High density UO2 and high thermal conductivity UO2 composites by spark plasma sintering (SPS)

Patent metadata

Field	Value
Publication number	US-10790065-B2
Application number	US-201314420702-A
Country	US
Kind code	B2
Filing date	Aug 15, 2013
Priority date	Aug 15, 2012
Publication date	Sep 29, 2020
Grant date	Sep 29, 2020

Abstract

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 UO2 and a highly thermally conductive material, such as SiC or diamond.

First claim

We claim:

1. A method for preparing a nuclear fuel pellet, comprising:
providing 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.;
treating an inner surface of a die assembly with graphite;
loading said die assembly with said nuclear fuel powder;
placing said die assembly in a sintering chamber of a spark plasma sintering (SPS) apparatus;
applying a pulsed current, a vacuum, and heat, wherein temperature is increased to a maximum temperature of 850 to 1600° C. and a rate of increase after achieving a temperature of 600° C. is at least 50° C./minute;
maintaining said maximum temperature for a period of 0.5 to 20 minutes; and
applying a controlled pressure of 25 to 100 MPa to said die assembly during said period for said maintaining of said maximum temperature, wherein a nuclear fuel pellet is formed, and wherein during the maintaining said period, holding said maximum temperature and said controlled pressure until a density of more than 95% TD (theoretical density) of the UO 2 , an average grain size of at least 4 μm with little or no inter-granular pores, and a thermal conductivity of at least 8 W/mK at 100° C. is attained.

2. The method according to claim 1 , wherein said nuclear fuel powder further comprises other oxidation states of uranium oxide, uranium nitride, thorium oxide, plutonium oxide, or any combination thereof.

3. The method according to claim 1 , wherein said controlled pressure is a uniaxial pressure of 30 to 45 MPa.

4. The method according to claim 3 , wherein said nuclear fuel powder comprising a nuclear fuel comprises a mixture of oxides with a stoichiometry UO 2+x where x is 0 to 0.25, and wherein said maximum temperature is 1050° C. to 1600° C.

5. The method according to claim 1 , wherein said nuclear fuel powder consists of UO 2 and said thermally conductive material, and said thermally conductive material is uranium or a uranium alloy.

6. The method according to claim 1 , further comprising reducing said fuel pellet by exposure to a reducing gas.

Assignees

• Univ Florida

Inventors

• Subhash Ghatu
• Baney Ronald Howard
• Tulenko James S
• Mckenna Iii Edward
• Ge Lihao
• Yeo Sunghwan

Classifications

• G21C3/62 Primary

  Ceramic fuel · CPC title

• Y02E30/30

  Nuclear fission reactors · CPC title

• G21C3/045

  Pellets · CPC title

• G21C21/02 Primary

  Manufacture of fuel elements or breeder elements contained in non-active casings · CPC title

• Y02E30/38

  Cross-Sectional Technologies · mapped topic