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
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.
PDF version available
DOWNLOAD full paper in PDF format.
Manuel M. Bretscher
Argonne National Laboratory – 362
9700 South Cass Avenue
Argonne, IL 60439 USA
Phone: (630) 252-8616
Fax: (630) 252-5161