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The design phase and safety assessment of Generation IV liquid metal-cooled fast reactors calls for the improvement of fuel pin performance codes, in particular the enhancement of their predictive capabilities towards uranium-plutonium mixed oxide fuels and stainless-steel cladding under irradiation in fast reactor environments. To this end, the current capabilities of fuel performance codes must be critically assessed against experimental data from available irradiation experiments. This work is devoted to the assessment of three European fuel performance codes, namely GERMINAL, MACROS and TRANSURANUS, against the irradiation of two fuel pins selected from the SUPERFACT-1 experimental campaign. The pins are characterized by a low enrichment (~ 2 wt.%) of minor actinides (neptunium and americium) in the fuel, and by plutonium content and cladding material in line with design choices envisaged for liquid metal-cooled Generation IV reactor fuels. The predictions of the codes are compared to several experimental measurements, allowing the identification of the current code capabilities in predicting fuel restructuring, cladding deformation, redistribution of actinides and volatile fission products. The integral assessment against experimental data is complemented by a code-to-code benchmark focused on the evolution of quantities of engineering interest over time. The benchmark analysis points out the differences in the code predictions of fuel central temperature, fuel-cladding gap width, cladding outer radius, pin internal pressure and fission gas release and suggests potential modelling development paths towards an improved description of the fuel pin behaviour in fast reactor irradiation conditions.

In light of the importance of helium production in influencing the behaviour of fast reactor fuels, in this work we present a burn-up module with the objective to calculate the production of helium in both in-pile and out-of-pile conditions tracking the evolution of 23 alpha-decaying actinides. This burn-up module relies on average microscopic cross-section look-up tables generated via SERPENT high-fidelity calculations and involves the solution of the system of Bateman equations for the selected set of actinide nuclides. The results of the burn-up module are verified in terms of evolution of actinide and helium concentrations by comparing them with the high-fidelity ones from SERPENT, considering two representative test cases of (U,Pu)O2 fuel in fast reactor conditions. In addition, a code-to-code comparison is made with the independent state-of-the-art module TUBRNP (implemented in the TRANSURANUS fuel performance code) for the same test cases. The herein presented burn-up module is available in the SCIANTIX code, designed for coupling with fuel performance codes.

Radiation-induced embrittlement of reactor pressure vessel (RPV) steels can potentially limit the operating life of nuclear power plants. Over extended exposure to radiation doses, these body-centered cubic (BCC) irons demonstrate irradiation damage. Here, we present a continuum dislocation density (CDD) crystal plasticity model to capture the interaction among dislocations and self-interstitial atom (SIA) loops in α-iron. We demonstrate the importance of modeling cross slip using a combined stochastic Monte Carlo approach and the role of slip system strength anisotropy in capturing stochastic cross slip interactions. Through these captured interactions, the CDD crystal plasticity model can capture both the stress response and the physical evolution of dislocations on different slip system planes. Single-crystal verification experiments are used to calibrate the CDD crystal plasticity model, and a set of simplified polycrystalline simulations demonstrates the model’s ability to capture the stress response from tensile experiments on α-iron.

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.

In this work, we propose a new mechanistic model for the treatment of helium behaviour which includes the description of helium solubility in oxide fuel. The proposed model has been implemented in SCIANTIX and validated against annealing helium release experiments performed on small doped fuel samples. The overall agreement of the new model with the experimental data is satisfactory, and given the mechanistic formulation of the proposed model, it can be continuously and easily improved by directly including additional phenomena as related experimental data become available.

The physics-based fission gas behaviour model available in the BISON fuel performance code provides satisfactory predictive capabilities for application to light-water reactor conditions. In this work, we present a model extension for application to fast reactor (U,Pu)O 2 fuel. In particular, we detail the introduction of a lower bound to the number density of grain-face bubbles, representing a limit to the coalescence process once extensive bubble interconnection is achieved. This new feature is tested first against an experimental database for UO 2 -LWR, and secondly is validated against integral irradiation experiments for fast reactor (U,Pu)O 2 fuel rods irradiated in the FFTF (Fast Flux Test Facility) and in the JOYO reactors. The comparisons of BISON results with the experimental data are satisfactory and demonstrate an improvement compared to the standard version of the code.

