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

Giorgi R.,Cechet A.,Cognini L.,Magni A.,Pizzocri D.,Zullo G.,Schubert A.,Van Uffelen P.,Luzzi L. 한국원자력학회 2022 Nuclear Engineering and Technology Vol.54 No.7

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.

Towards grain-scale modelling of the release of radioactive fission gas from oxide fuel. Part I: SCIANTIX

Zullo G.,Pizzocri D.,Magni A.,Van Uffelen P.,Schubert A.,Luzzi L. 한국원자력학회 2022 Nuclear Engineering and Technology Vol.54 No.8

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-ofthe-art semi-empirical approach (ANS 5.4e2010) is reviewed and compared with a mechanistic approach to evaluate the release of radioactive fission gases. At the intra-granular level, the diffusiondecay 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

A new burn-up module for application in fuel performance calculations targeting the helium production rate in (U,Pu)O₂ for fast reactors

Cechet, A.,Altieri, S.,Barani, T.,Cognini, L.,Lorenzi, S.,Magni, A.,Pizzocri, D.,Luzzi, L. Korean Nuclear Society 2021 Nuclear Engineering and Technology Vol.53 No.6

In light of the importance of helium production in influencing the behaviour of fast reactor fuels, in this work we present a burn-up module with the objective to calculate the production of helium in both in-pile and out-of-pile conditions tracking the evolution of 23 alpha-decaying actinides. This burn-up module relies on average microscopic cross-section look-up tables generated via SERPENT high-fidelity calculations and involves the solution of the system of Bateman equations for the selected set of actinide nuclides. The results of the burn-up module are verified in terms of evolution of actinide and helium concentrations by comparing them with the high-fidelity ones from SERPENT, considering two representative test cases of (U,Pu)O₂ fuel in fast reactor conditions. In addition, a code-to-code comparison is made with the independent state-of-the-art module TUBRNP (implemented in the TRANSURANUS fuel performance code) for the same test cases. The herein presented burn-up module is available in the SCIANTIX code, designed for coupling with fuel performance codes.

A surrogate model for the helium production rate in fast reactor MOX fuels

Pizzocri D.,Katsampiris M.G.,Luzzi L.,Magni A.,Zullo G. 한국원자력학회 2023 Nuclear Engineering and Technology Vol.55 No.8

Helium production in the nuclear fuel matrix during irradiation plays a critical role in the design and performance of Gen-IV reactor fuel, as it represents a life-limiting factor for the operation of fuel pins. In this work, a surrogate model for the helium production rate in fast reactor MOX fuels is developed, targeting its inclusion in engineering tools such as fuel performance codes. This surrogate model is based on synthetic datasets obtained via the SCIANTIX burnup module. Such datasets are generated using Latin hypercube sampling to cover the range of input parameters (e.g., fuel initial composition, fission rate density, and irradiation time) and exploiting the low computation requirement of the burnup module itself. The surrogate model is verified against the SCIANTIX burnup module results for helium production with satisfactory performance

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

Magni A.,Pizzocri D.,Luzzi L.,Lainet M.,Michel B. 한국원자력학회 2022 Nuclear Engineering and Technology Vol.54 No.7

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.

Recent Developments in Magnetic Measurements

V. Basso,F. Fiorillo,C. Beatrice,A. Caprile,M. Kuepferling,A. Magni,C. P. Sasso 한국자기학회 2013 Journal of Magnetics Vol.18 No.3

We present a few significant advances in methods and concepts of magnetic measurements, aimed both at providing novel routes in the characterization of hard and soft magnetic materials and at improving our basic knowledge of the magnetization process. We discuss, in particular, investigation methods and experimental arrangements that have been developed in recent times for: 1) Hysteresis loop determination in extra-hard magnets by means of Pulsed Field Magnetometry; 2) Broadband observation of domain wall dynamics by highspeed stroboscopical Kerr techniques; 3) Entropy measurements in magnetocaloric materials by calorimetry in magnetic field. While pertaining to somewhat independent fields of investigation, all these measuring techniques have in common a solid approach to the underlying physical phenomenology and have a potential for further developments.

Towards grain-scale modelling of the release of radioactive fission gas from oxide fuel. Part II: Coupling SCIANTIX with TRANSURANUS

Zullo G.,Pizzocri D.,Magni A.,Van Uffelen P.,Schubert A.,Luzzi L. 한국원자력학회 2022 Nuclear Engineering and Technology Vol.54 No.12

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.

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

L. Luzzi,T. Barani,B. Boer,A. Del Nevo,M. Lainet,S. Lemehov,A. Magni,V. Marelle,B. Michel,D. Pizzocri,A. Schubert,P. Van Uffelen,M. Bertolus 한국원자력학회 2023 Nuclear Engineering and Technology Vol.55 No.3

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 physicsbased 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.

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

Luzzi, L.,Barani, T.,Boer, B.,Cognini, L.,Nevo, A. Del,Lainet, M.,Lemehov, S.,Magni, A.,Marelle, V.,Michel, B.,Pizzocri, D.,Schubert, A.,Uffelen, P. Van,Bertolus, M. Korean Nuclear Society 2021 Nuclear Engineering and Technology Vol.53 No.10

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.