The SCIANTIX code for fission gas behaviour: Status, upgrades, separate-effect validation, and future developments

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

Synthesis of the presentation of INSPYRE results to the User Group

INSPYRE created a user group composed of key customers for the project’s results, which included representatives of the designers of the ESNII reactor concepts (ASTRID, MYRRHA, ALFRED, ALLEGRO), as well as of fuel manufacturers and utilities (ORANO, EDF). A first meeting had been organized in August 2018 to present the INSPYRE approach and activities to the users and discuss their needs in the area covered by INSPYRE activities. The synthesis of the meeting is reported in INSPYRE Deliverable D9.4. Three meetings with the user group were then organised throughout the project to present the approach and results of the project to the Users, get their feedback on these results and discuss follow-up activities. - The Second User Group meeting took place on January 17th, 2020 and was dedicated to the developments made in the fuel performance codes considered in the project and their assessment against the selected fast reactor irradiation experiments - The third User Group meeting was held on May 31st, 2021 and was dedicated to the simulation of the fuel elements in normal operating conditions of the ASTRID reactor concept - The final User Group meeting was organised jointly with the second Scientific Advisory Committee meeting on May 24th, 2022 and concerned the overall scientific outcomes of INSPYRE. The present deliverable reports the presentations made during the meetings and the synthesis of the exchanges that followed

Application of the SCIANTIX fission gas behaviour module to the integral pin performance in sodium fast reactor irradiation conditions

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

Synthesis of INSPYRE results

This document presents a synthesis of the results of the INSPYRE H2020 European Project, which was dedicated to the investigation and simulation of uranium-plutonium mixed oxide (MOX) fuels for fast reactors and their behaviour under irradiation. It first describes the progress in the understanding and description of the behaviour of MOX fuels obtained thanks to the combination of basic research (multiscale modelling and separate effect experiments) and examination of fuels irradiated in reactors in past experiments. Four operational issues were particularly studied: margin to fuel melting; irradiated fuel thermochemistry and interaction with the cladding; self-diffusion properties and inert gas behaviour; evolution of mechanical properties under irradiation. Second, the advances made in the simulation of fast reactor MOX fuels thanks to the development of more physically justified models and their implementations in three European fuel performance codes are presented. Then, the results of the simulation of two ESNII reactor cores are shown: normal operation conditions in the ASTRID sodium fast reactor prototype and normal and transient conditions in the lead-bismuth cooled reactor of the MYRRHA accelerator driven system. Finally, the activities conducted concerning education and training, dissemination, exploitation and communication are summarized

Spaceship Earth. Space-driven technologies and systems for sustainability on ground

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

Best practices and QA protocols for code development

The OperaHPC project aims to improve the numerical capabilities of 3D fuel performance modelling as part of its strategic objectives. To achieve this goal, an open-source approach has been chosen for the tools developed in the framework of the project, namely MMM and OFFBEAT, the latter coupled to the SCIANTIX code. As the open-source approach is relatively new in the domain of nuclear safety studies, this document presents a framework for achieving quality assurance targets for the open-source scientific computing tools within the OperaHPC project. First, the document provides a brief review of the most common QA programs and standards employed in the field, with a particular focus to the aspects that are more relevant to OperaHPC. Then, it discusses modern software development practices to improve code quality, highlighting the importance of revision control systems, testing methodologies, and documentation. Finally, it describes the concept of governance model for regulating interactions between contributors, users, and decision-makers. The framework presented in this document provides a backbone for the verification and validation actions that will be carried out within the project and contributes to the qualification of the MMM, OFFBEAT and SCIANTIX tools for nuclear safety studies

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

Towards simulations of fuel rod behaviour during severe accidents by coupling TRANSURANUS with SCIANTIX and MFPR-F

Among the applications of the multiscale modelling approach in nuclear fuel rod performance, the coupling of integral thermo-mechanical fuel performance codes with lower-length meso-scale modules is of great interest. This strategy allows to overcome correlation-based approaches with mechanistic ones and test their application in accidental conditions. In this work, we explore the coupling between the TRANSURANUS fuel performance code and two meso-scale modules for fission gas/product behaviour: MFPR-F and SCIANTIX. These modules, coupled within TRANSURANUS, are assessed against the IFA-650.10 loss-of-coolant accident test to analyse their overall impact and highlight future developments toward mechanistic modelling of fission gas during accident scenarios

Physics-based modelling and validation of inter-granular helium behaviour in SCIANTIX

In this work, we propose a new mechanistic model for the treatment of helium behaviour at the grain boundaries in oxide nuclear fuel. The model provides a rate-theory description of helium inter-granular behaviour, considering diffusion towards grain edges, trapping in lenticular bubbles, and thermal resolution. It is paired with a rate-theory description of helium intra-granular behaviour that includes diffusion towards grain boundaries, trapping in spherical bubbles, and thermal re-solution. The proposed model has been implemented in the meso-scale software designed for coupling with fuel performance codes SCIANTIX. It is validated against thermal desorption experiments performed on doped UO2 samples annealed at different temperatures. The overall agreement of the new model with the experimental data is improved, both in terms of integral helium release and of the helium release rate. By considering the contribution of helium at the grain boundaries in the new model, it is possible to represent the kinetics of helium release rate at high temperature. Given the uncertainties involved in the initial conditions for the inter-granular part of the model and the uncertainties associated to some model parameters for which limited lower-length scale information is available, such as the helium diffusivity at the grain boundaries, the results are complemented by a dedicated uncertainty analysis. This assessment demonstrates that the initial conditions, chosen in a reasonable range, have limited impact on the results, and confirms that it is possible to achieve satisfying results using sound values for the uncertain physical parameters