Bridging lower length-scale calculations with the engineering-scale simulations of fuel performance codes requires the development of dedicated intermediate-scale codes. In this work, we present SCIANTIX, an open source 0D stand-alone computer code designed to be included/coupled as a module in existing fuel performance codes. The models currently available in SCIANTIX cover intra- and inter-granular inert gas behaviour in UO2, and high burnup structure formation as well. Showcases of validation in both constant and transient conditions are presented in this work. As for the numerical treatment of the model equations, SCIANTIX is developed with full numerical consistency and entirely verified with the method of manufactured solutions – verification of different numerical solvers is also showcased in this work.

We propose a model describing the HBS formation and the progressive intra-granular xenon depletion in UO2. The HBS formation is modeled employing the Kolmogorov-Johnson-Mehl-Avrami (KJMA) formalism for phase transformations, which has been fitted to experimental data on the restructured volumetric fraction as a function of the local effective burnup. To this end, we employed available experimental data and novel data extracted in this work. The HBS formation model is coupled to a description of the intra-granular fission gas behavior, allowing to estimate the evolution of the retained xenon in order to consistently compute fission gas retention and its effect on the fuel matrix swelling. The satisfactory agreement of the model predictions to experimental data and state-of-the-art models’ results, in terms of both xenon depletion and fuel matrix swelling as a function of the local burnup, paves the way to the inclusion of the model in fuel performance codes.

The description of intra-granular fission gas behaviour is a fundamental part of any model for the prediction of fission gas release and swelling in nuclear fuel. In this work we present a model describing the evolution of intra-granular fission gas bubbles in terms of bubble number density and average size, coupled to gas release to grain boundaries. The model considers the fundamental processes of single gas atom diffusion, gas bubble nucleation, re-solution and gas atom trapping at bubbles. The model is derived from a detailed cluster dynamics formulation, yet it consists of only three differential equations in its final form; hence, it can be efficiently applied in engineering fuel performance codes while retaining a physical basis. We discuss improvements relative to previous single-size models for intra-granular bubble evolution. We validate the model against experimental data, both in terms of bubble number density and average bubble radius. Lastly, we perform an uncertainty and sensitivity analysis by propagating the uncertainties in the parameters to model results.

Uranium silicide ( U 3 Si 2 ) is a concept explored as a potential alternative to UO 2 fuel used in light water reactors (LWRs) since it may improve accident tolerance and economics due to its higher thermal conductivity and increased uranium density. U 3 Si 2 has been previously used in research reactors in the form of dispersion fuel which operates at lower temperatures than commercial LWRs. The research reactor data illustrated that significant gaseous swelling occurs as the fuel burnup increases. Therefore, it is imperative to understand the fission gas behavior of U 3 Si 2 under higher temperature LWR operating conditions. In this work, molecular dynamics and phase-field modeling techniques are used to reduce the uncertainty in select modeling assumptions made in developing the fission gas behavior model for U 3 Si 2 in the BISON fuel performance code. To support the implementation of a U 3 Si 2 fission gas model in BISON, cluster dynamics simulations of irradiation enhanced Xe diffusion have been carried out. Similarly, MD simulations were used to predict the athermal contribution due to atomic mixing during ballistic damage cascades. By combining our results with literature DFT data for thermal equilibrium diffusion, Xe diffusivity has been described over a wide range of temperatures for in-reactor conditions. These lower length scale informed models are then utilized in the assessment of BISON U 3 Si 2 modeling capabilities by simulating the ATF-1 experiments irradiated in the Advanced Test Reactor (ATR). Sensitivity analysis (SA) and uncertainty quantification (UQ) are included as part of the assessment process to identify where further experiments and lower length scale modeling would be beneficial. The multiscale modeling approach utilized in this work can be applied to new fuel concepts being explored for both LWRs and advanced reactors (e.g., uranium nitride, uranium carbide).

In the rim zone of UO2 nuclear fuel pellets, the combination of high burnup and low temperature drives a microstructural change, leading to the formation of the high burnup structure (HBS). In this work, we propose a semi-empirical model to describe the formation of the HBS, which embraces the polygonisation/recrystallization process and the depletion of intra-granular fission gas, describing them as inherently related. For this purpose, we performed grain-size measurements on samples at radial positions in which the restructuring was incomplete. Based on these new experimental data, we infer an exponential reduction of the average grain size with local effective burnup, paired with a simultaneous depletion of intra-granular fission gas driven by diffusion. The comparison with currently used models indicates the applicability of the herein developed model within integral fuel performance codes.

