This paper presents the results of the second phase of the benchmark exercise in the Plutonium Management for More Agility project (PuMMA), which aims to enhance the reliability of fuel performance codes (FPCs) for Mixed Oxide (MOX) fuels with high plutonium content. This phase involved the simulation of the CAPRIX irradiation (45% of Pu) carried out in the PHENIX reactor in France. Thirteen organizations from nine countries participated, using eight different fuel performance codes. In this phase, each code then improved its modelling by optimizing parameters and/or improving the formulation of some models involved in the computation. A comparative analysis was then conducted between the codes and the post-irradiation experimental (PIE) results, with the objective of identifying the improvements made, the discrepancies still observed, and to define future axes for improvement and development. The present paper demonstrates the results for the CAPRIX pin.The main results despite the discrepancies noted between the codes and/or with the experimental data, the work carried out through this Benchmark has demonstrated the capacity of the codes to optimise/improve their simulation in order to approximate the PIE. Moreover, the methodology employed yielded a comprehensive and precise analysis, thus facilitating the suggestion of approaches to the development of specific models. This finding thus demonstrates the potential of fuel performance codes modelling pins with high Pu contents.

Accurate prediction of fission gas release is essential to support fuel performance assessments across several plant states. This work presents an integral-scale assessment of the TRANSURANUS-SCIANTIX coupling using the updated modelling framework developed in the past two years. The assessment is performed through both code-to-data comparison against fuel central temperature, end-of-life fission gas release, rod internal pressure and cladding profilometry, and a supplementary code-to-code comparison against the previous TRANSURANUS-SCIANTIX (2024) version. In addition to model development, also the validation matrix has been extended with respect to the latest validation effort; it comprises water-reactor fuel rods under normal operation, anticipated operational occurrences, and design-basis-accident conditions represented by reactivity-initiated accident tests. In normal operation conditions, the model well reproduces the measured fission gas release, also providing improved kinetics thanks to athermal FGR modelling. Under accidental conditions, the physics-based model dedicated to grain-boundary micro-cracking reduces excessive activation of burst release at each power perturbation, as it is demonstrated via additional instrumented cases. When rod internal pressure data are available, the comparison highlights that reproducing pressure transients requires not only adequate release kinetics but also modelling gas transport toward the plenum. In general, the code predicts satisfactorily moderate releases, while high release conditions can be challenging. Cladding profilometry comparisons show a systematic tendency to overestimate the final cladding diameter, motivating future refinements in fuel swelling, relocation and recovery, as well as creep response. Overall, the code demonstrates a satisfactory performance of the fission-gas-related quantities.

Fission gas release modelling is a key aspect of fuel performance assessment, and quantifying the associated uncertainties remains a challenge. This work performs two sensitivity analyses to build confidence in recently developed athermal and burst fission gas release models. The computational framework adopted for fuel performance simulations couples the fission gas behaviour module of SCIANTIX with the TRANSURANUS fuel performance code. This system is interfaced with Dakota software to perform a global sensitivity analysis based on variance decomposition. Five parameters are investigated for the athermal release in the US PWR 16x16 base-irradiation case, i.e.: fabrication porosity, grain-edge length, burn-up, temperature, and fission rate. Results highlight the dominant role of fabrication porosity, suggesting the need for future modelling efforts on fuel densification. For the burst gas release from fuel microcracks five parameters are studied in the HATAC C2 rod irradiation experiment, i.e.: grain-boundary energy, bubble radius and internal pressure, temperature, and hydrostatic stress. Grain-boundary energy, which controls the microcracking threshold, emerges as the most relevant parameter. Furthermore, the analyses highlight the need for enhanced lower-scale physical modelling to further reduce empirical assumptions. The outcome of this study is twofold: an operational range of parameter values resulting in a calculated fission gas release within a factor two from the experimental value, and a methodology for testing and interpreting model behaviour within the TRANSURANUS–SCIANTIX framework prior to engineering-scale validation.

