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

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

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

Modeling volatile fission products transport into the Germinal V2 fuel performance code by coupling to thermochemical software

International audienceWithin the PLEIADES simulation platform co-developed by CEA, EDF and FRAMATOME, GERMINAL is the fuel performance code devoted to the in-pile behaviour of mixed oxide fuel pins for Sodium-cooled Fast Reactors. The feedback on such fuel elements, based on experimental observations mainly acquired during PHeNIX french reactor operation, shows that a layer of volatile fission products, called JOG for Joint Oxyde-Gaine, is formed between the fuel pellet and the clad for a burn-up of 6 to 8 %FIMA. At higher burnup, the volatile fission products compounds can react with the cladding material components (Fe, Ni, Cr), corroding the inner wall of the clad.Thus, it is important to improve our chemical modeling of the uranium and plutonium mixed oxide fuel behavior under typical irradiation conditions. For that purpose, the thermochemical code ANGE (close to SOLGASMIX) and, more recently, the OpenCalphad (OC) software, have been integrated into the GERMINAL V2 fuel performance code. The main motivations for choosing OpenCalphad are first its ability to use efficient models and its compatibility with the TAF-ID database.Coupled simulations have been performed on French reactor PHeNIX fuel pins irradiated to different burnup both with ANGE + TBASE using a description based on the Lindemer and Brynestad one, and OpenCalphad + TAF-ID using the description of Gueneau et al. for the modeling of the fuel with its fission products in solution.The calculation results show quite good qualitative agreement with post-irradiation observations regarding the JOG thickness evaluations and the fuel pellet chemical composition. Nevertheless, the simulations are strongly dependent on the volatile fission products release rate as well as on the oxygen content considered in the calculations. Thus, in near future, it will be important to focus on these aspects in order to confirm our evaluations

3d simulation in the pleiades software environment for sodium fast reactor fuel pin behavior under irradiation

International audienceIn the framework of the basic design of ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) the GERMINAL fuel performance code is developed in the PLEIADES software environment. In order to improve one dimensional modelling of GERMINAL, a 3D simulation for the SFR fuel pin behavior under irradiation has been proposed. The 3D model represents a single pellet fragment and its associated piece of cladding. The scale transfer between this single fragment model and the fuel pin scale is achieved through appropriate boundary conditions given by GERMINAL results. The 3D thermo-mechanical computation scheme is implemented in the LICOS code of the PLEIADES platform. In this approach, chemo-physical state variables are pre-computed by the GERMINAL code and are introduced in the 3D computation scheme as some input data in a two-step procedure. First studies have been achieved in order to analyse pellet-to-cladding gap closure mechanisms at the beginning of irradiation. Two mechanisms of fuel relocation have been identified through the 3D simulation. The first one is linked to the hourglass shape of the fragmented fuel pellet under thermal gradient, and the second one is induced by the mass transfer due the central-hole formation and fuel restructuration. Thanks to our 3D results, the gap closure rate given by the fuel relocation displacement model of GERMINAL can be interpreted. The next step is now to propose a full coupling formulation between fuel mass transfer model and the radial relocation displacement model

Results of the applicative benchmark between TRANSURANUS and GERMINAL on the ASTRID case study

This deliverable addresses one of the benchmark activities foreseen in Task 7.3 of INSPYRE, which focuses on the simulation of irradiation case studies representative of the ESNII reactor prototypes. The goal of this work is to assess the predictive design capabilities of fuel performance codes before and after the improvements brought about by the INSPYRE Project. The ASTRID case study, which was selected by the Task Force on fuel performance codes taking into account the needs expressed by the User Group, is a base irradiation for an axially heterogeneous fuel pin representative of the maximal loading condition in the ASTRID reactor. This case study was analysed by POLIMI using the TRANSURANUS fuel performance code and by CEA using the GERMINAL code

GERMINAL, a fuel performance code of the PLEIADES platform to simulate the in-pile behaviour of mixed oxide fuel pins for sodium-cooled fast reactors

International audienceGERMINAL is a fuel performance code developed by the French Commission of Alternative and Atomic Energies (CEA) to simulate the in-pile behaviour of mixed oxide fuel pins for Sodium-cooled Fast Reactors. The code is continuously being improved.GERMINAL was initially designed to simulate the fuel pin behaviour of the PHENIX and SUPER-PHENIX reactors, which were built in France and have been in operation over the last decades. The GERMINAL models were then extended and improved to meet the needs of the design studies of ASTRID, a project of a technological Sodium Fast Reactor demonstrator in France. The goal of this article is to introduce the current modelling implemented in GERMINAL. The code is validated and the validation work is illustrated here by a selection of comparisons between calculations and measurements. Working perspectives for further modelling improvements are finally presented, through more mechanistic approaches sustained by three-dimensional computations or based on extended physical couplings

Current status and progression of germinal fuel performance code for sfr oxide fuel pins

International audienceA fuel performance code for SFR oxide fuel pins, GERMINAL, is developed by CEA within the PLEIADES simulation framework. The present main goal of GERMINAL is to meet the needs of the design studies of ASTRID, the future Advanced Sodium Technological Reactor for Industrial Demonstration in France. Recent works have been conducted to improve the modelling of different physical mechanisms having a strong influence on the design criteria evaluation. Thus, the formulation of the fuel pellet fragments relocation model has been revisited, by introducing a dependence to the thermal gradient inside the pellet. The description of this mechanism represents a key point to evaluate the pellet-to-cladding gap closure and the margin to melting at beginning of life. Another evolution concerns the pellet-clad mechanical interaction. The ability to simulate a stronger interaction for fuel pins with a higher filling fraction has been acquired with a focused work on fuel mechanical behavior. A stronger mechanical interaction may also happen with lower power operating conditions and a cladding material remaining stable under irradiation. Moreover, the description of the thermochemistry of oxide fuel is currently being improved by coupling GERMINAL with the OpenCalphad thermodynamic calculation software. In doing this, the goal is to obtain a better prediction of the amount of volatile fission products being transported outside the fuel pellet, and then contributing to the Joint Oxyde-Gaine formation. With refined estimations of JOG volume and composition, we expect further to improve the evaluation of heat transfer through pellet-to-cladding gap at high burn-up, and also a more mechanistic description of cladding corrosion due to released fission products. These works are based on a systematic comparison of calculation results to post-irradiation measures, by integrating progressively additional objects to our validation base. This process leads to a wider validity range targeting ASTRID design, and brings out new working perspectives