In this work, we present a model of fission gas behavior in U3Si2 under light water reactor (LWR) conditions for application in engineering fuel performance codes. The model includes components for intra-granular and inter-granular behavior of fission gases. The intra-granular component is based on cluster dynamics and computes the evolution of intra-granular fission gas bubbles and swelling coupled to gas diffusion to grain boundaries. The inter-granular component describes the evolution of grain-boundary fission gas bubbles coupled to fission gas release. Given the lack of experimental data for U3Si2 under LWR conditions, the model is informed with parameters calculated via atomistic simulations. In particular, we derive fission gas diffusivities through density functional theory calculations, and the re-solution rate of fission gas atoms from intra-granular bubbles through binary collision approximation calculations. The developed model is applied to the simulation of an experiment for U3Si2 irradiated under LWR conditions available from the literature. Results point out a credible representation of fission gas swelling and release in U3Si2. Finally, we perform a sensitivity analysis for the various model parameters. Based on the sensitivity analysis, indications are derived that can help in addressing future research on the characterization of the physical parameters relative to fission gas behavior in U3Si2. The developed model is intended to provide a suitable infrastructure for the engineering scale calculation of fission gas behavior in U3Si2 that exploits a multiscale approach to fill the experimental data gap and can be progressively improved as new lower-length scale calculations and validation data become available.

In oxide nuclear fuel, the temperature gradient from the centerline to the radial edge of the pellet induces cracking due to thermal stress. We present a model that represents the effect of cracking as an isotropic softening of the material. The model considers a reduction of the elastic constants as a function of the number of cracks present in the fuel. This pragmatic approach aims to represent the average effect of cracking on the fuel stresses without attempting to explicitly describe the crack pattern or localization. Albeit simplistic, this approach has definite advantages in terms of computational expense and numerical convergence behavior for the mechanical analysis. It is also consistent with the uncertainties inherent in modeling the fuel cracking process, and suitable for engineering calculations aimed at representing the global fuel element behavior. The applied scaling of the elastic constants conserves the principal strain values and redistributes the principal stresses isotropically. Moreover, the model allows for the effect of an evolving number of cracks. In particular, an empirical correlation for the number of cracks as a function of the rod average linear heat rate is developed. Although the basic concept of the model was known, in this work we revisit and improve the original formulation, and implement the model in the BISON fuel performance code. Application in BISON is demonstrated through simulations of an idealized single fuel pellet irradiation and two integral fuel rod experiments. Results showcase the impact on calculated fuel stresses, the coupling to fuel creep, and the effect on cladding strains during pellet-cladding mechanical interaction.

This work applies reconstruction methods based on a genetic algorithm to derive 3D material properties, namely porosity and percolation fraction, in irradiated U-Pu-Zr fuel with minor actinides. We provide two-dimensional experimental data regarding the radial distribution of fission gas bubbles in the fuel and apply the algorithm successfully developed in a companion paper to reconstruct the fuel pore structure in 3D which is unknown a priori. The algorithm returned a set of best structures that constituted the best candidate solutions representing the pore phase. From these, it was possible to extract statistics on the 3D percolation fraction of the reference medium and infer a mean value, the related uncertainty, and an upper and lower bound of the percolation fraction. The algorithm proved able to infer this 3D property from 2D information of the metallic fuel with confidence intervals, thus establishing a path to infer 3D properties directly from 2D experimental images. The knowledge of such a relationship can be used to extrapolate the percolation threshold with confidence interval, which is a crucial property in defining microstructure-based fission gas release models of metallic fuels.

A new methodology is introduced to analyse porosity data in the high burnup structure. Image analysis is coupled with the adaptive kernel density estimator to obtain a detailed characterisation of the pore size distribution, without a-priori assumption on the functional form of the distribution. Subsequently, stereological analysis is carried out. The method shows advantages compared to the classical approach based on the histogram in terms of detail in the description and accuracy within the experimental limits. Results are compared to the approximation of a log-normal distribution. In the investigated local burnup range (80–200 GWd/tHM), the agreement of the two approaches is satisfactory. From the obtained total pore density and mean pore diameter as a function of local burnup, pore coarsening is observed starting from ≈100 GWd/tHM, in agreement with a previous investigation.

