• Nuclear Engineering and Technology
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• v.50 no.2
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• pp.259-267
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• 2018

Focused ion beam-scanning electron microscope and electron backscattered diffraction examinations were conducted in the center of a $73\;GWd/t_U\;UO_2$ fuel. They showed the formation of subdomains within the initial grains. The local crystal orientations in these domains were close to that of the original grain. Most of the fission gas bubbles were located on the boundaries. Their shapes were far from spherical and far from lenticular. No interlinked bubble network was found. These observations shed light on previous unexplained observations. They plead for a revision of the classical description of fission gas release mechanisms for the center of high burn-up $UO_2$. Yet, complementary detailed observations are needed to better understand the mechanisms involved.

• Nuclear Engineering and Technology
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• v.28 no.3
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• pp.229-246
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• 1996

A mechanistic model has been developed to predict the release behavior of fission gas during steady-state irradiation of LWR UO$_2$ fuel. Under the assumption that UO$_2$ grain surface is composed of fourteen identical circular faces and grain edge bubble can be represented by a triangulated tube around the circumference of three circular grain faces, it introduces the concept of continuous formation of open grain edges tunnels that is proportional to grain edge swelling. In addition, it takes into account the interaction between the gas release from matrix to grain boundary and the reintroduction of gas atoms into the matrix by the irradiation-induced re-solution of grain face bubbles. It also treats analytically the behavior of intragranular, intergranular, and grain edge bubbles under the assumption that both intragranular and intergranular bubbles are uniform in both radius and number density. Comparison of the present model with experimental data shows that the model's prediction produces reasonable agreement for fuel with centerline temperatures of 1000 to 140$0^{\circ}C$, wide scatter band for fuel with centerline temperatures lower than 100$0^{\circ}C$, and underprediction for fuel with centerline temperatures higher than 140$0^{\circ}C$.

• Nuclear Engineering and Technology
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• v.41 no.2
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• pp.155-162
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• 2009

Since the 90's, EDF and AREVA-NP have irradiated, up to very high burnups, lead assemblies housing $M5^{(R)}$ cladded fuels. Post-irradiation examination of high burnup $UO_2$ pellets show an increase in the fission-gas release rate, an increase in fuel swelling, and formation of fission-gas bubbles throughout the pellets. Xenon abundances were quantified, and phenomena leading to this bubble formation were identified. All examinations provided valuable data on the complex state of the fuel during irradiation. They show the good behavior of these fuels, exhibiting various microstructures at very high burnups, none of which is likely to lead to problems during irradiation.

• Nuclear Engineering and Technology
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• v.26 no.3
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• pp.374-380
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• 1994

A method to simulate the extent of interlinkage of grain boundary gas bubbles to the free surfaces of fuel pellet was developed. With the shape of UO$_2$gain treated as tetrakaidecahedron (TKD)), the interlinked fraction of fission gas bubbles to free surfaces at grain comers was calculated as a function of the radius of grain corner bubbles by the Monte Carlo technique. In spite of two dimensional analysis, the present method shooed reasonable agreement between predicted and measured fuel swelling at the moment that complete bubble interlinkage was achieved. However, for more realistic simulation of interlinkage, grain comer bubbles should be treated three dimensionally.

• Nuclear Engineering and Technology
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• v.46 no.2
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• pp.147-158
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• 2014

In order to investigate how the microstructure of fuel/matrix-interaction (FMI) layers change during irradiation, different U-7Mo dispersion fuel plates have been irradiated to high fission density and then characterized using scanning electron microscopy (SEM). Specifially, samples from irradiated U-7Mo dispersion fuel elements with pure Al, Al-2Si and AA4043 (~4.5 wt.%Si) matrices were SEM characterized using polished samples and samples that were prepared with a focused ion beam (FIB). Features not observable for the polished samples could be captured in SEM images taken of the FIB samples. For the Al matrix sample, a relatively large FMI layer develops, with enrichment of Xe at the FMI layer/Al matrix interface and evidence of debonding. Overall, a significant penetration of Si from the FMI layer into the U-7Mo fuel was observed for samples with Si in the Al matrix, which resulted in a change of the size (larger) and shape (round) of the fission gas bubbles. Additionally, solid fission product phases were observed to nucleate and grow within these bubbles. These changes in the localized regions of the microstructure of the U-7Mo may contribute to changes observed in the macroscopic swelling of fuel plates with Al-Si matrices.

• Nuclear Engineering and Technology
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• v.49 no.4
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• pp.692-699
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• 2017

The determination of the shape and rising velocity of gas bubbles in a liquid pool is of great importance in analyzing the radioactive aerosol emissions from nuclear power plant accidents in terms of the fission product release rate and the pool scrubbing efficiency of radioactive aerosols. This article suggests a simple parameterization for the gas bubble rising velocity as a function of the volume-equivalent bubble diameter; this parameterization does not require prior knowledge of bubble shape. This is more convenient than previously suggested parameterizations because it is given as a single explicit formula. It is also shown that a bubble shape diagram, which is very similar to the Grace's diagram, can be easily generated using the parameterization suggested in this article. Furthermore, the boundaries among the three bubble shape regimes in the $E_o-R_e$ plane and the condition for the bypass of the spheroidal regime can be delineated directly from the parameterization formula. Therefore, the parameterization suggested in this article appears to be useful not only in easily determining the bubble rising velocity (e.g., in postulated severe accident analysis codes) but also in understanding the trend of bubble shape change due to bubble growth.

• Nuclear Engineering and Technology
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• v.54 no.1
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• pp.226-233
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• 2022

In the present work, a phase-field model was developed to investigate the influence of recrystallization on bubble evolution during irradiation. Considering the interaction between bubbles and grain boundary (GB), a set of modified Cahn-Hilliard and Allen-Cahn equations, with field variables and order parameters evolving in space and time, was used in this model. Both the kinetics of recrystallization characterized in experiments and point defects generated during cascade were incorporated in the model. The bubble evolution in recrystallized polycrystalline of U-Mo alloy was also investigated. The simulation results showed that GB with a large area fraction generated by recrystallization accelerates the formation and growth of bubbles. With the formation of new grains, gas atoms are swept and collected by GBs. The simulation results of bubble size and distribution are consistent with the experimental results.

• Nuclear Engineering and Technology
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• v.45 no.7
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• pp.839-846
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• 2013

This paper describes a finite element model of the microstructure of dispersion type nuclear fuels, which can be used to determine the effective thermal conductivity of the fuels during irradiation. The model simulates a representative region of the fuel as a prism shaped unit cell made of brick elements. The elements within the unit cell are assigned material properties of either the fuel or the matrix depending on position, in such a way as to represent randomly distributed fuel particles with a size distribution similar to that of the as manufactured fuel. By applying an appropriate heat flux across the unit cell it is possible to determine the effective thermal conductivity of the unit cell as a function of the volume fraction of the fuel particles. The presence of a fuel/matrix interaction layer is simulated by the addition of a third set of material properties that are assigned to the finite elements that surround each fuel particle. In this way the effective thermal conductivity of the material may also be determined as a function of the volume fraction of the interaction layer. Work is on going to add fission gas bubbles in the fuel as a fourth phase to the model.