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SiC and SiC/SiC-composite claddings

Accident tolerant fuels (ATF) are a series of new nuclear fuel concepts researched to improve fuel performance during normal operation, transient conditions, and accident scenarios, such as loss-of-coolant accidents (LOCA) or reactivity-initiated accidents (RIA). Following the Fukushima Daiichi accident, a fuel behavior review was initiated. Zirconium alloy clad fuel operates successfully to high burnup and is the result of 40 years of continuous development and improvement. However, under severe accident conditions, the high-temperature zirconium–steam interaction can be a major source of damage to the power plant.

These upgrades include:

  • specially designed additives to standard fuel pellets intended to improve various properties and performance
  • robust coatings applied to the outside of standard claddings intended to reduce corrosion, increase wear resistance, and reduce the production of hydrogen under high-temperature (accident) conditions
  • development of completely new fuel designs with ceramic cladding and different fuel materials

Current fuel cladding is the outer layer of the fuel rods, standing between the reactor coolant and the nuclear fuel (i.e., fuel pellets). It is made of corrosion-resistant material with a low absorption cross section for thermal neutrons (~ 0.18 × 10–24 cm2), usually zirconium alloy. It prevents radioactive fission products from escaping the fuel matrix into the reactor coolant and contaminating it. Cladding constitutes one of the barriers to the ‘defence-in-depth‘ approach; therefore, its coolability is one of the key safety aspects.

Special Reference: Nuclear Energy Agency, State-of-the-Art Report on Light Water Reactor Accident-Tolerant Fuel. NEA No.7317, OECD, 2018.

SiC and SiC/SiC-composite claddings

Silicon carbide is an exceedingly hard, synthetically produced crystalline compound of silicon and carbon, and its chemical formula is SiC. Silicon carbide has a Mohs hardness rating of 9, approaching that of a diamond. In addition to hardness, silicon carbide crystals have fracture characteristics that make them extremely useful in grinding wheels. 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 superior physical/chemical properties. The composite consists of SiC fibers wrapped around a SiC inner layer and surrounded by a SiC outer layer. Problems have been reported with the ability to join the pieces of the SiC composite.

SiC cladding is intended to provide groundbreaking safety margin improvements. SiC cladding reacts many orders of magnitude slower with water and steam than zirconium at critical temperatures (above 800°C), resulting in the minimal generation of heat and hydrogen in beyond-design-basis accident scenarios. The SiC-composite claddings and fuel components are expected to provide excellent passive safety features both in design-basis accidents and design extension conditions severe accidents (SAs). Moreover, the SiC/SiC composites are anticipated to provide additional benefits over the Zr-alloys, such as a reduced neutron absorption cross-section enabling a smaller uranium enrichment. These attractive features make the SiC composites one of the leading candidates for accident-tolerant LWR fuel cladding and core structures.

There are three main disadvantages of the SiC/SiC composite fuel clads:

  • Fabrication. Fabrication of thin fuel clad requires further development. A technology for end-plug joining with gas tightness and adequate strength should also be developed because SiC ceramics cannot be welded.
  • Tritium releases. There is a potential increase in tritium release into the reactor coolant. Tritium is produced as a fission product (FP). SiC does not react with hydrogen to form stable hydrides like a zirconium-based alloy, resulting in higher permeability of tritium through cladding to the reactor coolant. The choice of using a proper liner material could help mitigate this issue.
  • SiC/SiC has significantly lower thermal conductivity than zirconium alloys. This fact negatively influences pellet centerline temperatures and also the coping time.

One possible SiC composite claddings have been developed by General Atomics and are known as SiGA™ silicon-carbide (SiC) composite, in which the SiC matrix material is reinforced with flexible SiC fiber in much the same way that steel rebar reinforces concrete. This creates an extremely hard and durable material that can withstand the harshest reactor conditions.

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