• Nuclear Engineering and Technology
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• v.46 no.2
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• pp.159-168
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• 2014

In order to advance understanding of the breakaway swelling behavior of U-Mo/Al dispersion fuel under a high-power irradiation condition, the effects of fuel-matrix interaction on the fuel performance of U-Mo/Al dispersion fuel were investigated. Fission gas release into large interfacial pores between interaction layers and the Al matrix was analyzed using both mechanistic models and observations of the post-irradiation examination results of U-Mo dispersion fuels. Using the model predictions, advantageous fuel design parameters are recommended to prevent breakaway swelling.

• Nuclear Engineering and Technology
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• v.40 no.5
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• pp.409-418
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• 2008

Because the interaction layers that form between U-Mo particles and the Al matrix degrade the thermal properties of U-Mo/Al dispersion fuel, an investigation was undertaken of the undesirable feedback effect between an interaction layer growth and a centerline temperature increase for dispersion fuel. The radial temperature distribution due to interaction layer growth during irradiation was calculated iteratively in relation to changes in the volume fractions, the thermal conductivities of the constituents, and the oxide thickness with the burnup. The interaction layer growth, which is estimated on the basis of the temperature calculations, showed a reasonable agreement with the post-irradiation examination results of the U-Mo/Al dispersion fuel rods irradiated at the HANARO reactor. The U-Mo particle size was found to be a dominant factor that determined the fuel temperature during irradiation. Dispersion fuel with larger U-Mo particles revealed lower levels of both the interaction layer formation and the fuel temperature increase. The results confirm that the use of large U-Mo particles appears to be an effective way of mitigating the thermal degradation of U-Mo/Al dispersion fuel.

• 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.

• Nuclear Engineering and Technology
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• v.45 no.7
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• pp.827-838
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• 2013

Interaction layer growth between U-Mo alloy fuel particles and Al in a dispersion fuel is a concern due to the volume expansion and other unfavorable irradiation behavior of the interaction product. To reduce interaction layer (IL) growth, a small amount of Si is added to the Al. As a result, IL growth is affected by the Si content in the Al matrix. In order to predict IL growth during fabrication and irradiation, empirical models were developed. For IL growth prediction during fabrication and any follow-on heating process before irradiation, out-of-pile heating test data were used to develop kinetic correlations. Two out-of-pile correlations, one for the pure Al matrix and the other for the Al matrix with Si addition, respectively, were developed, which are Arrhenius equations that include temperature and time. For IL growth predictions during irradiation, the out-of-pile correlations were modified to include a fission-rate term to consider fission enhanced diffusion, and multiplication factors to incorporate the Si addition effect and the effect of the Mo content. The in-pile correlation is applicable for a pure Al matrix and an Al matrix with the Si content up to 8 wt%, for fuel temperatures up to $200^{\circ}C$, and for Mo content in the range of 6 - 10wt%. In order to cover these ranges, in-pile data were included in modeling from various tests, such as the US RERTR-4, -5, -6, -7 and -9 tests and Korea's KOMO-4 test, that were designed to systematically examine the effects of the fission rate, temperature, Si content in Al matrix, and Mo content in U-Mo particles. A model converting the IL thickness to the IL volume fraction in the meat was also developed.

• 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.45 no.7
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• pp.859-870
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• 2013

The paper summarizes the irradiation test and post-irradiation examination (PIE) data for the U-Mo low-enriched fuel that was irradiated in the MIR reactor under the RERTR Program. The PIE data were analyzed for both full-size fuel rods and mini-rods with atomized powder dispersed in Al matrix as well as with additions of 2%, 5% and 13% of silicon in the matrix and ZrN protective coating on the fuel particles. The full-size fuel rods were irradiated up to an average burnup of ${\sim}60%^{235}U$; the mini-rods were irradiated to an average burnup of ${\sim}85%^{235}U$. The presented data show a significant increase of the void fraction in the U-Mo alloy as the U-235 burnup rises from ~ 40% up to ~ 85%. The effect of irradiation test conditions and U-235 burnup were analyzed with regard to the formation of an interaction layer between the matrix and fuel particles as well as generation of porosity in the U-Mo alloy. Shown here are changes in distribution of U fission products as the U-235 burnup increases from ~ 40% up to ~ 85%.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.27 no.10
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• pp.1420-1426
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• 2003

The stability of turbulent nonpremixed interacting flames is investigated in terms of nozzle configuration shapes and kind of fuels. Four nozzle arrangements - cross 5, matrix 8, matrix 9 and circle 8 nozzles - are used in the experiment. There are many parameters affecting flame stability in multi-nozzle flames such as nozzle separation distance, fuel flowrates and nozzle configuration etc. Key factors to enhance blowout limit are the nozzle configuration and the existence of center nozzle. Even nozzle exit velocity equal 204 m/s, flame is not extinguished when there is not a center nozzle and s/d=15.3∼27.6 in matrix-8 and circular-8 configurations. At these conditions, recirculation of burnt gas is related with stability augmentation. Fuel mole fraction measurements using laser induced fluorescence reveal lifted flame base is not located at the stoichiometric contour.

• Nuclear Engineering and Technology
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• v.45 no.7
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• pp.847-858
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• 2013

Since 2001, a series of five irradiation test campaigns for atomized U-Mo dispersion fuel rods, KOMO-1, -2, -3, -4, and -5, has been conducted at HANARO (Korea) in order to develop high performance low enriched uranium dispersion fuel for research reactors. The KOMO irradiation tests provided valuable information on the irradiation behavior of U-Mo fuel that results from the distinct fuel design and irradiation conditions of the rod fuel for HANARO. Full size U-Mo dispersion fuel rods of 4-5 $g-U/cm^3$ were irradiated at a maximum linear power of approximately 105 kW/m up to 85% of the initial U-235 depletion burnup without breakaway swelling or fuel cladding failure. Electron probe microanalyses of the irradiated samples showed localized distribution of the silicon that was added in the matrix during fuel fabrication and confirmed its beneficial effect on interaction layer growth during irradiation. The modifications of U-Mo fuel particles by the addition of a ternary alloying element (Ti or Zr), additional protective coatings (silicide or nitride), and the use of larger fuel particles resulted in significantly reduced interaction layers between fuel particles and Al.

• 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.46 no.2
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• pp.125-146
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• 2014

This review aims to provide a synthesis of the knowledge generated and the lessons learned in roughly 15 years of UMo dispersion fuel R&D in Europe through a series of irradiation experiments. A lot of irradiations were also performed outside of Europe, particularly in the USA, Russia, Canada, Korea and Argentina. In addition, a large number of out-of-pile investigations were done throughout the world, providing support to the understanding of the phenomena governing the UMo behaviour in pile. However, the focus of this article will be on the irradiations and Post-Irradiation Examination (PIE) results obtained in European experiments. The introduction of the article provides a historic overview of the evolution and progress in the high density UMo dispersion fuel development. The ensuing sections then provide further details on the various phases of the development, from the UMo dispersion in a pure Al matrix through the addition of Si to the matrix to address the interaction layer formation and finally to the more advanced solutions to the excessive swelling encountered in various experiments. This review was based only on published results or results that are currently in the process of being published.