Towards the Development of Novel Cladding Materials based on Nanolaminated Ternary Carbides (MAX phases) for Different Nuclear Systems

After the Fukushima Daiichi accident, the interest in novel materials to improve nuclear safety and fuel cladding performances not only during service conditions but also in accident scenarios increased. For light water reactors (LWR) such as the Fukushima Daiichi reactor, the accident tolerant fuel (ATF) concept was developed to increase the coping-time during an accident by increasing oxidation resistance of the clad materials while satisfying all other design requirements. In addition to LWRs, material-based challenges exist for the new generation reactor designs such as GEN-IV lead fast reactors (LFR) with higher fuel efficiency, reduced waste generation and increased safety. For LFRs, the use of liquid metals as reactor primary coolant raised the corrosion problems. MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) being built by the Belgian Nuclear Research Center will have a liquid lead-bismuth eutectic alloy as primary coolant. Candidate cladding materials like stainless steels corrode in liquid metal in the absence of a protective oxide layer. Therefore, intrinsically oxidation and corrosion-resistant novel materials that are suitable for cladding applications are needed. MAX phase ceramics are considered as one of the candidate materials, either as monoliths or as protective coatings on existing clad materials.

MAX phases are inherently nanolaminated ceramics with Mn+1AXn (n=1, 2 or 3) stoichiometry. M is an early transition metal, A is mostly an element from group 13-14 in the periodic table and X is either C or N. Their unit cell consists of transition metal carbide or nitride blocks separated by an atomic layer of metal. Due to this ceramic interleaved with metal stacking, they possess properties from both metals and ceramics. They are soft and easily machinable, good conductors of heat and electricity, chemically resistant to corrosive environments, elastically stiff, have good mechanical properties at high temperature and some have good irradiation tolerance. In addition to ternary MAX phases, substitutional solid solution MAX phases can be synthesized using multiple elements for M, A or X in the MAX phase structure. The expanding list of MAX phase forming elements in the periodic table allows the flexibility to fine-tune their properties, to better suit targeted applications.

The aim of this PhD research was to explore the potential of MAX phases as candidate nuclear cladding materials for LWR and LFR nuclear systems. (Zr,Ti)n+1AlCn solid solution MAX phases were initially synthesized using powder metallurgical routes. The aim was to improve the properties of Zrn+1AlCn MAX phases which were interesting due to their low neutron absorption, a requirement especially important for LWRs. An out-of-plane chemically ordered solid solution MAX phase was discovered in this solid solution system but phase purity, an important criterion, was not met. Next, double solid solution (Zr,Ti)2(Al,Sn)C MAX phases were synthesized. With the addition of Sn, obtaining near phase purity was possible. Multiple synthesis routes were studied including reactive hot pressing and pressureless sintering. Porous (Zr,Ti)2(Al,Sn)C MAX phases with near phase purity were produced by pressureless sintering which is a less costly method for up-scaling the production of phase pure MAX phase powders. The thermal expansion properties of (Zr,Ti)2(Al,Sn)C phases with different Zr:Ti ratios were investigated for their use at elevated temperatures. These MAX phases also had an improved oxidation resistance compared to Zr2(Al,Sn)C.

In literature, among the MAX phases tested in LBE, Zr-rich MAX phases were reported to have the most severe interaction which leds to the in-situ formation of new solid solution MAX phases containing Bi and Pb. To understand this interaction mechanism, LBE tested Zr2AlC MAX phase was selected and examined in detail. An interaction mechanism and potential ways to mitigate the reaction were proposed. Potential mitigation techniques were later tested in Pb and LBE coolants and found successful. In addition, phase pure double solid solution MAX phases were tested in oxygen-poor LBE and Pb to assess their inherent corrosion resistance without any potential protective oxide layers, revealing no interaction.

Finally, the near phase pure double solid solution (Zr0.5,Ti0.5)2(Al0.5,Sn0.5)C MAX phase was in-situ irradiated in a transmission electron microscope using He ions to simulate reactor-like conditions. The evolution of irradiation-induced defects as a function of damage levels and reactor relevant temperatures were studied. Due to He ion implantation, He-bubbles or platelets were observed. Dimensions of these He-complexes were overall small and stable, indicating manageable He accumulation in the material without large void formations. Recovery of damage at high temperatures and no amorphization or decomposition were observed. The overall findings indicated a good He ion radiation tolerance.