High density UO
Embodiments of the invention are directed to a method for production of a nuclear fuel pellet by spark plasma sintering (SPS), wherein a fuel pellet with more than 80% TD or more than 90% TD is formed. The SPS can be performed with the imposition of a controlled uniaxial pressure applied at the maximum temperature of the processing to achieve a very high density, in excess of 95% TD, at temperatures of 850 to 1600° C. The formation of a fuel pellet can be carried out in one hour or less. In an embodiment of the invention, a nuclear fuel pellet comprises UO 2 and a highly thermally conductive material, such as SiC or diamond.
1 . A system for preparing a nuclear fuel pellet, comprising:
a homogeneous dispersion of powders, as a nuclear fuel powder, the homogeneous dispersion of powders comprising uranium dioxide (UO 2 ) and a thermally conductive material selected from a metal or metal alloy, where the homogeneous dispersion of powders comprises 90 to 99 vol % of the UO 2 and the thermally conductive material has a thermal conductivity that is greater than 10 W/mK at 100° C.;
a spark plasma sintering (SPS) apparatus;
a die assembly loaded with the homogeneous dispersion of powders;
a power supply associated with the SPS apparatus, the power supply configured to apply a pulsed current to the die assembly; and
the SPS apparatus comprising a sintering chamber that receives the die assembly, where the SPS apparatus depressurizes the sintering chamber and increases temperature of the die assembly to a maximum temperature of 850 to 1600° C. as the pulsed current is applied to the die assembly, where the temperature is increased with a rate of increase of at least 50° C./minute after achieving a temperature of 600° C. and maintained at the maximum temperature for a period of 0.5 to 20 minutes while applying a controlled pressure of 25 to 100 MPa to the die assembly to form the nuclear fuel pellet, where the maximum temperature and the controlled pressure are held during the period until a density of more than 95% TD (theoretical density) of the UO 2 , an average grain size of at least 4 μm absent inter-granular pores, and a thermal conductivity of at least 8 W/mK at 100° C. is attained.
2 . The system of claim 1 , wherein an inner surface of the die assembly is treated with graphite prior to the homogeneous dispersion of powders being loaded, the graphite treatment forming a graphite layer between the inner surface of the die and the homogeneous dispersion of powders loaded in the die assembly.
3 . The system of claim 1 , wherein the controlled pressure is a uniaxial pressure.
4 . The system of claim 1 , wherein the die assembly comprises a die loaded with the homogeneous dispersion of powders between a first punch and a second punch.
5 . The system of claim 4 , wherein the die is a graphite die and the first and second punches are graphite punches.
6 . The system of claim 4 , wherein the first and second punches are cylindrical punches configured to insert into opposite sides of the die.
7 . The system of claim 4 , wherein an inner surface of the die is covered by graphite foil.
8 . The system of claim 4 , wherein an end surface of the first and second punches are coated with an aerosol layer of graphite.
9 . The system of claim 1 , wherein an outer surface of the die assembly is at least partially wrapped in graphite felt.
10 . The system of claim 1 , comprising a temperature measurement device configured to monitor the temperature of the die assembly in the sintering chamber.
11 . The system of claim 10 , wherein the temperature measurement device is a radiation pyrometer.
12 . The system of claim 1 , wherein the nuclear fuel pellet is reduced by exposure to a reducing gas.
13 . The system of claim 1 , wherein the UO 2 comprises a mixture of oxides with a stoichiometry UO 2+x where x is 0 to 0.25, and wherein the maximum temperature is 1050° C. to 1600° C.
14 . The system of claim 13 , wherein the controlled pressure is a uniaxial pressure of 30 to 45 MPa.
15 . The system of claim 1 , wherein the nuclear fuel powder comprises other oxidation states of uranium oxide, uranium nitride, thorium oxide, plutonium oxide, or any combination thereof.
16 . The system of claim 1 , wherein the nuclear fuel powder consists of UO 2 and the thermally conductive material.
17 . The system of claim 16 , wherein the thermally conductive material is uranium or uranium alloy.