This work presents and validates a novel approach for modelling non-inert fission products behaviour during a power transient. The proposed model, MARGARET Active Fission Products, leverages all the microstructural quantities computed by MARGARET—a code originally developed to evaluate the release of inert fission products under normal and incidental loading sequences—to simulate the production, decay, and transport of non-inert fission products. The MARGARET Active Fission Products model is capable of handling 75 different isotopes, ranging from short-lived to long-lived species. The tendency of non-inert fission products to form different chemical compounds in different phases within the fuel compared to inert ones (always gaseous at equilibrium), requires the incorporation of thermochemistry in the calculations, since it directly impacts their release kinetics from the fuel. The thermochemical behaviour of fission products is included by performing calculations with the OpenCalphad thermochemical solver and the TAF-ID database. In this coupled thermochemistry-fission gas release approach, thermochemistry leads to the assessment of the quantity of non-inert fission products in the gas phase, that will percolate towards the free volume and, consequently, be released in the fuel rod. To validate the model, a power transient simulation was performed, replicating an experiment conducted on a refabricated fuel rod in the OSIRIS experimental reactor. The simulation results are compared with experimental data, including the total xenon release (measured by puncturing), the release of 137Cs, 131I, and 132Te at each interpellet (obtained from gamma spectrometry), and the radial distribution of xenon in the pellet, before and after the power ramp (analysed by Secondary-Ion Mass Spectrometry measurements). Furthermore, a detailed discussion on the thermochemical results is provided.

Accurate thermochemical modelling is essential for predicting the behaviour of nuclear fuels during both normal operation and accident conditions. However, direct coupling of Fuel Performance Codes with CALPHAD-based thermochemistry solvers remains computationally expensive and can hinder their use in iterative or multi-physics simulations. In this work, we develop a machine-learning approach that accelerates the evaluation of phase fractions and thermodynamic quantities in mixed oxide fuel by training neural-network surrogates on a large dataset of CALPHAD equilibrium calculations. The resulting models accurately reproduce the highly nonlinear correlations between fuel composition, temperature and thermodynamic behaviour, providing fast and reliable predictions without the numerical overhead or convergence issues typically associated with thermochemistry solvers. The methodology is implemented within SCIANTIX, a meso-scale code dedicated to the mechanistic modelling of fission gas behaviour in oxide fuels. The integration of the neural-network surrogates into SCIANTIX enables on-the-fly thermochemistry evaluations during fuel simulations with a substantial computational speedup, offering a practical pathway towards incorporating comprehensive thermochemical effects into fission gas models. This work therefore provides a general and scalable framework for embedding thermochemistry into nuclear fuel performance tools and represents a significant step towards the development of a full thermochemistry module applicable to irradiated fuel systems.

Foreseen applications of nuclear fuel operating at lower temperatures than current reactors call for the description of specific phenomena occurring in this temperature range. Among these, athermal fission gas release is of particular interest since it is expected to dominate the overall fission gas behaviour at low temperatures. This holds particular significance for certain Small Modular Reactor designs and even for fast neutron reactors employing fuels with high thermal conductivity, such as nitride fuels. Additionally, its implications extend to normal irradiation conditions in light water reactors and to spent fuel storage conditions. In this work, an AI-enhanced physics-based model of the athermal fission gas behaviour is presented, extending mechanistic models available in the open literature. The athermal release, defined as the fraction of gas vented from the fuel through its open porosity, is accounted for via the solution of the gas diffusion within the fuel grain, evaluating the fraction of the concentration gradient in the proximity of grain edges. The results of such computation are included in the SCIANTIX code thanks to a dedicated neural network, aimed at encapsulating the complex dependencies affecting the athermal release, whilst ensuring a computational time in line with fuel performance applications. Additionally, in the context of this analysis, semi-empirical models for solid fission products swelling and fuel densification are incorporated as well, providing a modelling suite for low-temperature conditions. The consistency of the model is tested with data available in the literature.

