• 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.5
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• pp.715-722
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• 2009

A model to predict failure of designed-to-fail (dtf) fuel particles is discussed. The dtf fuel under study consisted of a uranium oxycarbide kernel coated with a single pyrocarbon seal coat. Coating failure was assumed to be due to fission gas recoil and knockout mechanisms and direct diffusive release of fission gas from the kernel, which acted to increase pressure and stress in the pyrocarbon layer until it ruptured. Predictions of dtf fuel failure using General Atomics' particle fuel performance code for HRB-17/18 and HFR-B1 irradiation tests were reasonably accurate; however, the model could not predict the failure for COMEDIE BD-1. This was most likely due to insufficient information on reported particle fuel failure at the beginning.

• 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.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.335-335
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 1998.10a
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• pp.254-254
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• 1998

• Proceedings of the Korean Nuclear Society Conference
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• 2001.10a
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• pp.226-226
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 2002.05a
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• pp.241.2-241
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• 2002

• Nuclear Engineering and Technology
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• v.38 no.2
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• pp.163-174
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• 2006

The $Ph{\acute{e}}bus$ FP project is an international reactor safety project. Its main objective is to study the release, transport and retention of fission products in a severe accident of a light water reactor (LWR). The FPT4 test was performed with a fuel debris bed geometry, to look at late phase core degradation and the releases of low volatile fission products and actinides. Post Test Analyses results indicate that releases of noble gases (Xe, Kr) and high-volatile fission products (Cs, I) were nearly complete and comparable to those obtained during $Ph{\acute{e}}bus$ tests performed with a fuel bundle geometry (FPT1, FPT2). Volatile fission products such as Mo, Te, Rb, Sb were released significantly as in previous tests. Ba integral release was greater than that observed during FPT1. Release of Ru was comparable to that observed during FPT1 and FPT2. As in other $Ph{\acute{e}}bus$ tests, the Ru distribution suggests Ru volatilization followed by fast redeposition in the fuelled section. The similar release fraction for all lanthanides and fuel elements suggests the released fuel particles deposited onto the plenum surfaces. A blockage by molten material induced a steam by-pass which may explain some of the low releases. The revaporisation testing under different atmospheres (pure steam, $H_2/N_2$ and steam /$H_2$) and up to $1000^{\circ}C$ was performed on samples from the first upper plenum. These showed high releases of Cs for all the atmospheres tested. However, different kinetics of revaporisation were observed depending on the gas composition and temperature. Besides Cs, significant revaporisations of other elements were observed: e.g. Ag under reducing conditions, Cd and Sn in steam-containing atmospheres. Revaporisation of small amounts of fuel was also observed in pure steam atmosphere.

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.929-930
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• 2004

• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.258-259
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• 2005