Uranium Silicide Fuel

Most of PWRs use the 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 the uranium dioxide has 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), which is made of zirconium alloys due to its very low absorption cross-section (unlike the stainless steel). The surface of the tube, which covers the pellets, is called fuel cladding. Fuel rods are 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. Typical fuel rod, has a length of some 4 m, with a 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, such as oxide-doped UO2 and high-thermal conductivity UO2 (designed by adding metallic or ceramic dopant).
  • 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.

High-density fuel

Most of the metallic materials suggested for use as cladding to reduce the steam oxidation present fairly large reactivity penalties compared to the traditional Zr-based claddings. These penalties can be compensated by either an increase of the 235U enrichment and/or a decrease in the cycle length. To compensate for this without the previous concessions, the fissile density in the pellet has to be increased. 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 sevaral proposed designs of high-density fuel, but it must be noted, 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 possible designs of accident tolerant fuel pellet materials proposed. Advantages are higher percentage of uranium and higher thermal conductivity. With a density uranium silicide of 12.2 g/cm3 (vs. ), uranium silicide (U3Si2) offers a boost in fuel economics. Uranium dioxide has a density of 10.97 g/cm3. Moreover, there is a surplus from its stechiometric composition. In final, there is about 17% higher uranium density than that of uranium dioxide. A direct replacement of UO2 with U3Si2 should enable a reactor to generate more energy from a set of fuel rods and also provide more “coping time” in the case 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 such that fuel operating and safety margins are improved.

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 programme. The improved thermal performance of U3Si2 compared to UO2 fuel allows 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
TRISO, TRI-structural ISO-tropic, is a type of micro fuel particle, which consists of fissile material-bearing kernels that are coated with multiple layers of porous or dense carbon and silicon carbide. Source: energy.gov License: Public Domain

TRISO, TRI-structural ISO-tropic, is a type of micro fuel particle, which consists of fissile material-bearing kernels that are 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 own containment system thanks to its triple-coated layers. This allows them to retain fission products under all reactor conditions. TRISO particles can withstand extreme temperatures that are well beyond the threshold of current nuclear fuels. TRISO-SiC-composite pellets consist of TRISO fuel particles embedded in an 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 generally called a fully ceramic microencapsulated (FCM) fuel.

According to NEA report, the TRISO-SiC fuel is conceived as a promising medium-term concept to replace current UO2 fuel pellets. It has potentially superior safety characteristics relative to other fuel forms as a result of 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. In order to increase the fissile loading, the combination of uranium enrichment up to the practical upper limit of LEU(~19.7% of 235-U), increasing kernel-to-particle volume fraction and TRISO-packing fraction, and enlarging fuel pin diameter was proposed.

Silicon carbide is 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 diamond. Its high thermal conductivity, together with its high-temperature strength, low thermal expansion, and resistance to chemical reaction, makes silicon carbide valuable in the manufacture of high-temperature applications and other refractories. In 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 other superior physical/chemical properties.

References:
Materials Science:

U.S. Department of Energy, Material Science. DOE Fundamentals Handbook, Volume 1 and 2. January 1993.
U.S. Department of Energy, Material Science. DOE Fundamentals Handbook, Volume 2 and 2. January 1993.
William D. Callister, David G. Rethwisch. Materials Science and Engineering: An Introduction 9th Edition, Wiley; 9 edition (December 4, 2013), ISBN-13: 978-1118324578.
Eberhart, Mark (2003). Why Things Break: Understanding the World by the Way It Comes Apart. Harmony. ISBN 978-1-4000-4760-4.
Gaskell, David R. (1995). Introduction to the Thermodynamics of Materials (4th ed.). Taylor and Francis Publishing. ISBN 978-1-56032-992-3.
González-Viñas, W. & Mancini, H.L. (2004). An Introduction to Materials Science. Princeton University Press. ISBN 978-0-691-07097-1.
Ashby, Michael; Hugh Shercliff; David Cebon (2007). Materials: engineering, science, processing and design (1st ed.). Butterworth-Heinemann. ISBN 978-0-7506-8391-3.
J. R. Lamarsh, A. J. Baratta, Introduction to Nuclear Engineering, 3d ed., Prentice-Hall, 2001, ISBN: 0-201-82498-1.

See above:
Accident Tolerant Fuel