Accident tolerant fuels (ATF) are a series of new nuclear fuel concepts, researched in order to improve fuel performance during normal operation, transient conditions, and accident scenarios, such as loss-of-coolant accident (LOCA) or reactivity-initiated accidents (RIA). Following the Fukushima Daiichi accident, a review of fuel behaviour has been 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 a corrosion-resistant material with 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 constitute one of barriers in ‘defence-in-depth‘ approach, therefore its coolability is one of 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.
Chromium-coated Fuel Cladding
Chromium is one of possible coating elements for accident tolerant fuel. Cr-coated zirconium cladding and other metallic-coated claddings significantly reduce the high-temperature oxidation rates. The coating thickness is usually between 20 and 30 mm. All investigated coating materials (Cr, FeCrAl, Cr-Al, CrN) are harder than zirconium alloys so if the coating is sufficiently thick (> 30μm), then mechanical properties will be modified with increased strength and reduced ductility. The increased hardness of the coating materials has the benefit of potentially protecting the cladding against fretting and wear. Therefore Cr-coating may significantly reduce the risk for cladding damages due to debris or grid-to-rod fretting.
But the main advantage is that the coated cladding inherits all of the benefits of the base zirconium material properties but improves its oxidation and corrosion resistance for both normal operation and accident conditions. According to several investigations, Cr-coated cladding exhibits significantly increased post-quench strength and residual ductility. The strengthening effect of the Cr-coated cladding observed at high temperature is beneficial in that it delays the time to rupture and better preserves the coolable geometry of the nuclear fuel channel by mitigation of the flow blockage. Moreover, the corrosion of Cr-coated zirconium alloys is reduced to close to zero, thus also decreasing the hydrogen uptake by the cladding. The cladding will therefore not exhibit hydrogen embrittlement, leading to increased operational margins and potentially longer fuel rod irradiations.