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Inertia and LLNL accelerate fusion fuel manufacturing
Cut-away schematic and example radiograph of the fusion fuel capsule, showing the outer spherical carbon shell and the D-T fuel “ice” layer with a gas core. For Inertia, this capsule has a diameter of about 4–5 mm. Formation of a sufficiently smooth D-T ice layer is crucial for ignition. (Image: Inertia)
Inertia Enterprises and Lawrence Livermore National Laboratory have developed a manufacturing process for the thin layer of cryogenically frozen deuterium-tritium (D-T) used in its target design, reducing production time from days to hours, according to the company.
The advance makes target fueling cheaper and reduces the amount of tritium that Inertia would need to hold at a pilot plant, which lowers material handling costs, regulatory burden, and dependence on scarce fuel inventories.
Argha Dutta, Apu Sarkar, Sandip Bysakh, Uttiyoarnab Saha, N. Gayathri, Santu Dey, P. Mukherjee
Nuclear Science and Engineering | Volume 197 | Number 12 | December 2023 | Pages 3160-3174
Regular Research Article | doi.org/10.1080/00295639.2023.2191580
Articles are hosted by Taylor and Francis Online.
One of the proposed materials for structural application in compact high-temperature reactors (CHTRs) is Nb-1Zr-0.1C alloy. Using the Variable Energy Cyclotron at the Variable Energy Cyclotron Centre, Kolkata, Nb-1Zr-0.1C alloy was irradiated with a 160-MeV oxygen (O6+) ion up to three different doses. Emulation of neutron irradiation by ion irradiation could be achieved as the weighted recoil spectra of the oxygen ion are found to be similar to the neutron recoil spectra of CHTRs within recoil energy ranging from 100 eV to 100 keV. The irradiated materials along with one as-received sample were characterized using different X-ray diffraction line profile analyses (XRDLPAs) to systematically evaluate microstructural parameters. A decrease in the domain size with an increase in microstrain and dislocation density is observed at first dose and then found to saturate with further irradiation. An increase in the Wilkens arrangement parameter indicates the formation of less correlated dislocations (clusters) after irradiation. Transmission electron microscopy analysis of as-received and highest-dose samples shows the formation of densely populated defect clusters after irradiation. Nanohardness increased after irradiation due to pinning of the dislocation movement by point defects and defect clusters/loops, as well as carbides in the matrix. The results extracted from the XRDLPAs are compared with our earlier studies of light ion–irradiated (H+) Nb-1Zr-0.1C alloy and oxygen-irradiated pure Nb to understand the effect of the type of ion and the alloying elements, respectively, on the evolution of the microstructure. It may be concluded that changes in dose and dose rate affect the movement of point defects toward sinks. Hence, possible correlated dislocation formation is observed in light ion–irradiated Nb alloy, but correlation is found to decrease with dose for heavy ion–irradiated Nb alloy. On the other hand, the presence of finely dispersed carbides restricts the formation of dislocation loops by making complexes with the defects in heavy ion (O6+)–irradiated Nb-1Zr-0.1C alloy, which is in contrast to pure Nb irradiated using the heavy ion (O6+) in a similar environment.
