Ceramics in the Nuclear Fuel Cycle: Ceramics in the Nuclear Materials
Sponsored by: ACerS Energy Materials and Systems
Program Organizers: Cory Trivelpiece, Savannah River National Laboratory; Kyle Brinkman, Clemson University; Philip Edmondson, The University Of Manchester; Djamel Kaoumi, North Carolina State University
Tuesday 2:00 PM
November 3, 2020
Room: Virtual Meeting Room 11
Location: MS&T Virtual
2:00 PM
Evaluation of the Corrosion of High Purity CVD SiC in Light Water Reactor Environments: Peter Doyle1; Stephen Raiman2; Steven Zinkle1; 1University of Tennessee; 2Oak Ridge National Laboratory
SiCf/SiC composites have been identified as a potential accident tolerant fuel cladding. Among the R&D topics requiring investigation, aqueous corrosion behavior under normal operating conditions requires improved understanding. In the present work, SiC was exposed to high purity pressurized water in a constantly recirculating autoclave environment. Exposures ranged between 288°C and 350°C for times up to 2000 h, with either 1-4ppm O2 or 0.15-3ppm H2 dissolved in the water. Oxygen reacted with SiC with a reaction order of 1 and was initially linear with time until grain fallout became prevalent. No localized attack was observed in the absence of oxygen and uniform dissolution is predicted to be below 4µm/5 years, an acceptable rate. A predictive equation is given and compared to other published data. Recommendations are made for future testing parameters, include sample preparation. Funding was provided by the U.S. Department of Energy Office of Nuclear Energy, Advanced Fuel Campaign.
2:20 PM
First-principles Study on the Trapping and Recombination of Tritium in Lithium Vacancy of the γ-LiAlO2 (100) Surface: Ting Jia1; David Senor2; Yuhua Duan1; 1National Energy Technology Laboratory, United States Department of Energy; 2Pacific Northwest National Laboratory
γ-LiAlO2 enriched with the 6Li isotope absorbs neutrons and produce tritium (T) in tritium-producing burnable absorber rods (TPBARs). By increasing the number of T diffused from bulk to surface, upon trapped at lithium vacancies (VLi),recombination and desorption from surface sites, they will escape from the γ-LiAlO2 surface to form different T species (possibly being as T, OT, T2, or T2O) which need to be identified. Using the first-principles calculations, we investigate the trapping, recombination, and followed by desorption of T from γ-LiAlO2 (100) surface. Our results indicate that the trapped T atoms can be combined into T2 or T2O molecules after over-accumulation on the surface. The T2 molecule should be the main product at beginning. As the number of VLi increases under irradiation, the T2O yield could be increased. Such results are useful for improving the TPBAR’s performance with T high-yields.
2:40 PM
Nb and Ti Alloying Effects on the Phase and Oxidation of U3Si2: Geronimo Robles1; Joshua White2; Elizabeth Sooby Wood1; 1University of Texas at San Antonio; 2Los Alamos National Laboratory
U3Si2 has been identified as an accident tolerant nuclear fuel candidate for light water reactors due to its superior thermal conductivity and increased uranium density, when compared to traditional uranium dioxide (UO2). While reducing internal thermal stresses and increasing efficiency with this improvement, U3Si2 exhibits an energetic oxidation response less favorable than UO2 during off normal and accident scenarios including coolant or steam exposure. To mitigate this, Ti and Nb were chosen as alloying constituents for their corrosion resistance and strength benefits when alloyed in materials like steels. The work presented investigates the response of alloyed U3Si2 to water bearing atmospheres. Phase characterization of as-melted, thermally annealed and post oxidation compositions with up to 12 volume percent Nb and Ti by powder x-ray diffraction, scanning electron microscopy, and energy dispersive spectroscopy is reported. Included are select compositions in sintered pellet form. Thermogravimetric analysis conducted in flowing steam assesses oxidation resistance.
3:00 PM
Radiolytic Damage and Hydrogen Generation at Carbide – Water Interfaces: Simon Pimblott1; Jay LaVerne2; 1Idaho National Laboratory; 2University of Notre Dame
Carbides have great potential for application in the nuclear industry; however, various properties require detailed understanding for the materials to be properly utilized in situations where they will be exposed to extreme temperatures and mixed radiation fields. The behavior of SiC and ZrC as well as water slurries have been systematically studied using gamma and alpha radiation. Damage to the carbides was determined using analytical techniques including TGA, ICP-OES and SEM-EDS. Gamma irradiation of ZrC in air results in a reduction in Zr:C ratio with SEM examination showing oxidation of the surface. SiC is relatively stable under gamma irradiation except for some conversion of beta to the alpha phase: alpha radiolysis leads to the formation of SiO2. Aqueous slurry irradiations yield a large increase in the radiolytic yield of hydrogen compared to water, an effect that could have significant deleterious consequences if the material is deployed in nuclear energy environs.
3:20 PM
Thermophysical Properties of Sintered Yttrium Dihydride: Aditya Shivprasad1; Vedant Mehta1; Joshua White1; Michael Cooper1; Tarik Saleh1; Joseph Wermer1; Erik Luther1; Holly Trellue1; D.V. Rao1; 1Los Alamos National Laboratory
One current challenge to the nuclear industry is the ability to integrate nuclear energy with microgrids. Owing to their size, cost, and power output, microreactors could be designed to meet these needs. One such proposed microreactor design uses metal hydrides as the moderator due to their high hydrogen density, allowing enhanced fuel utilization and cost-effectiveness while keeping the core transportable. Yttrium dihydride is a promising candidate for this application due to its high thermal stability. Despite these advantages, it is difficult to produce YH2-x in geometries required for reactor design concepts.In this study, yttrium dihydride pellets were fabricated using powder metallurgical methods. Pellets were then analyzed for thermal diffusivity and heat capacity to determine thermal conductivity. Results will relate the thermophysical properties of the sintered pellets with those of hydrided monoliths in literature.