• Title/Summary/Keyword: fuel energy

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USE OF A CENTRIFUGAL ATOMIZATION PROCESS IN THE DEVELOPMENT OF RESEARCH REACTOR FUEL

  • Kim, Chang-Kyu;Park, Jong-Man;Ryu, Ho-Jin
    • Nuclear Engineering and Technology
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    • v.39 no.5
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    • pp.617-626
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    • 2007

  • A centrifugal atomization process for uranium fuel was developed in order to fabricate high uranium density dispersion fuel for advanced research reactors. Spherical powders of $U_3Si$ and U-Mo were successfully fabricated and dispersed in aluminum matrices. Thermal and mechanical properties of dispersion fuel meat were characterized. Irradiation tests at the research reactor HANARO confirm the excellent performance of high uranium density dispersion fuel.

  • https://doi.org/10.5516/NET.2007.39.5.617 인용 PDF KSCI

Enthalpy and Void Distributions in Subchannels of PHWR Fuel Bundles

  • Park, J.W.;Choi, H.;Rhee, B.W.
    • Proceedings of the Korean Nuclear Society Conference
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    • 1998.05a
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    • pp.502-507
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    • 1998

  • Two different types the CANDU fuel bundles hue been modeled for the ASSERT-IV code subchannel analysis. From calculated values of mixture enthalpy and void paction distributions in the fuel bundles, it is found that net buoyancy effect is pronounced in the central region of the DUPIC fuel bundle when compared with the standard CANDU fuel bundle. It is also found that the central region of the DUPIC fuel bundle can be cooled more efficiently than that of the standard fuel bundle. From calculated mixture enthalpy distribution at the exit of fuel channel, it is found that the mixture enthalpy and void fraction can be highest in the peripheral region of the DUPIC fuel bundle. On the other hand, the enthalpy and the void fraction were found to be highest in the central region of the standard CANDU fuel bundle at the exit of the fuel channel. This study shows that the subchannel analysis is very useful assessing thermal behavior of the fuel bundle that could be used in CANDU reactors.

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FUEL PERFORMANCE CODE COSMOS FOR ANALYSIS OF LWR UO2 AND MOX FUEL

  • Lee, Byung-Ho;Koo, Yang-Hyun;Oh, Jae-Yong;Cheon, Jin-Sik;Tahk, Young-Wook;Sohn, Dong-Seong
    • Nuclear Engineering and Technology
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    • v.43 no.6
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    • pp.499-508
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    • 2011

  • The paper briefs a fuel performance code, COSMOS, which can be utilized for an analysis of the thermal behavior and fission gas release of fuel, up to a high burnup. Of particular concern are the models for the fuel thermal conductivity, the fission gas release, and the cladding corrosion and creep in $UO_2$ fuel. In addition, the code was developed so as to consider the inhomogeneity of MOX fuel, which requires restructuring the thermal conductivity and fission gas release models. These improvements enhanced COSMOS's precision for predicting the in-pile behavior of MOX fuel. The COSMOS code also extends its applicability to the instrumented fuel test in a research reactor. The various in-pile test results were analyzed and compared with the code's prediction. The database consists of the $UO_2$ irradiation test up to an ultra-high burnup, power ramp test of MOX fuel, and instrumented MOX fuel test in a research reactor after base irradiation in a commercial reactor. The comparisons demonstrated that the COSMOS code predicted the in-pile behaviors well, such as the fuel temperature, rod internal pressure, fission gas release, and cladding properties of MOX and $UO_2$ fuel. This sufficient accuracy reveals that the COSMOS can be utilized by both fuel vendors for fuel design, and license organizations for an understanding of fuel in-pile behaviors.

  • https://doi.org/10.5516/NET.2011.43.6.499 인용 PDF KSCI