Multi-scale methodologies have been developed and applied successfully in the frame of nuclear fuel performance analyses, but the complexity of the tools involved hinders their extensive application. Gaps in modelling capabilities of specific input/outputs in particular limits code-to-code communication. In this work, we propose a multi-fidelity methodology to tackle this issue. The application presented here concerns the inclusion of a meso-scale module describing fission gas behaviour (SCIANTIX) in a fuel performance code (FRAPCON). A critical input parameter of the meso-scale module, the local hydro-static stress in the fuel, is not predicted by such fuel performance code, hence limiting this coupling. This gap is filled by using a second fuel performance code (TRANSURANUS) to construct a virtual dataset of local hydro-static stress values, on which an artificial neural network is trained and included in the FRAPCON/SCIANTIX coupled suite. This multi-fidelity methodology is demonstrated by simulating the Ris & oslash; AN3 irradiation experiment.

Overpressurisation of gas-filled bubbles and pores is considered the main driver of oxide fuel fragmentation and fission gas release into the rod free volume of irradiated fuel during transients. In this framework, advanced modelling of fission gas behaviour is crucial to enhance the predictive capabilities of fuel performance codes. This study develops a physics-based model for fission gas release from grain boundaries in UO2 fuel and implements it into SCIANTIX, an open-source code developed at Politecnico di Milano to simulate fission gas behaviour in nuclear fuels. The model first describes gas release through continuous bubble networks at the grain face, exploiting data from irradiated fuel. It then focuses on gas release from damaged grain boundaries, applying fracture mechanics to predict micro-cracking induced by bubble overpressurisation. Finite elements simulations are performed using ABAQUS software, in order to assess stress intensification as a function of bubble density, shape, and size. The model is assessed against three separate-effect experiment datasets, including annealing tests and grain-face observations via scanning electron microscopy, demonstrating promising predictive capabilities for gaseous swelling and fission gas release. This work provides a valuable tool for improving fission gas behaviour modelling through a physics-based approach and lays the groundwork for future extensions to the high burn-up structure, offering a framework for a more comprehensive description of fuel fragmentation.

DarkSide-20k is a novel liquid argon dark matter detector currently under construction at the Laboratori Nazionali del Gran Sasso (LNGS) of the Istituto Nazionale di Fisica Nucleare (INFN) that will push the sensitivity for Weakly Interacting Massive Particle (WIMP) detection into the neutrino fog. The core of the apparatus is a dual-phase Time Projection Chamber (TPC), filled with 50 tonnes of low radioactivity underground argon (UAr) acting as the WIMP target. NUV-HD-cryo Silicon Photomultipliers (SiPM)s designed by Fondazione Bruno Kessler (FBK) (Trento, Italy) were selected as the photon sensors covering two $$10.5~\text {m}^2$$ 10.5 m 2 Optical Planes, one at each end of the TPC, and a total of $$5~\text {m}^2$$ 5 m 2 photosensitive surface for the liquid argon veto detectors. This paper describes the Quality Assurance and Quality Control (QA/QC) plan and procedures accompanying the production of FBK NUV-HD-cryo SiPM wafers manufactured by LFoundry s.r.l. (Avezzano, AQ, Italy). SiPM characteristics are measured at 77 K at the wafer level with a custom-designed probe station. As of March 2025, 1314 of the 1400 production wafers (94% of the total) for DarkSide-20k were tested. The wafer yield is $$93.2\pm 2.5$$ 93.2 ± 2.5 %, which exceeds the 80% specification defined in the original DarkSide-20k production plan.

Developing fuel performance codes requires lengthy technical and legal verification procedures. This work proposes an approach to automate and facilitate these processes by using a data assimilation method. The method considered is Gaussian process regression, which allows for exploiting experimental data to correct/update material property correlations in fuel performance codes. As a demonstration of this method, the correlation for the single-atom xenon diffusivity is updated in the SCIANTIX code with recent lower-length scale results, without the need to modify the SCIANTIX code itself. The goal is to demonstrate the validity of this approach for data-driven developments of physics-based fuel performance codes.

