Modelling of Stable Fission Gases Diffusion in UO2 Nuclear Fuel

2008 ◽  
Vol 273-276 ◽  
pp. 693-698 ◽  
Author(s):  
Isabel Vallejo ◽  
Luisen E. Herranz
Keyword(s):  
Nuclear Fuel ◽  
Current Trend ◽  
Nuclear Industry ◽  
Temperature Threshold ◽  
Empirical Correlations ◽  
Modified Model ◽  
Fission Gas ◽  
Specific Temperature ◽  
Void Volumes ◽  
Fission Gas Release

The current trend in nuclear industry to extend the burnup limits up to high burnup requires the analysis of the phenomenology related to the nuclear fuel, specifically the diffusion and fission gas release (FGR) related to this phenomenon plays an important role. In this paper, the diffusion of stable fission gases (xenon and krypton) in nuclear fuel UO2 is addressed through the predictions for FGR to rod void volumes obtained with FRAPCON-3 fuel performance code. The theoretical base of diffusion model in FRAPCON-3 code is shown and some modifications are proposed, such as the removing of empirical correlations not related to diffusion phenomenon and fitting parameters included in the model. Besides, the resolution process in the proximities of grain boundaries is considered in a different way, and the grain growth mechanism from a specific temperature threshold is implemented into the code. The database applied for evaluation is presented and the results with the original and modified model are shown.

Axial Fission Gas Transport in Nuclear Fuel Rods

2009 ◽  
Vol 283-286 ◽  
pp. 262-267
Author(s):  
M.T. del Barrio ◽  
Luisen E. Herranz
Keyword(s):  
Nuclear Fuel ◽  
Gas Transport ◽  
Fuel Pellet ◽  
Gas Release ◽  
Gas Mixing ◽  
Fuel Performance ◽  
Fuel Rods ◽  
Fuel Rod ◽  
Fission Gas ◽  
Fission Gas Release

Fission of fissile uranium or plutonium nucleus in nuclear fuel results in fission products. A small fraction of them are volatile and can migrate under the effect of concentration gradients to the grain boundaries of the fuel pellet. Eventually, some fission gases are released to the rod void volumes by a thermally activated process. Local transients of power generation could distort even further the already non-uniform axial power and fission gas concentration profiles in fuel rods. Most of the current fuel rod performance codes neglects these gradients and the resulting axial fission gas transport (i.e., gas mixing is considered instantaneous). Experimental evidences, however, highlight axial gas mixing as a real time-dependent process. The thermal feedback between fission gas release, gap composition and fuel temperature, make the “prompt mixing assumption” in fuel performance codes a key point to investigate due to its potential safety implications. This paper discusses the possible scenarios where axial transport can become significant. Once the scenarios are well characterized, the available database is explored and the reported models are reviewed to highlight their major advantages and shortcomings. The convection-diffusion approach is adopted to simulate the axial transport by decoupling both motion mechanisms (i.e., convection transport assumed to be instantaneous) and a stand-alone code has been developed. By using this code together with FRAPCON-3, a prospective calculation of the potential impact of axial mixing is conducted. The results show that under specific but feasible conditions, the assumption of “prompt axial mixing” could result in temperature underestimates for long periods of time. Given the coupling between fuel rod thermal state and fission gas release to the gap, fuel performance codes predictions could deviate non-conservatively. This work is framed within the CSN-CIEMAT agreement on “Thermo-Mechanical Behaviour of the Nuclear Fuel at High Burnup”.

Fuel Performance Modelling At High Burn-Up by FEMAXI-6 Code

2021 ◽  
Author(s):  
Jakub Luley ◽  
Branislav Vrban ◽  
Stefan Cerba ◽  
Filip Osuský ◽  
Vladimir Necas
Keyword(s):  
Pressure Sensors ◽  
Nuclear Industry ◽  
Performance Limits ◽  
Performance Modelling ◽  
Fuel Performance ◽  
Fuel Burn ◽  
Power Cycling ◽  
Current Requirement ◽  
Fission Gas ◽  
Fission Gas Release

Abstract The scope of current research in the field of fuel performance is primary aimed to an improvement of the operating reliability, safety and cost effectiveness of the reactors in operation. The current requirement of nuclear industry is to have fuel suitable for load follow operation. Fission gas release, Pellet-Cladding Mechanical Interaction and stress corrosion cracking are the main phenomena that limit the variability of reactor operation from a safety perspective. To reasonable predict the fuel performance limits it is necessary to benchmark the computational tools against high quality experimental data. This work is devoted to the calculation of fuel performance using the code FEMAXI-6 based on the longest irradiation experiment in the Halden reactor. The fuel burn-up was approaching 90 MWd/kgUO2 in three selected rods which were equipped by the pressure sensors and were subjected to extensive post-irradiation examination. During the experiment, the rods were exposed to several periods of power cycling. The rods were manufactured with different fuel grain size and fuel-to-clad gap size.

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