Most PWRs use uranium fuel, which is in the form of uranium dioxide. Uranium dioxide is a black semiconducting solid with very low thermal conductivity. On the other hand, uranium dioxide has a very high melting point and has well-known behavior. The UO2 is pressed into pellets. These pellets are then sintered into the solid cylinder (with a height and diameter of about 1 centimeter, the height being greater than the diameter). The dimensions of the fuel pellets and other components of the fuel assembly are precisely controlled to ensure consistency in the characteristics of the fuel. These pellets are then loaded and encapsulated within a fuel rod (a metallic cladding tube) made of zirconium alloys due to their very low absorption cross-section (unlike stainless steel). The surface of the tube, which covers the pellets, is called fuel cladding. Fuel rods are the base element of a fuel assembly. Fuel rods have the purpose of containing fission products, ensuring mechanical support for the pellets, and allowing the heat removal to the coolant fluid of the heat generated by nuclear reactions. A typical fuel rod has a 4 m diameter of around 1 cm. An 1100 MWe (3300 MWth) nuclear core may contain 157 fuel assemblies composed of over 45,000 fuel rods and some 15 million fuel pellets.
Advanced Fuel Pellets
According to the NEA report, the fuel designs covered by the Task Force on Advanced Fuel Designs consist of three different concepts:
- Improved UO2 fuel. Regarding the improved UO2 fuel, this particular design was divided into two sub-concepts: oxide-doped UO2 and high-thermal conductivity UO2 (designed by adding metallic or ceramic dopants).
- High-density fuel.
- Encapsulated fuel (TRISO-SiC-composite pellets).
Special Reference: Nuclear Energy Agency, State-of-the-Art Report on Light Water Reactor Accident-Tolerant Fuel. NEA No.7317, OECD, 2018.
Improved UO2 fuel
Most PWRs use uranium fuel, which is in the form of uranium dioxide. Uranium dioxide is a black semiconducting solid with very low thermal conductivity. On the other hand, uranium dioxide has a very high melting point and has well-known behavior. The UO2 is pressed into pellets. These pellets are then sintered into the solid cylinder (with a height and diameter of about 1 centimeter, the height being greater than the diameter).
Regarding the improved UO2 fuel, this particular design was divided into two sub-concepts, such as:
- Doped UO2. Desirable attributes for accident-tolerant fuel (ATF) pellets include enhancing the retention of fission products (FPs) and minimizing pellet-cladding interaction. According to Westinghouse proposals, chromium (Cr2O3) and aluminium (Al2O3) doped UO2 pellet, known as our ADOPT pellet, achieves greater uranium efficiency through:
- Increased density of fissile material
- A higher creep rate than standard UO2 at high temperatures
- A higher thermal stability
- Reduced wash-out in the event of a fuel rod leaker
- Reduction of fission gas release in a transient scenario
- High-thermal conductivity UO2 (designed by adding metallic or ceramic dopant). Uranium dioxide is a black semiconducting solid with very low thermal conductivity. Thermal conductivity is one of the parameters which determine the fuel centerline temperature. This low thermal conductivity can result in localized overheating in the fuel centerline; therefore, this overheating must be avoided. The concept of cermet (ceramic-metallic) fuel for LWRs is considered. With a low volume fraction of highly conductive metallic additive, the CERMET fuel pellets present a higher conductivity than UO2 standard pellets, lowering the fuel temperature in normal operating conditions and increasing the margins concerning fuel melting in case of an accident.
Most metallic materials suggested for cladding to reduce steam oxidation present fairly large reactivity penalties compared to the traditional Zr-based claddings. These penalties can be compensated by either increasing the 235U enrichment and/or decreasing the cycle length. The fissile density in the pellet has to be increased to compensate for this without the previous concessions. The fissile density can be increased in several ways. One possible way is to increase the density of the material, and another one is to increase the metal to non-metal ratio in the metal compound fuels.
There are several proposed designs of high-density fuel, but it must be noted that all the high-density fuels are far from ready to be used as fuels in commercial light water reactors. The concepts include:
- Nitride Fuels
- Silicide Fuels
- Carbide Fuels
- Metallic Fuels
Uranium Silicide Fuel
Uranium silicide is an inorganic compound of uranium. It is one of the possible designs of accident-tolerant fuel pellet materials proposed. Advantages are a higher percentage of uranium and higher thermal conductivity. With a density of uranium silicide of 12.2 g/cm3 (vs. ), uranium silicide (U3Si2) boosts fuel economy. Uranium dioxide has a density of 10.97 g/cm3. Moreover, there is a surplus from its stoichiometric composition. Finally, there is about 17% higher uranium density than uranium dioxide. A direct replacement of UO2 with U3Si2 should enable a reactor to generate more energy from a set of fuel rods and provide more “coping time” in the case of severe accidents. Its thermal conductivity (~8.5 W/m.K at 300 K) is significantly higher than that of uranium dioxide at operating temperatures, and it increases as a function of temperature (uranium dioxide’s thermal conductivity decreases as a function of temperature). This thermal conductivity offsets its lower melting point, improving fuel operating and safety margins.
Westinghouse, the Idaho National Laboratory (INL), and the Los Alamos National Laboratory began developing and manufacturing uranium silicide and its composite fuels through DOE’s Accident Tolerant Fuel program. The improved thermal performance of U3Si2 compared to UO2 fuel allows the implementation of a more advanced cladding such as a SiC-SiC-composite, which besides the expected operational and safety benefits, also offers superior neutron economy and further fuel cycle cost savings relative to Zr-based claddings.
Encapsulated fuel – TRISO-SiC-composite pellets
TRISO, TRI-structural ISO-tropic, is a type of micro fuel particle consisting of fissile material-bearing kernels coated with multiple layers of porous or dense carbon and silicon carbide. Historically, TRISO particles have been utilized in fuel elements consisting of spherical pebbles or hexagonal prismatic blocks with graphite used as a matrix and coating for the fuel element. Each particle acts as its containment system thanks to its triple-coated layers, allowing them to retain fission products under all reactor conditions. TRISO particles can withstand extreme temperatures well beyond the threshold of current nuclear fuels. TRISO-SiC-composite pellets consist of TRISO fuel particles embedded in a SiC matrix. Using SiC as a matrix instead of graphite improves the radiation tolerance of the fuel matrix while also enhancing FP retention. The TRISO-SiC-composite fuel is a fully ceramic microencapsulated (FCM) fuel.
According to the NEA report, the TRISO-SiC fuel is conceived as a promising medium-term concept to replace current UO2 fuel pellets. It has superior safety characteristics relative to other fuel forms due to its multiple barriers to FP dispersion, high mechanical stability, and good thermal conductivity. A low fissile material loading density is the major issue for this concept. The combination of uranium enrichment up to the practical upper limit of LEU(~19.7% of 235-U) to increase the fissile loading, increase kernel-to-particle volume fraction and TRISO-packing fraction and enlarge fuel pin diameter was proposed.
Silicon carbide is an exceedingly hard, synthetically produced crystalline compound of silicon and carbon. Its chemical formula is SiC. Silicon carbide has a Mohs hardness rating of 9, approaching that of a diamond. Its high thermal conductivity, high-temperature strength, low thermal expansion, and resistance to a chemical reaction make silicon carbide valuable in the manufacture of high-temperature applications and other refractories. In the nuclear industry, silicon carbide composite material has been investigated for use as a replacement for zirconium alloy cladding in light water reactors. Silicon carbide (SiC) based ceramics and their composites have superior high-temperature (HT) properties, excellent irradiation resistance, inherent low activation, and another superior physical/chemical properties.