DarkSide-20k (DS-20k) is a dark matter detection experiment under construction at the Laboratori Nazionali del Gran Sasso (LNGS) in Italy. It utilises similar to 100 t of low radioactivity argon from an underground source (UAr) in its inner detector, with half serving as target in a dual-phase time projection chamber (TPC). The UAr cryogenics system must maintain stable thermodynamic conditions throughout the experiment's lifetime of over 10 years. Continuous removal of impurities and radon from the UAr is essential for maximising signal yield and mitigating background. We are developing an efficient and powerful cryogenics system with a gas purification loop with a target circulation rate of 1000 slpm. Central to its design is a condenser operated with liquid nitrogen which is paired with a gas heat exchanger cascade, delivering a combined cooling power of more than 8 kW. Here we present the design choices in view of the DS-20k requirements, in particular the condenser's working principle and the cooling control, and we show test results obtained with a dedicated benchmarking platform at CERN and LNGS. We find that the thermal efficiency of the recirculation loop, defined in terms of nitrogen consumption per argon flow rate, is 95% and the pressure in the test cryostat can be maintained within +/-(0.1-0.2) mbar. We further detail a 5-day cool-down procedure of the test cryostat, maintaining a cooling rate typically within -2K/h, as required for the DS-20k inner detector. Additionally, we assess the circuit's flow resistance, and the heat transfer capabilities of two heat exchanger geometries for argon phase change, used to provide gas for recirculation. We conclude by discussing how our findings influence the finalisation of the system design, including necessary modifications to meet requirements and ongoing testing activities.

Minor Actinides (MAs) are produced inside nuclear reactor fuels under irradiation and are responsible of large part of the waste radiotoxicity. Partitioning and Transmutation (P&T) is identified as the strategy that can relax constraints on geological disposal, since transmutation of MAs via critical or subcritical fast spectrum irradiation facilities like MYRRHA is a valid path to reduce the waste hazard. One of the focuses of the PATRICIA Project is the development of fuel performance codes towards Am-bearing fuels and the computational study of their behaviour under irradiation, supporting the licencing and development of advanced designs of the MYRRHA reactor. This work focuses on the modelling of thermophysical properties of transmutation-type fuel concepts for fast reactor application, specifically of melting temperature, specific heat capacity and thermal conductivity of (U,Pu,Am)O2-x and (U,Am)O2-x. The modelling activity involves the latest atomic-scale/CALPHAD data on Am-bearing fuels achieved in the framework of PATRICIA. Developed models are verified and validated against separate-effect data, and implemented in the TRANSURANUS fuel performance code to simulate a past irradiation experiment (HEDL P-19 power-to-melt transient test) for the evaluation of the impact of new models on the pin performance and their integral validation.

Abstract The knowledge of scintillation quenching of $$\alpha $$ α -particles plays a paramount role in understanding $$\alpha $$ α -induced backgrounds and improving the sensitivity of liquid argon-based direct detection of dark matter experiments. We performed a relative measurement of scintillation quenching in the MeV energy region using radioactive isotopes ( $$^{222}$$ 222 Rn, $$^{218}$$ 218 Po and $$^{214}$$ 214 Po isotopes) present in trace amounts in the DEAP-3600 detector and quantified the uncertainty of extrapolating the quenching factor to the low-energy region.

Mechanistic multi-scale modelling holds the potential to inform fuel performance codes by incorporating high-fidelity models, algorithms, parameters, and material properties. In this context, meso-scale codes emerge as valuable tools for developing detailed models and performing separate verification and validation steps. This work focuses on SCIANTIX, an open-source 0D meso-scale code designed to describe the behaviour of gaseous and volatile fission products in nuclear oxide fuel. The code predominantly employs engineering physics-based behavioural models featuring computational times that align with typical fuel performance code requirements. Given the numerical foundation of the code, it is applicable to both stationary and transient conditions. Following a recent work outlining the standalone SCIANTIX (version 2.0) performance and its separate-effect validation database, we present its performance when coupled with fuel performance codes to simulate light water reactor fuel rods. The experiments selected for the comparative analysis constitute an initial integral validation database. The comparison focuses on conventional engineering quantities of interest, such as integral fission gas release, demonstrating the satisfactory performance of the code. Additionally, it highlights the potential advantages of multi-scale modelling over conventional semi-empirical approaches.

