The release of radioactivity to the environment is a major threat to people’s health in the case of a nuclear accident. The most critical barrier to the release of radioactivity to the environment is the fuel cladding. As discussed in the previous article, the innovation of materials in the nuclear sector has been rather slow compared to other sectors.
The conventional Zircaloys discovered in the 1950s, with some modifications, are still being used as cladding materials in current reactors. Efforts are underway to find and test the alternative fuel cladding materials better than Zircaloy, this has gained momentum after the recent Fukushima accident.
The aim of this article is to give a general idea about the promise, status, and challenges of Accident Tolerant Cladding (ATC) materials.
In 2011, a tsunami wave spanning a height of 15 m hit three Fukushima Daiichi reactors . After the fission reaction was stopped, the fission products were still in the radioactive state, hence they emitted beta, alpha and gamma rays. These rays produce what is called decay heat in the reactor. Hence, the coolant needed to be kept circulating to cool the reactor.
The diesel generators, which were planned to be used as an emergency backup for circulating the coolant in the reactor failed. Because of failure of heat removal by circulation to an outside heat exchanger, the temperature close to the zirconium cladding rose to about 1200℃. Zirconium oxidises easily above 1200℃, reacting with water to release hydrogen and forming ZrO2. This reaction is exothermic and produces large amounts of heat and hydrogen.
Zr + 2H2O → ZrO2 + 2H2 
This is called a Loss of Coolant Accident (LOCA). The amount of generated hydrogen gas blew off the containment rooftop. Due to severe oxidation, the Zr cladding failed and lead to the release of radioactivity to the environment.
Accident tolerant cladding
An ideal cladding metal should prevent the leak of radioactivity even under a worst-case scenario. Accident tolerant fuel and cladding materials are currently hot topics of research. The major emphasis is on materials which have significantly higher oxidation resistance than Zircaloys.
It is known that SiC offers the best oxidation resistance followed by FeCrAl (ferritic steel) and austenitic stainless steel. One or more of the elements silicon, aluminium and chromium must be present in accident tolerant cladding material, since these elements form a protective oxide layer, providing corrosion resistance.
FeCrAl alloys have been primarily used as heating elements and components in high-temperature furnaces due to their superior oxidation resistance compared to many other common materials. Over the past half-century, FeCrAl alloys have also been considered for structural applications for varying industrial applications including for the nuclear power industry .
FeCrAl alloys possess excellent oxidation resistance up to 1500℃ (close to its melting point) . If the FeCrAl tube has an alumina layer on the surface, it will dissolve in high-temperature (300℃) water and a protective chromium oxide will form in its place. If FeCrAl tube has a chromium oxide layer on the surface and it is exposed to accident steam conditions, the Cr oxide layer will evaporate, and an alumina layer will form to protect the tube. So, Cr protects the alloy under normal operation conditions and Al protects the alloy at temperatures higher than 1100℃ .
There are still some major challenges that need to be overcome for FeCrAl alloys to be used as a cladding material:
- One of the major challenges is the increased parasitic neutron absorption of FeCrAl compared to zirconium alloys . The current cladding thickness of zirconium alloys is close to 600µm. For FeCrAl to have similar neutron absorption, the calculated cladding thickness should be around 300µm.
- Another major problem is alpha prime precipitation in FeCrAl after neutron irradiation, which makes it brittle.
- The third problem is that the FeCrAl cladding may release more tritium to the coolant than zirconium alloys . One of the suggested solutions for this problem is that the FeCrAl tubes can be coated with alumina coating from inside.
The most advanced and mature cladding solution, in development by AREVA NP, is the Cr-coated M5™ cladding, which consists of a 15 mm-thick dense chromium coating layer deposited on the surface of an M5™ cladding tube . In the same way, several other institutions worldwide are developing coatings, and especially Cr coatings, as a near-term ATC solution.
The Electric Power Research Institute is developing a Mo coating and the Korea Atomic Energy Research Institute is working on Cr-Al alloy coatings using 3-D laser deposition. Recent studies have shown improvement in oxidation resistance . A lot of variations in coating technologies require more research on this ATC. Literature suggests that this ATC is still in its infancy.
Silicon carbide (SiC) fiber–reinforced SiC matrix (SiC/SiC) composites
SiC fibers are braided, knitted or stitched into 3-dimensional (3D) tubes. This is referred to as the architecture of the tubes. An interphase layer is added by chemical vapor deposition (CVD). The matrix is added by either chemical vapor infiltration (CVI) or by nano-infiltration transient eutectic phase (NITE) sintering using hot pressing.
Inner or outer coating layers may be added using different techniques. The composite properties are, to a large extent, determined by the volume fractions and orientations of the fibers in relation to the orientation of interest for certain properties .
When silicon carbide is exposed to hot water it slowly forms a SiO2 protective layer but, surprisingly, this layer comes off by readily dissolving in water. This is referred to as SiC recession. It is doubtful that significant amounts of deposited silica in the reactor core can be allowed during operation, a topic which still needs clarification. In the case of processing defects, there is the possibility of fission gas escape from cladding .
Trends and predictions
Despite extensive research, we are still far away from the ideal cladding material for nuclear reactors. At one time, Zircaloy was thought to be the ideal material but after the LOCA accident in Fukushima, that’s not the case due to its poor oxidation resistance above 1200℃.
FeCrAl appears to be the closest candidate to replace Zircaloy in terms of the work carried out, ease of availability and low cost however the data on its high-temperature properties is limited which needs attention.
More research is needed on chromium-coated zirconium to comment on its feasibility.
SiC/SiC composites are anticipated to provide additional benefits over zirconium alloys: they provide smaller neutron absorption cross sections, general chemical inertness, ability to withstand higher fuel burn-ups and higher temperatures, exceptional inherent radiation resistance, lack of progressive irradiation growth and low induced activation/low decay heat. Moreover, SiC is considered to be permanently stable in nuclear waste .
Although SiC-based cladding appears to be attractive, critical feasibility issues such as hydrothermal corrosion, potential loss of fission gas retention due to cracking under normal operation conditions, development of fuel performance modeling capability and high processing cost must be addressed .
Low thermal conductivity in SiC leads to elevated fuel temperatures and large temperature gradients across the cladding that in turn induces large thermal stresses across the cladding thickness .
Although SiC/SiC composites are deemed the ideal ATC material, a lot of research is needed to overcome their current limitations.
I consider myself fortunate to be a part of the nuclear materials community involved in finding solutions to such grand challenges which were discussed briefly in the above two articles. In such critical applications, material failure is not an option!
This article is a part of the series. Please find the full reference list in the first article by Pratik Joshi Which High-Performance Materials Are Used in Nuclear Reactors?.
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