Page 233 - Managing Global Warming

Current and future nuclear power reactors and plants 191

4.5.3.2 Carbide fuels

Carbides of uranium and thorium have been considered as nuclear fuels. The use of
carbides of plutonium has also been investigated, but as mixed carbides such as
UC-PuC. Compared to MOX fuel, mixed carbide fuel has a higher thermal conduc-
tivity, higher heavy-metal density, and better neutron economy. Carbides of thorium,
ThC and ThC 2 , are the most stable compounds of thorium after thorium dioxide. ThC
is stable up to temperatures close to its melting point. Carbides of uranium have desir-
able properties such as high thermal conductivities and high melting points. Uranium
carbide (UC) and uranium dicarbide (UC 2 ) are two carbides of uranium, which can be
used as nuclear fuels. Uranium sesquicarbide (U 2 C 3 ) is another carbide of uranium.
U 2 C 3 cannot be manufactured through casting or compaction of a powder. But
UC 2 may transform to U 2 C 3 at high temperatures and under stress.

4.5.3.3 Nitride fuels
There are three compounds of uranium nitride system, namely, uranium mononitride
(UN), uranium dinitride (UN 2 ), and uranium sesquinitride (U 2 N 3 ). Among these com-
pounds, UN has been considered as nuclear fuel for use in space nuclear reactors and
sodium-cooled fast-breeder reactors, because of its superior properties such as high
thermal conductivity, high melting point, and high uranium-atom density. The fuel
residence time in the reactor core can be increased, when UN is used as a fuel.

4.5.4 Hydride fuels
A promising fuel for future use in LWRs is the hydride fuel, which has been used in
TRIGA reactors. Uranium-thorium-zirconium fuels are also of interest due to a sig-
nificant amount of thorium resources and utilization of the fuels in breeder reactors.

4.5.5 Composite fuels
Currently, there is a great interest in developing high thermal-conductivity fuel and/or
improving the thermal conductivity of low thermal-conductivity fuels such as UO 2 .
High thermal conductivities result in lower fuel centerline temperatures and limit
the release of gaseous fission products. As shown previously, UO 2 has a very low ther-
mal conductivity, especially, at high temperatures compared to other fuels such as UC,
UC 2 , and UN. However, research has shown that the thermal conductivity of oxide
fuels such as UO 2 can be increased by either adding a continuous solid phase or long,
thin fibbers of a high thermal-conductivity material.
A high thermal conductivity material must have a low neutron absorption cross sec-
tion depending on the reactor. In addition, it must have a high melting point and be
chemically compatible with the fuel, the cladding, and the coolant. The need to meet
these requirements narrows the potential materials to silicon carbide (SiC), beryllium
oxide (BeO), and graphite (C).