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

Description of new meso-scale models and their implementation in fuel performance codes

This deliverable illustrates the new (2.0) version of the SCIANTIX meso-scale code, developed within Task 5.2 of the PATRICIA Project, highlighting first the code structure and its numerical features. Then, the SCIANTIX models for various physics involved in the inert gas behaviour are described in detail along with their corresponding separate-effect validation database and validation results. The coupling of SCIANTIX with integral, pin-level fuel performance codes is also introduced, presenting the different strategy and interface details for the coupling with the TRANSURANUS and GERMINAL fuel performance codes. Finally, conclusions and future perspectives are provided, mentioning several envisaged developments targeted in the framework of multiple research initiatives at a European and international level, and outlining the strategy foreseen for further developments of the code (in both its stand-alone and coupled fashion)

Fuel performance simulations of ESNII prototypes: Results on the MYRRHA case study

Nominal and transient conditions of the ESNII prototypes were investigated in the INSPYRE Project using the European fuel performance codes GERMINAL, MACROS and TRANSURANUS. This Deliverable presents the results of the simulations of the MYRRHA case study: MYRRHA nominal irradiation conditions and the occurrence of a beam power jump (over‐power) transient at the beginning and end of life of the fuel pin in reactor. Besides the application of the reference (“pre‐INSPYRE”) code versions, the activity involves the evaluation of the impact of the improved models of MOX fuel properties developed in INSPYRE and implemented in the three fuel performance codes. These modelling advances concern the thermal properties (thermal conductivity, melting temperature), mechanical properties (thermal expansion, Young’s modulus) and the mechanistic treatment of fission gas behaviour and release from MOX fuels. The results yielded by the pre‐INSPYRE and post‐INSPYRE versions of the codes involved are presented and assessed in terms of evolution in time, as well as axial and radial profiles of significant quantities, both integral and local. Then, the code results are compared with the design limits set for the MYRRHA fuel pins, in particular the maximal fuel temperature admitted, which prevents fuel melting, and the maximal allowed cladding plasticity that ensures the cladding integrity. The outcome is a complete compliance of the pin behaviour with the design limits, respecting adequate margins even in the case of the hottest fuel pin and in the case of beam power jump transients

Assessment of INSPYRE-extended fuel performance codes against the SUPERFACT-1 fast reactor irradiation experiment

Design and safety assessment of fuel pins for application in innovative Generation IV fast reactors calls for a dedicated nuclear fuel modelling and for the extension of the fuel performance code capabilities to the envisaged materials and irradiation conditions. In the INSPYRE Project, comprehensive and physics- based models for the thermal-mechanical properties of UePu mixed-oxide (MOX) fuels and for fission gas behaviour were developed and implemented in the European fuel performance codes GERMINAL, MACROS and TRANSURANUS. As a follow-up to the assessment of the reference code versions (“pre- INSPYRE”, NET 53 (2021) 3367e3378), this work presents the integral validation and benchmark of the code versions extended in INSPYRE (“post-INSPYRE”) against two pins from the SUPERFACT-1 fast reactor irradiation experiment. The post-INSPYRE simulation results are compared to the available integral and local data from post-irradiation examinations, and benchmarked on the evolution during irradiation of quantities of engineering interest (e.g., fuel central temperature, fission gas release). The comparison with the pre-INSPYRE results is reported to evaluate the impact of the novel models on the predicted pin performance. The outcome represents a step forward towards the description of fuel behaviour in fast reactor irradiation conditions, and allows the identification of the main remaining gaps

Helium solubility in oxide nuclear fuel: Derivation of new correlations for Henry’s constant

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, kH (at m-3 MPa-1 ): kH = 1. 8·10^25·exp(-0.41/kT) for powders and kH = 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

Helium diffusivity in oxide nuclear fuel: Critical data analysis and new correlations

Helium is relevant in determining nuclear fuel behaviour. It affects the performance of nuclear fuel both in reactor and in storage conditions. Helium becomes important in reactor conditions when high burnups are targeted or MOX fuel is used, whereas for storage conditions it can represent a threat to the fuel rods integrity. The accurate knowledge of helium behaviour combined with predictive model capabilities is fundamental for the safe management of nuclear fuel, with helium diffusivity being a critical property. For this reason, a considerable number of separate effect experiments in the last fifty years investigated helium diffusivity in nuclear fuel. The aim of this work is to critically review and assess the experimental results concerning the helium diffusivity. Experimental results are critically analysed in terms of the helium introduction technique used (either infusion, implantation or doping) and of sample characteristics (single crystal, poly-crystal or powder). Accordingly, we derived two different correlations for the diffusivity. Clearly, each of the new correlations corresponds to a limited range of application conditions, depending on the experimental data used to derive it. We provide recommendations regarding the proper application conditions for each correlation (e.g., in reactor or storage conditions)

Results of the benchmark between pre- and post-INSPYRE code versions on selected experimental cases

This report presents the results of the simulation of the SUPERFACT-1, RAPSODIE-I and NESTOR-3 irradiation experiments using the fuel performance codes TRANSURANUS, MACROS, GERMINAL. The simulations aim at the evaluation of the code improvements made during the INSPYRE project. The comparison of the integral pin performance results with experimental measurements available from the irradiation experiments considered and the comparison between the code results are presented. Both the results obtained using the ‘pre-INSPYRE’ code versions and the improved ‘post-INSPYRE’ ones, in which novel data and models originating from other Work Packages of the INSPYRE Project were implemented, are provided