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The behaviour of the fission gas plays an important role in the fuel rod performance. In a previous work, we presented a physics-based model describing intra- and inter-granular behaviour of radioactive fission gas. The model was implemented in SCIANTIX, a mesoscale module for fission gas behaviour, and assessed against the CONTACT 1 irradiation experiment. In this work, we present the multi-scale coupling between the TRANSURANUS fuel performance code and SCIANTIX, used as mechanistic module for stable and radioactive fission gas behaviour. We exploit the coupled code version to reproduce two integral irradiation experiments involving standard fuel rod segments in steady-state operation (CONTACT 1) and during successive power transients (HATAC C2). The simulation results demonstrate the predictive capabilities of the code coupling and contribute to the integral validation of the models implemented in 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 interrelated 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.

The sodium-cooled fast reactor is among the innovative nuclear technologies selected in the framework of the development of Generation IV concepts, allowing the irradiation of uranium-plutonium mixed oxide fuels (MOX). A fundamental step for the safety assessment of MOX-fuelled pins for fast reactor applications is the evaluation, by means of fuel performance codes, of the integral thermal-mechanical behaviour under irradiation, involving the fission gas behaviour and release in the fuel-cladding gap. This work is dedicated to the performance analysis of an inner-core fuel pin representative of the ASTRID sodium-cooled concept design, selected as case study for the benchmark between the GERMINAL and TRANSURANUS fuel performance codes. The focus is on fission gas-related mechanisms and integral outcomes as predicted by means of the SCIANTIX module (allowing the physics-based treatment of inert gas behaviour and release) coupled to both fuel performance codes. The benchmark activity involves the application of both GERMINAL and TRANSURANUS in their “pre-INSPYRE” versions, i.e., adopting the state-of-the-art recommended correlations available in the codes, compared with the “post-INSPYRE” code results, obtained by implementing novel models for MOX fuel properties and phenomena (SCIANTIX included) developed in the framework of the INSPYRE H2020 Project. The SCIANTIX modelling includes the consideration of burst releases of the fission gas stored at the grain boundaries occurring during power transients of shutdown and start-up, whose effect on a fast reactor fuel concept is analysed. A clear need to further extend and validate the SCIANTIX module for application to fast reactor MOX emerges from this work; nevertheless, the GERMINAL-TRANSURANUS benchmark on the ASTRID case study highlights the achieved code capabilities for fast reactor conditions and paves the way towards the proper application of fuel performance codes to safety evaluations on Generation IV reactor concepts.

Besides recent progresses in the physics-based modelling of fission gas and helium behaviour, the scarcity of experimental data concerning their combined behaviour (i.e., cocktail) hinders further model developments. For this reason, in this work, we propose a modelling methodology aimed at providing recommendations for accelerated experimental investigations. By exploring a wide range of annealing temperatures and cocktail compositions with a physics-based modelling approach we identify the most interesting conditions to be targeted by future experiments. To corroborate the recommendations arising from the proposed methodology, we include a sensitivity analysis quantifying the impact of the model parameters on fission gas and helium release, in conditions representative of high and low burnup.

Increasing the average grain size of fuel pellets by doping them with chromium oxide is one strategy to improve oxide nuclear fuels performance. The promoted fission gas retention is thought to improve the performance of the fuel at high burnup. In this work, we review models for the solubility of chromium in UO 2 , and the evolution of the chromium phases in the fuel matrix during irradiation. These models are implemented in SCIANTIX, an open-source mesoscale code describing inert gas behaviour in nuclear fuel. We adjusted the chromium solubility model keeping each parameter within its range of compatibility with experimental data, targeting a better representation of available electron probe microanalysis data of chromium content in fuel after irradiation. As for fission gas behaviour, we considered a physics-based description of the chromium impact on the fission gas diffusivity in fuel grains. The expression for the fission gas diffusivity in standard non-doped uranium oxide has been extended by introducing the impact of the concentration of defects introduced by interstitial oxygen excess representing the effect of chromium content in the fuel itself. A preliminary integral assessment of the proposed models has been carried out against the available experimental data. • Chromium-doped oxide fuels target increased safety and performance. • Chromium-doping increases fuel grain size and promotes effective retention of gases. • Recent experimental findings highlighted the phases arising in chromium-doped fuels. • An engineering model linking phase evolution and fission gas diffusivity is proposed. • The proposed model is tested in SCIANTIX and targets application in engineering tools. [ABSTRACT FROM AUTHOR]

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. • The integral validation database of SCIANTIX is detailed. • The code is coupled with the thermo-mechanical fuel performance codes TRANSURANUS, FRAPCON and OFFBEAT. • The validation database includes Light Water Reactor fuel rods operating in nominal operative and transient conditions. • The comparison is performed against engineering quantities such as the integral fission gas release. • Potential advantages of the multi-scale modelling and future developments are outlined. [ABSTRACT FROM AUTHOR]

• The new version of the SCIANTIX code is described. • The new SCIANTIX modelling capabilities are detailed. • The code structure and its numerical features are presented. • Each model is presented with the separate-effect validation database. • Future model developments and qualification actions are outlined. SCIANTIX is a 0D, open-source code designed to model inert gas behaviour within nuclear fuel at the scale of the grain. The code predominantly employs mechanistic approaches based on kinetic rate-theory models to calculate engineering quantities, such as fission gas release and gaseous fuel swelling. Since its release, SCIANTIX has undergone significant improvements, including the incorporation of new modelling and numerical capabilities. The code architecture has been revamped, embracing an object-orientated structure improving the overall efficiency and usability. This work provides a concise overview of the current state of the SCIANTIX code, highlighting recent updates and advancements. Each SCIANTIX model is presented along with the corresponding separate-effect validation database, which is used to assess its accuracy and predictions. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

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 UO 2 , 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. • The characteristics of the SCIANTIX computer code are described. • The models currently available in SCIANTIX are detailed (with all the parameters). • The verification of the numerical solvers is presented. • Showcases of validation in both constant and transient conditions are presented. • Future (short and long term) development plans are outlined. [ABSTRACT FROM AUTHOR]