Abstracts and Available Papers Presented at the
2000 International RERTR Meeting

NEUTRONIC FEASIBILITY STUDIES USING U-Mo DISPERSION FUEL
(9 Wt % Mo, 5.0 gU/cm3) FOR LEU CONVERSION OF THE
MARIA (POLAND), IR-8 (RUSSIA), AND WWR-SM (UZBEKISTAN) RESEARCH REACTORS

M. M. Bretscher, J. R. Deen, N. A. Hanan, and J. E. Matos

Argonne National Laboratory
Argonne, Illinois 60439-4841 USA

ABSTRACT

U-Mo alloys dispersed in an Al matrix offer the potential for high-density uranium fuels needed for the LEU conversion of many research reactors.  On-going fuel qualification tests by the US RERTR Program show good irradiation properties of U-Mo alloy dispersion fuel containing 7-10 weight percent molybdenum.  For the neutronic studies in this paper the alloy was assumed to contain 9 wt % Mo (U-9Mo) with a uranium density in the fuel meat of 5.00 gU/cm3 which corresponds to 32.5 volume % U-9Mo.  Fuels containing U-9Mo have been used in Russian reactors since the 1950’s.  For the three research reactors analyzed here, LEU fuel element thicknesses are the same as those for the Russian-fabricated HEU reference fuel elements.

Relative to the reference fuels containing 80-90% enriched uranium, LEU U-9Mo Al-dispersion fuel with 5.00 gU/cm3 doubles the cycle length of the MARIA reactor and increases the IR-8 cycle length by about 11%.  For the WWR-SM reactor, the cycle length, and thus the number of fuel assemblies used per year, is nearly unchanged.  To match the cycle length of the 36% enriched fuel currently used in the WWR-SM reactor will require a uranium density in the LEU U-9Mo Al-dispersion fuel of about 5.4 gU/cm3.  The 5.00 gU/cm3 LEU fuel causes thermal neutron fluxes in water holes near the edge of the core to decrease by (6-8)% for all three reactors.

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