Next Generation Reactors
Background
Commercial nuclear fuel cladding is currently made from various alloys of zirconium metal. These alloys are
used because they have reasonable strength at normal operating conditions and have fairly low neutron capture cross sections. Unfortunately, there
are several drawbacks to the material:
- The current maximum amount of time that a fuel assembly can remain in the reactor is limited a burnup of 62
GWD/MTU. The reason that this amount of time has been chosen is because the current cladding becomes embrittled due to increased
ZrH2 formation within the cladding. This hydride formation decreases the strength and ductility of the metal, which increases the
possibility that the cladding could fail if there is an accident.
- Zirconium alloy cladding will undergo a metal-water reaction above a certain temperature in which hydrogen
gas is generated leading to the possibility that the there could be a hydrogen explosion . The possibility of this reaction currently
limits the fuel centerline temperature to 2200°F.
- Zirconium alloy cladding forms an oxide layer that reduces the effective thermal conductivity of the
cladding. One of the current limitation to higher burnups is the 17% oxide layer limitation. Above that amount, the fuel rod unable to
withstand an accident at the end-of-life.
- The current generation of cladding is susceptible to fretting wear where the cladding interacts with the
spacer grids. In severe cases, this wear actually leads to leaking fuel pins increasing the level of radioactivity in the primary
system. This translates into increased costs for the utilities both because of the need to replace the damaged fuel rods and because of
increased dose to workers when the plant is being serviced.
- Zircalloy loses strength rapidly as temperature increases, as shown in the graph below. It also
experiences significant thermal and irradiation enhanced creep which causes the cladding to contract down until it contacts the fuel
pellets. The susceptibility to creep limits the internal pressure of the cladding to below primary loop operating pressure to prevent
"ballooning." The low stiffness of zircalloy requires a large number of spacer grids to prevent flow induced vibration.
These grids add to the pressure drop in the core and are the primary source of fretting wear.
SiC Composite Cladding Advantages
A leading commercial nuclear fuel supplier has estimated that the high-temperature tolerance of SiC cladding could allow for power upgrades
in existing commercial LWRs by as much as 30%. This could substantially increase the capacity of existing plants, thereby improving their
economics
Because of the maximum irradiation limit, commercial power reactors operate on an 18 month refueling cycle. If the cladding could
withstand higher irradiation, then it would be possible to increase the refueling outages to once every 24 months. This could save the utility
$240 million over the lifetime of the power plant.
SiC has a 25% lower thermal neutron cross-section than zirconium alloys resulting in greater neutron efficiency of the reactor.
The greater stiffness of SiC cladding should result in fewer spacer grids which could decrease the pressure drop through the core
helping with the power upgrades. The enhanced heat transfer from tailored surface roughness on the outside of the composite may eliminate the need
for intermediate flow mixing grids. The SiC cladding would also not be susceptible to fretting wear which could eliminate leaking fuel rods.
Safety
Because of its very high-temperature tolerances the SiC cladding will reduce accident risk and consequences. An increase in the cladding
operating temperature would allow for the fuel cladding to reach higher temperatures during accident or off-normal operating conditions. This
increase in temperature would give a higher safety margin if a LOCA or departure from nucleate boiling (DNB) were to occur, since the cladding
would not deform as quickly and would result in less potential for fuel damage and allow for a longer amount of time before safety injection
would need to occur.
There is no reaction with the water coolant at elevated temperatures. In fact, it appears that even the corrosion rate of the SiC
cladding is orders of magnitude lower than that of zircalloy.
Long Term Storage Reduction
SiC duplex cladding technology could increase the fuel burn-up of LWR fuels by as much as 70% and thereby reduce the repository burden from
commercial LWRs. The increase in the length of time that fuel stays in the reactor decreases the amount of spent fuel generated.
Next Generation Reactors
The advanced nuclear reactor concepts selected by the Generation IV International Forum for future development have high reactor coolant
outlet temperatures. For example, one of the designs of the Gas Cooled Fast Reactor (GFR) is a 600MWth helium cooled system
operating with an outlet temperature of 850° C, and sodium cooled reactors have temperatures ranging from 480°C for pool type reactors to 550°C.
These high temperatures are desired for several reasons. Liquid metal coolants, such as lead and sodium, have excellent nuclear
properties as required for the fast spectrum reactors needed to burn the higher actinides and reduce the long lived toxicity of spent fuel as
called for in the Advanced Fuel Cycle Initiative and in the DOE's recent GNEP (Global Nuclear Energy Partnership) proposal to Congress. In order
to function effectively, these liquid metal coolants must operate at temperatures where they can flow efficiently and remove heat from the
solid fuel elements.
Most metal and refractory alloy systems are not viable at these high temperatures because of creep rupture concerns. Although the sodium
cooled fast reactors already developed, such as the EBR-2 and FFTF in the US, Phoenix in France, and the BN-600 in Russia, have operated
successfully with stainless steel cladding, a significant penalty in neutron economy is incurred because of the high parasitic neutron
absorption of stainless steel.