The transport of fission gas from within the fuel grains to the grain boundaries (intra-granular fission gas release) is a fundamental controlling mechanism of fission gas release and gaseous swelling in nuclear fuel. Hence, accurate numerical solution of the corresponding mathematical problem needs to be included in fission gas behaviour models used in fuel performance codes. Under the assumption of equilibrium between trapping and resolution, the process can be described mathematically by a single diffusion equation for the gas atom concentration in a grain. In this paper, we propose a new numerical algorithm (PolyPole-1) to efficiently solve the fission gas diffusion equation in time-varying conditions. The PolyPole-1 algorithm is based on the analytic modal solution of the diffusion equation for constant conditions, combined with polynomial corrective terms that embody the information on the deviation from constant conditions. The new algorithm is verified by comparing the results to a finite difference solution over a large number of randomly generated operation histories. Furthermore, comparison to state-of-the-art algorithms used in fuel performance codes demonstrates that the accuracy of PolyPole-1 is superior to other algorithms, with similar computational effort. Finally, the concept of PolyPole-1 may be extended to the solution of the general problem of intra-granular fission gas diffusion during non-equilibrium trapping and resolution, which will be the subject of future work.

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.

The description of intra-granular fission gas behavior during irradiation is a fundamental part of models used for the calculation of fission gas release and gaseous swelling in nuclear fuel performance codes. The relevant phenomena include diffusion of gas atoms towards the grain boundaries coupled to the evolution of intra-granular bubbles. While intra-granular bubbles during normal operating conditions are limited to sizes of a few nanometers, experimental evidence exists for the appearance of a second population of bubbles during transients, characterized by coarsening to sizes of tens to hundreds of nanometers and that can significantly contribute to gaseous fuel swelling. In this work, we present a model of intra-granular fission gas behavior in uranium dioxide fuel that includes both nanometric fission gas bubble evolution and bubble coarsening during transients. While retaining a physical basis, the developed model is relatively simple and is intended for application in engineering fuel performance codes. We assess the model through comparisons to a substantial number of experimental data from SEM observations of intra-granular bubbles in power ramp tested uranium dioxide samples. The results demonstrate that the model reproduces the coarsening of a fraction of the intra-granular bubbles and correspondingly, predicts gaseous swelling during power ramps with a significantly higher accuracy than is allowed by traditional models limited to the evolution of nanometric intra-granular bubbles.

Helium plays an important role in determining nuclear fuel performance both in-pile (especially for MOX fuels and those at high burnup) and in storage conditions. Predictive models of helium behaviour are therefore a fundamental element in fuel performance codes. These models are based on the accurate knowledge of helium diffusivity (addressed in a previous paper, Luzzi et al. (2018)) and of helium solubility in oxide nuclear fuel. Based on all the experimental data available in the literature and after verification of the validity of Henry’s law we propose two correlations for Henry’s constant, k H ( at m - 3 MPa - 1 ) : k H = 1 . 8 · 10 25 exp - 0 . 41 / kT for powders and k H = 4.1 · 10 24 exp - 0 . 65 / kT for single crystals, with the Boltzmann factor 1/kT in ( eV - 1 ). The correlation for Henry’s constant in powder samples is of interest for the analysis of helium behaviour in the fuel after the pulverization occurring during LOCA-like temperature transients, while the correlation for Henry’s constant in single-crystals is usable in meso-scale models describing helium behaviour at the level of fuel grains. The current lack of data for this fundamental property, especially for poly-crystalline samples, calls for new experiments.

