MULTISCALE MODELLING FOR THE FISSION GAS BEHAVIOUR IN THE TRANSURANUS CODE

P. VAN UFFELEN,G. PASTORE,V. DI MARCELLO,L. LUZZI 한국원자력학회 2011 Nuclear Engineering and Technology Vol.43 No.6

A formulation is proposed for modelling the process of intra-granular diffusion of fission gas during irradiation of UO2under both normal operating conditions and power transients. The concept represents a simple extension of the formulation of Speight, including an estimation of the contribution of bubble motion to fission gas diffusion. The resulting equation is formally identical to the diffusion equation adopted in most models that are based on the formulation of Speight, therefore retaining the advantages in terms of simplicity of the mathematical-numerical treatment and allowing application in integral fuel performance codes. The development of the new model proposed here relies on results obtained by means of molecular dynamics simulations as well as finite element computations. The formulation is proposed for incorporation in the TRANSURANUS fuel performance code.

MULTISCALE MODELLING FOR THE FISSION GAS BEHAVIOUR IN THE TRANSURANUS CODE

Van Uffelen, P.,Pastore, G.,Di Marcello, V.,Luzzi, L. Korean Nuclear Society 2011 Nuclear Engineering and Technology Vol.43 No.6

A formulation is proposed for modelling the process of intra-granular diffusion of fission gas during irradiation of $UO_2$ under both normal operating conditions and power transients. The concept represents a simple extension of the formulation of Speight, including an estimation of the contribution of bubble motion to fission gas diffusion. The resulting equation is formally identical to the diffusion equation adopted in most models that are based on the formulation of Speight, therefore retaining the advantages in terms of simplicity of the mathematical-numerical treatment and allowing application in integral fuel performance codes. The development of the new model proposed here relies on results obtained by means of molecular dynamics simulations as well as finite element computations. The formulation is proposed for incorporation in the TRANSURANUS fuel performance code.

Investigation on the effect of eccentricity for fuel disc irradiation tests

Scolaro, A.,Van Uffelen, P.,Fiorina, C.,Schubert, A.,Clifford, I.,Pautz, A. Korean Nuclear Society 2021 Nuclear Engineering and Technology Vol.53 No.5

A varying degree of eccentricity always exists in the initial configuration of a nuclear fuel rod. Its impact on traditional LWR fuel is limited as the radial gap closes relatively early during irradiation. However, the effect of misalignment is expected to be more relevant in rods with highly conductive fuels, large initial gaps and low conductivity filling gases. In this paper, we study similar characteristics in the experimental setup of two fuel disc irradiation campaigns carried out in the OECD Halden Boiling Water Reactor. Using the multi-dimensional fuel performance code OFFBEAT, we combine 2-D axisymmetric and 3-D simulations to investigate the effect of eccentricity on the fuel temperature distribution. At the same time, we illustrate how the advent of modern tools with multi-dimensional capabilities might further improve the design and interpretation of in-pile separate-effect tests and we outline the potential of such an analysis for upcoming experiments.

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.

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

Towards a physics-based description of intra-granular helium behaviour in oxide fuel for application in fuel performance codes

Cognini, L.,Cechet, A.,Barani, T.,Pizzocri, D.,Van Uffelen, P.,Luzzi, L. Korean Nuclear Society 2021 Nuclear Engineering and Technology Vol.53 No.2

In this work, we propose a new mechanistic model for the treatment of helium behaviour which includes the description of helium solubility in oxide fuel. The proposed model has been implemented in SCIANTIX and validated against annealing helium release experiments performed on small doped fuel samples. The overall agreement of the new model with the experimental data is satisfactory, and given the mechanistic formulation of the proposed model, it can be continuously and easily improved by directly including additional phenomena as related experimental data become available.

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.

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.

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.