04-19 TRISO nuclear fuel

[Image above] Drawing of a tristructural isotropic coated particle fuel, which is meant for use in nuclear reactors. Researchers are exploring the potential of various ceramics to replace the silicon carbide layer in these fuels. Credit: Electric Power Research Institute, YouTube

As the world looks to transition away from fossil fuels, public views on nuclear power remained mixed, according to a recent Pew Research Center survey.

Safety of nuclear reactor operations and radioactive waste disposal remain top concerns for people when debating use of nuclear power. The U.S. Department of Energy is actively funding research on these topics, and one method that has been in development since the 1960s—TRi-structural ISOtropic (TRISO) coated particle fuel—is finally moving toward commercial scale deployment.

TRISO coated particle fuel is a way of designing nuclear fuel that is expected to reduce the likelihood of reactor meltdown and improve containment of radioactive products. In modern nuclear reactors, uranium-based fuel pellets are stacked together in sealed metal tubes called fuel rods, which are bundled together to form a fuel assembly. Fission reactions are kept under control by inserting so-called control rods among the fuel rods to absorb some of the neutrons released from the fission reaction. The reactor is also enclosed in a massive containment structure designed to prevent radioactive material from escaping if something goes wrong.

In contrast to this current setup, TRISO particle fuel encapsulates each uranium-based pellet in multiple layers of carbon- and ceramic-based materials. When fission occurs, these layers contain the radioactive fission products within each pellet and moderate the movement of neutrons, thus rendering modern safety measures of control rods and containment structures somewhat redundant, opening the door for smaller reactor designs.

Many current TRISO coated particle fuels consist of a silicon carbide layer between inner and outer pyrolytic carbon layers. The silicon carbide layer provides most of the structural strength and dimensional stability, and it serves as the primary barrier to prevent release of fission products.

Silicon carbide is used due to its superior irradiation tolerance as well as high thermal conductivity and low thermal expansion. However, silicon carbide does face some drawbacks, specifically the loss of mechanical strength above 1,700°C, which limits the operating temperature of the nuclear reactor.

To allow for higher operating temperatures, researchers have explored using alternative ceramics such as zirconium carbide, titanium carbide, and titanium nitride in TRISO particle fuels. However, for many possible carbides and nitrides, much fundamental data remains unknown.

In a recent study, researchers from the University of Tennessee, Knoxville; Los Alamos National Laboratory; and Oak Ridge National Laboratory looked to provide fundamental data for possible layer materials silicon nitride (Si3N4) and zirconium nitride (ZrN).

The researchers explain that both ceramics have shown good overall properties in previously published papers, such as low linear swelling, superior retention of thermal diffusivity, and greater flexural strength compared to oxide ceramics. However, there are gaps in the literature regarding their irradiation behavior, specifically in the 300–700°C temperature range.

They conducted ion irradiation experiments at the Ion Beam Material Laboratory at the University of Tennessee, Knoxville. Samples of silicon nitride and zirconium nitride were irradiated using 15 MeV Ni5+ ions at 300, 500, and 700°C to midrange irradiation doses of 1 and 15 displacements per atom (dpa) and at 700°C to 50 dpa. They used nickel ion due to the ability to achieve relatively deep irradiation regions with relatively high beam currents and primary knock-on atom energies.

Analysis of the samples led to several important observations, including the following.

Hardness degradation of Si3N4 was due to sintering aid

Analysis of the silicon nitride samples showed some irradiation induced hardening at 300°C, but the ceramic experienced a drop in mechanical properties starting at 15 dpa at 500°C and accentuated at 700°C. Scanning transmission electron microscopy and energy-dispersive X-ray spectroscopy revealed that the sintering aid yttrium aluminum garnet was present in the samples, mainly along grain boundaries, and it dramatically degraded under ion irradiation.

“Therefore, the observed hardness degradation (softening) of Si3N4 measured in this work for the higher doses and irradiation temperatures is considered to be an artifact due to the presence of the intergranular YAG phase and is not representative of the intrinsic mechanical properties’ behavior of Si3N4,” the researchers write. “[A]dditional research is needed to identify a suitable sintering aid for Si3N4 that doesn’t react poorly during irradiation.”

In contrast, zirconium nitride, which did not contain a sintering aid, experienced a rapid increase in hardness at low dose and reached saturation by 20 dpa. This increase held steady at higher doses, from 50 dpa (this work) up to 200 dpa (Robertson, 2019), “reinforcing the idea that the hardness increase is conserved with increasing dose,” the researchers write.

Lattice swelling presented both expected and unexpected results

Overall, both ceramics displayed a maximum volumetric swelling below 0.7%, which indicates good dimensional stability up to 50 dpa. But the samples also displayed unique behavior at 700°C.

Silicon nitride irradiated at 50 dpa showed a moderate lattice swelling increase up to 0.55%, in contrast to the expected saturation value of 0.22%. Zirconium nitride experienced an increase in lattice swelling with increasing dose at 700°C, without showing saturation behavior, up to 50 dpa.

“Radiation enhanced diffusion of the implanted nickel ions into the midrange region could contribute to lattice parameter increase and therefore needs to be further evaluated,” the researchers write.

High density of small cavities was observed under all irradiation conditions

No previous studies have reported cavities forming in silicon nitride and zirconium nitride under neutron or conventional nongaseous ion irradiation. Yet the new study revealed a slight coarsening above 15 dpa (1.5 to 2 nm diameter) and increased cavity swelling with increasing dose (from 0.05 to 0.08% in silicon nitride; from 0.06 to 0.16% in zirconium nitride).

While origin of the cavity formation remains mostly unsolved, the researchers suggest that “Ionization enhanced radiolysis is a potential reason for the formation of these cavities, where freed nitrogen atoms would stabilize cavity formation.”

Ultimately, “Both silicon nitride and zirconium nitride appeared to be quite promising for demanding nuclear applications, showing mechanical properties increase at all studied temperatures and an overall low volumetric swelling following irradiation,” they conclude.

The paper, published in Journal of Nuclear Materials, is “Properties and microstructure evolution of silicon nitride and zirconium nitride following Ni ion irradiation” (DOI: 10.1016/j.jnucmat.2022.153643).