Fission gas release and gaseous swelling in nuclear fuel are driven by the transport of fission gas from within the fuel grains to grain boundaries (intra-granular fission gas release). The process involves gas atom diffusion in conjunction with trapping in and resolution from intra-granular bubbles, and is described mathematically by a system of two partial differential equations (PDE). Under the assumption of equilibrium between trapping and resolution (quasi-stationary approximation) the system can be reduced to a single diffusion equation with an effective diffusion coefficient. Numerical solutions used in engineering fuel performance calculations invariably rely on this simplification. First, we investigate the validity of the quasi-stationary approximation compared to the solution of the general system of PDEs. Results demonstrate that the approximation is valid under most conditions of practical interest, but is inadequate to describe intra-granular fission gas release during rapid transients to relatively high temperatures such as postulated reactivity-initiated accidents (RIA). Then, we develop a novel numerical algorithm for the solution of the general PDE system in time-varying conditions. We verify the PolyPole-2 algorithm against a reference finite difference solution for a large number of randomly generated operation histories including prototypical RIA transients. Results demonstrate that PolyPole-2 captures the solution of the general system with a high accuracy and a low computational cost. The PolyPole-2 algorithm overcomes the quasi-stationary approximation and the concept of an effective diffusion coefficient for the solution of the intra-granular fission gas release problem in nuclear fuel analysis.

The formation of the high burnup structure (HBS) is possibly the most significant example of the restructuring processes affecting commercial nuclear fuel in-pile. The HBS forms at the relatively cold outer rim of the fuel pellet, where the local burnup is 2–3 times higher than the average pellet burnup, under the combined effects of irradiation and thermo-mechanical conditions determined by the power regime and the fuel rod configuration. The main features of the transformation are the subdivision of the original fuel grains into new sub-micron grains, the relocation of the fission gas into newly formed intergranular pores, and the absence of large concentrations of extended defects in the fuel matrix inside the subdivided grains. The characterization of the newly formed structure and its impact on thermo-physical or mechanical properties is a key requirement to ensure that high burnup fuel operates within the safety margins. This paper presents a synthesis of the main findings from extensive studies performed at JRC-Karlsruhe during the last 25 years to determine properties and behaviour of the HBS. In particular, microstructural features, thermal transport, fission gas behaviour, and thermo-mechanical properties of the HBS will be discussed. The main conclusion of the experimental studies is that the HBS does not compromise the safety of nuclear fuel during normal operations.

Iron-chromium-aluminum (FeCrAl) alloys are candidates to be used as nuclear fuel cladding for increased accident tolerance. An analysis of the response of FeCrAl under normal operating and loss of coolant conditions has been performed using fuel performance modeling. In particular, recent information on FeCrAl material properties and phenomena from separate effects tests has been implemented in the BISON fuel performance code and analyses of integral fuel rod behavior with FeCrAl cladding have been performed. BISON simulations included both light water reactor normal operation and loss-of-coolant accidental transients. In order to model fuel rod behavior during accidents, a cladding failure criterion is desirable. For FeCrAl alloys, a failure criterion is developed using recent burst experiments under loss of coolant like conditions. The added material models are utilized to perform comparative studies with Zircaloy-4 under normal operating conditions and oxidizing and non-oxidizing out-of-pile loss of coolant conditions. The results indicate that for all conditions studied, FeCrAl behaves similarly to Zircaloy-4 with the exception of improved oxidation performance. Further experiments are required to confirm these observations.

The modelling of fission gas behaviour is a crucial aspect of nuclear fuel performance analysis in view of the related effects on the thermo-mechanical performance of the fuel rod, which can be particularly significant during transients. In particular, experimental observations indicate that substantial fission gas release (FGR) can occur on a small time scale during transients (burst release). To accurately reproduce the rapid kinetics of the burst release process in fuel performance calculations, a model that accounts for non-diffusional mechanisms such as fuel micro-cracking is needed. In this work, we present and assess a model for transient fission gas behaviour in oxide fuel, which is applied as an extension of conventional diffusion-based models to introduce the burst release effect. The concept and governing equations of the model are presented, and the sensitivity of results to the newly introduced parameters is evaluated through an analytic sensitivity analysis. The model is assessed for application to integral fuel rod analysis by implementation in two structurally different fuel performance codes: BISON (multi-dimensional finite element code) and TRANSURANUS (1.5D code). Model assessment is based on the analysis of 19 light water reactor fuel rod irradiation experiments from the OECD/NEA IFPE (International Fuel Performance Experiments) database, all of which are simulated with both codes. The results point out an improvement in both the quantitative predictions of integral fuel rod FGR and the qualitative representation of the FGR kinetics with the transient model relative to the canonical, purely diffusion-based models of the codes. The overall quantitative improvement of the integral FGR predictions in the two codes is comparable. Moreover, calculated radial profiles of xenon concentration after irradiation are investigated and compared to experimental data, illustrating the underlying representation of the physical mechanisms of burst release.

