Integration of OpenCalphad thermo-chemical solver in PLEIADES/ALCYONE fuel performance code

International audienceThe ALCYONE fuel performance code, co-developed by CEA, EDF and Framatome, within the PLEIADES software environment provides a multidimensional modeling for detailed analysis of PWR fuel elements behavior under irradiation [1]. Iodine-Stress Corrosion Cracking is one of the physical phenomena of major interest for cladding design and long term operation of PWRs. In a first step towards I-SCC simulations, the thermochemical code ANGE was integrated in PLEIADES [2]. ANGE, a modified version of SOLGASMIX, enables to compute thermo-chemical equilibria using the TBASE database [3] and associate species description [4] but has some limitations and cannot be used to solve chemical systems based on the Compound Energy Formalism, such as the one proposed in the TAF-ID [5]. Consequently, a robust, efficient and free numerical tool, OpenCalphad [6], was introduced in PLEIADES. In this work, we focus our presentation on the calculation of complex multi-component systems representative of fuel elements behavior under irradiation. From the results of in-reactor power transient calculations (1D-2D-3D), we show that ALCYONE/OpenCalphad is much faster than ALCYONE/ANGE. We note a decrease of the CPU time by almost a factor 4 that can be explained by the OpenCalphad solver itself and by a set of numerical strategies implemented to start a thermodynamic calculation on a mesh node by using another calculated equilibrium as an initial solution. We also show through first results the capacity and the robustness of the ALCYONE/OpenCalphad coupling to do in-reactor power transients calculations (1D-2D-3D) using the TAF-ID. In the latter, the models are more complicated and the possible phases are greater in number than in the TBASE database. For calculations performed in the same conditions as those done with the TBASE database, we note a slight increase of the CPU time that can be reduced by calculating several thermodynamic equilibria simultaneously with a multithread approach.References[1] V. Marelle, et al. New developments in ALCYONE 2.0 fuel performance code, Top Fuel, Boise ID (2016)[2] B. Baurens, et al., J. Nucl. Mater. 452 (2014) 578[3] E.H.P. Cordfuncke, R.J.M. Konings, J. Phase Equilibria 14 [4] (1993)[4] T.M. Besmann, Comprehensive Nucl. Mater. 1.17 (2012)[5] C. Gueneau et al., J. Nucl. Mater. 419 (2011) 147[6] B. Sundman, et al, Integ. Mater. Manuf. Innov. 4 (2015)

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

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

Clean air policies are key for successfully mitigating Arctic warming

A tighter integration of modeling frameworks for climate and air quality is urgently needed to assess the impacts of clean air policies on future Arctic and global climate. We combined a new model emulator and comprehensive emissions scenarios for air pollutants and greenhouse gases to assess climate and human health co-benefits of emissions reductions. Fossil fuel use is projected to rapidly decline in an increasingly sustainable world, resulting in far-reaching air quality benefits. Despite human health benefits, reductions in sulfur emissions in a more sustainable world could enhance Arctic warming by 0.8 °C in 2050 relative to the 1995–2014, thereby offsetting climate benefits of greenhouse gas reductions. Targeted and technically feasible emissions reduction opportunities exist for achieving simultaneous climate and human health co-benefits. It would be particularly beneficial to unlock a newly identified mitigation potential for carbon particulate matter, yielding Arctic climate benefits equivalent to those from carbon dioxide reductions by 2050

Fuel Performance Codes

International audienc

Integration of OpenCalphad thermo-chemical solver in PLEIADES/ALCYONE fuel performance code

International audienceThe ALCYONE fuel performance code, co-developed by CEA, EDF and Framatome, within the PLEIADES software environment provides a multidimensional modeling for detailed analysis of PWR fuel elements behavior under irradiation [1]. Iodine-Stress Corrosion Cracking is one of the physical phenomena of major interest for cladding design and long term operation of PWRs. In a first step towards I-SCC simulations, the thermochemical code ANGE was integrated in PLEIADES [2]. ANGE, a modified version of SOLGASMIX, enables to compute thermo-chemical equilibria using the TBASE database [3] and associate species description [4] but has some limitations and cannot be used to solve chemical systems based on the Compound Energy Formalism, such as the one proposed in the TAF-ID [5]. Consequently, a robust, efficient and free numerical tool, OpenCalphad [6], was introduced in PLEIADES. In this work, we focus our presentation on the calculation of complex multi-component systems representative of fuel elements behavior under irradiation. From the results of in-reactor power transient calculations (1D-2D-3D), we show that ALCYONE/OpenCalphad is much faster than ALCYONE/ANGE. We note a decrease of the CPU time by almost a factor 4 that can be explained by the OpenCalphad solver itself and by a set of numerical strategies implemented to start a thermodynamic calculation on a mesh node by using another calculated equilibrium as an initial solution. We also show through first results the capacity and the robustness of the ALCYONE/OpenCalphad coupling to do in-reactor power transients calculations (1D-2D-3D) using the TAF-ID. In the latter, the models are more complicated and the possible phases are greater in number than in the TBASE database. For calculations performed in the same conditions as those done with the TBASE database, we note a slight increase of the CPU time that can be reduced by calculating several thermodynamic equilibria simultaneously with a multithread approach.References[1] V. Marelle, et al. New developments in ALCYONE 2.0 fuel performance code, Top Fuel, Boise ID (2016)[2] B. Baurens, et al., J. Nucl. Mater. 452 (2014) 578[3] E.H.P. Cordfuncke, R.J.M. Konings, J. Phase Equilibria 14 [4] (1993)[4] T.M. Besmann, Comprehensive Nucl. Mater. 1.17 (2012)[5] C. Gueneau et al., J. Nucl. Mater. 419 (2011) 147[6] B. Sundman, et al, Integ. Mater. Manuf. Innov. 4 (2015)

Simulation of Pellet-Cladding interaction with the PLEIADES fuel performance software environment

International audienceThis paper focuses on the PLEIADES fuel performance software environment and its application to the modeling of pellet-cladding interaction (PCI). The PLEIADES platform has been under development for 10 yr; a unified software environment, including the multidimensional finite element solver CAST3M, has been used to develop eight computation schemes now under operation. Among the latter, the ALCYONE application is devoted to pressurized water reactor fuel rod behavior. This application provides a three-dimensional (3-D) model for a detailed analysis of fuel element behavior and enables validation through comparing simulation and postirradiation examination results (cladding residual diameter and ridges, dishing filling, pellet cracking, etc.). These last years the 3-D computation scheme of the ALCYONE application has been enriched with a complete set of physical models to take into account thermomechanical and chemical-physical behavior of the fuel element under irradiation. These models have been validated through the ALCYONE application on a large experimental database composed of approximately 400 study cases. The strong point of the ALCYONE application concerns the local approach of stress-corrosion-cracking rupture under PCI, which can be computed with the 3-D finite element solver. Further developments for PCI modeling in the PLEIADES platform are devoted to a new mesh refinement method for assessing stress-and-strain concentration (multigrid technique) and a new component for assessing fission product chemical recombination

Integration of OpenCalphad thermo-chemical solver in PLEIADES/ALCYONE fuel performance code

International audienc

Assessment of three European fuel performance codes against the SUPERFACT-1 fast reactor irradiation experiment

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