Minor actinides are the main contributors to medium- and long-term radiotoxicity and heat production in spent nuclear fuels. Research efforts are currently ongoing to explore different options to dispose of such radionuclides, e.g., their burning in fast reactors within mixed -oxide fuels. The MYRRHA sub -critical reactor is one of the future facilities with envisaged burning and transmutation capabilities. This work assesses the thermal-mechanical performance of a homogeneous Am -bearing fuel pin both in the In -Pile test Section position of the MYRRHA "Revision 1.8" core and under driver irradiation. The normal operating conditions of MYRRHA are considered, with a focus on the safety design limits and involving sensitivity analyses to evaluate the impact of increasing americium contents (in the range 0-5 wt%) on safety -relevant simulation outcomes. The simulations are performed with the TRANSURANUS fuel performance code (version v1m4j22) coupled with the SCIANTIX physics -based module for inert gas behaviour, and rely on a dedicated surrogate model for the helium source term during MYRRHA irradiation accounting for the relevant contribution of the fuel americium enrichment, besides advanced models for the properties and behaviour of the specific pin materials. The analyses reveal the suitability and safety under irradiation of MOX fuels with low Am enrichments according to the current MYRRHA design.

The present work describes original experimental results obtained on the melting and solidification behaviour of mixed PuO2 - Fe3O4 samples. The samples were laser-heated into the liquid phase in a pressurised autoclave under a controlled atmosphere of argon or air, whereby the sample temperature was measured in-situ with the help of fast pyrometers. In addition, the oxygen release in the argon experiments was measured by an O2 probe. By these means, solidus and liquidus temperatures in the Fe3O4-PuO2 system were derived as a function of the PuO2 mole fraction at two oxygen partial pressures (pO2 = 1e-6 and pO2 = 0.2 bar). It was shown that the system exhibits a eutectic reaction, and that the observed liquidus and eutectic temperatures are higher for higher oxygen contents. This is due to the stabilisation of stoichiometric oxides in an oxidising atmosphere, as opposed to the appearance of oxygen-defective, lower-melting phases in an inert/reducing environment. The occurrence of the eutectic in the system was confirmed by electron microscopy analysis. It occurs at a composition between 22 mol % and 23 mol % of PuO2 in the Fe3O4-PuO2 pseudo-binary system. Outside the Fe3O4 - PuO2 plane, the eutectic temperature was observed to depend upon the oxygen potential, varying from approximately 1800 K in air to approximately 1700 K in inert (Ar) gas. The experimental solidus/liquidus data were used to derive a thermodynamic model on the system. Thermodynamic calculations supported the oxygen release of the samples in both argon and air and the interpretation of the quenched microstructure of the samples.

This chapter summarizes some specific aspects concerning the thermal hydraulic analysis of liquid-fueled molten salt reactors (MSRs), including the presence of an internally heated fluid that serves both as fuel and coolant. Throughout the chapter, in order to focus the attention on some fundamentals from a physical point of view, reference is made to very simple case studies, namely, a typical graphite-moderated MSR channel and two single-phase natural circulation rectangular loops. The effects of internal heat generation on heat transfer, pressure losses, and stability of natural circulation (both in laminar and turbulent regimes) are investigated. The presented case studies also offer a useful analytical framework to assess more sophisticated numerical simulation tools, in view of their adoption for more realistic and complex 3-D analyses. Solid-fueled MSRs are not considered in this chapter as they feature different thermohydraulic issues, with the molten salt acting only as a coolant.

The SPHERE experiment aimed at studying the behaviour of Minor Actinide-bearing Driver Fuel (U,Pu,Am)O2-x by comparing sphere-packed and pelletized fuels. The irradiation experiment was performed in the High Flux Reactor at Petten from August 2013 to April 2015, and was followed by post-irradiation examinations up to mid-2017. The present work consists in a new analysis of the SPHERE experiment, focusing on the pelletized fuel, by the means of both neutronics and fuel performance codes. This study is performed in the frame of the European Project PATRICIA. The adopted methodology and the main results achieved, assessed in particular against inert gas-related experimental data, are presented in the paper.