As awareness towards the problem is growing, eco-friendliness is today a paramount requirement for all space activities and in particular for the ground segment, fully comparable to other industrial sectors. The present work focuses on the assessment and the sustainable development enhancement of a ground-based space facility, the European Astronaut Centre (EAC), located in Germany. The project is framed within the European Space Agency development of an environmental outlook, which aims not only at the full compliance with the legislation and at assessing the impact of its activities, but also at laying the foundation for future evolution through innovation. Indeed, ESA promotes the sustainable use of space as a necessity and duty for Europe. As history teaches us, technical knowledge emerged within the space sector serves as innovation driver in other industrial branches: the goal of the project is to transform the EAC building into a spaceship integrated with the territory through the conscious management of this spontaneous process, fostering the combination between the space sector and the architecture and civil engineering fields. The work explores the potential of space technologies, processes and systems applied on ground and presents a range of space-driven innovative concepts which may improve the sustainability of the EAC building, focusing on different aspects of its resource demand – energy, water and waste management – and defining the integration with the pre-existing compound, the limitation of the impact on the surrounding landscape and the participation of the local community as additional fundamental requirements. Indeed, the project embraces the full concept of sustainability, which considers not only eco-friendliness but also its balance with economic and social aspects. Two factors – a certain urgency for action, which leaves little space for research and experimentation, and a call for ground-breaking solutions – guided the design activity: taking advantage of these conflicting requirements, a comparison between standard technologies and innovative space-related concepts was performed. When dealing with complex and uncertain scenarios, decision among the possible solutions is not straightforward and needs to be supported by appropriate methodologies: a multi-criteria and quantitative decision-making tool, able to concentrate on the main goal while considering all other relevant aspects – environmental, economic, social sustainability – was therefore developed. Furthermore, the project promotes local community participation in the decisional process, as a way to enhance knowledge, generate understanding and promote towards the EAC redesign, space activities and their potential innovative impact on sustainability.

The Advanced Lead Fast Reactor European Demonstrator (ALFRED) is the 300 MWth pool-type reactor aimed at proving the feasibility of the design concept adopted for the European Lead-cooled Fast Reactor (ELFR) of Generation-IV, whose preliminary design has been proposed within the LEADER (Lead Cooled European Advanced Demonstration Reactor) EURATOM Project. In the frame of investigating different options to optimize the conceptual fuel pin design of ALFRED, the fuel performance analysis of the reactor core is carried out by means of the TRANSURANUS code, which is presented in this paper. Results of the average and the hottest reactor conditions are discussed concerning both fuel and cladding performance on the basis of indicative design limits. A sensitivity analysis is performed in order to assess the impact of models and parameters affected by a larger uncertainty, and to identify a “worst case” scenario. This strategy allows identifying some design parameters which could be modified to improve the fuel pin behaviour, in order to gain benefits mainly in terms of maximum fuel temperature. The results of the present work are useful for giving feedback to the conceptual design of the ALFRED reactor and, in perspective, to improve the safety-by-design characteristics of the LFR systems.

Vickers microhardness ( HV 0.1 ) and Young's modulus ( E ) measurements of LWR UO 2 fuel at burn-up ≥60 GWd/tHM are presented. Their ratio HV 0.1 / E was found constant in the range 60–110 GWd/tHM. From the ratio and the microhardness values vs porosity, the Young's modulus dependence on porosity was derived and extended to the full radial profile, including the high burn-up structure (HBS). The dependence is well represented by a linear correlation. The data were compared to fuel performance codes correlations. A burn-up dependent factor was introduced in the Young's modulus expression. The modifications extend the experimental validation range of the TRANSURANUS correlation from un-irradiated to irradiated UO 2 and up to 20% porosity. First simulations of LWR fuel rod irradiations were performed in order to illustrate the impact on fuel performance. In the specific cases selected, the simulations suggest a limited effect of the Young's modulus decrease due to burn-up on integral fuel performance.