Ceramics for a New Generation of Nuclear Energy Systems and Applications: Ceramic Nuclear Fuel
Sponsored by: TMS Nuclear Materials Committee, ACerS Energy Materials and Systems (EMSD) Division
Program Organizers: Ming Tang, Clemson University; Enrique Martinez Saez, Clemson University; Yongfeng Zhang, University of Wisconsin; Krista Carlson, University of Nevada, Reno; Yutai Katoh, Oak Ridge National Laboratory; Jean Paul Crocombette, CEA Saclay; Erofili Kardoulaki, Los Alamos National Laboratory; Levi Gardner, Argonne National Laboratory; Jian Zhang, Xiamen University; Charmayne Lonergan, Missouri University of Science and Technology
Monday 2:00 PM
October 10, 2022
Room: 402
Location: David L. Lawrence Convention Center
Session Chair: Lingfeng He, Idaho National Laboratory; Kathy Lu, University of Alabama Birmingham
2:00 PM Invited
Cluster Dynamics Simulations of Point Defects and Fission Gas Evolution in Irradiated Ceramic Nuclear Fuels: David Andersson1; Christopher Matthews1; Jason Rizk1; Romain Perriot1; Michael Cooper1; Benjamin Liu1; Christopher Stanek1; 1Los Alamos National Laboratory
Point defect evolution under irradiation influences important nuclear fuel performance characteristics such as fission gas evolution and dislocation growth. The Centipede cluster dynamics code, based on the MOOSE framework, was developed to simulate these processes, specifically targeting the known sensitivity to oxygen non-stoichiometry and complex defect clusters in UO2 based fuels, which has since been extended to U3Si2 and UN fuels. The cluster dynamics models require a large set of parameters as input, which are derived from atomic scale simulations. The results from the Centipede cluster dynamics simulations are analyzed in terms of parameter uncertainty and the need for limited parameter calibration will be discussed. Finally, application of defect and fission gas diffusion models in engineering scale fuel performance analysis is discussed.
2:30 PM Invited
Microstructural Evolution in Ceramic Nuclear Fuels and their Surrogates under Irradiation: Lingfeng He1; Kaustubh Bawane1; Pengyuan Xiu1; Tiankai Yao1; Chao Jiang1; Marat Khafizov2; Miaomiao Jin3; Yi Xie4; Lin Shao5; 1Idaho National Laboratory; 2The Ohio State University; 3The Pennsylvania State University; 4Purdue University; 5Texas A&M University
Oxide nuclear fuels have been widely used in light water reactors (LWRs) and nitride nuclear fuels are proposed as accident tolerant fuels for LWRs or candidates for advanced reactors. In reactor environments, radiation induced microstructure changes in ceramic nuclear fuels can affect their thermal conductivity and mechanical properties. Investigating early-stage microstructural changes is of significance in understanding the performance degradation of ceramic nuclear fuels in reactor environments. In this work, we study the microstructural evolution as a function of temperature and irradiation dose in oxide and nitride nuclear fuels and their surrogates using a combination of in situ/ex situ ion irradiation, advanced characterization, and modeling. The irradiation induced dislocation loops and phase changes are characterized using electron microscopy techniques. Loop density and diameter are analyzed using a kinetic rate theory that considers stoichiometric loop evolution. The energetics of dislocation loop types and phase relationships are studied using multiscale modeling.
3:00 PM
Fabrication and Properties of Sintered Yttrium Hydride: 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. One proposed microreactor design to integrate with microgrids uses yttrium hydride as the neutron moderator due to its high hydrogen density and thermal stability, enabling the operation of thermal spectrum reactors to high temperatures and, thus, enhancing fuel utilization and cost-effectiveness while keeping the core transportable. However, it is difficult to produce yttrium hydride in geometries required for reactor design concepts.In this study, yttrium hydride pellets were fabricated using powder metallurgical methods. Thermogravimetric analysis was used to determine mass change as a function of temperature and sintering environment. Pellets were then analyzed for elastic moduli, thermal diffusivity, coefficient of thermal expansion, and heat capacity. Results will relate the properties of the sintered pellets with those of hydrided monoliths in literature.
3:20 PM Break
3:40 PM Invited
SiC Oxidation and Irradiation Resistance in Advanced Nuclear Reactor TRISO Fuel: Kathy Lu1; Yi Je Cho2; 1Virginia Polytechnic Institute and State University; 2Virginia Polytechnic Institute and State University; Sunchon National University
Under accidental conditions for high temperature gas-cooled reactors (HTGR), the SiC layer in tri-structural-isotropic (TRISO) fuel particles can be exposed to water vapor. In this study, oxidation behaviors of surrogate TRISO fuel particles were investigated in a He-20 vol% water vapor mixed atmosphere at temperatures up to 1600 °C. Volatilization of the oxide layer was analyzed using a mechanistic model. The prediction indicates that the oxidized SiC layer should retain fission products. In addition, microstructure and defect evolution in the oxidized SiC layer of surrogate TRISO fuel particles under ion irradiation were observed by in-situ transmission electron microscopy. The defect number density at 800 °C was an order of magnitude lower than that in the sample irradiated at room temperature. Also, crystalline SiO2 had higher radiation resistance compared to SiC. A defect reaction rate theory was utilized to understand the fundamental defect evolution process and irradiation resistance difference.
4:10 PM Invited
Integration of Nuclear Fuel and Embedded Sensors within Additively Manufactured SiC Components: Christian Petrie1; 1Oak Ridge National Laboratory
The adoption of silicon carbide (SiC) ceramic components within nuclear reactor designs has historically been limited due to challenges in manufacturing components with complex geometries. Oak Ridge National Laboratory has recently demonstrated an additive manufacturing (AM) process to overcome these challenges using binder jet 3D printing of SiC components that are densified using chemical vapor infiltration. Leveraging this process, the Transformational Challenge Reactor (TCR) fuel form was conceptualized to include a dense packing of conventionally fabricated uranium nitride tristructural isotropic fuel particles within an AM SiC matrix. This same process is used to embed sensors at critical locations within 3D printed SiC fuel forms or structural components. This presentation will provide an overview of the SiC AM and sensor embedding processes, the challenges and opportunities for designing, fabricating, and instrumenting TCR-like fuels, and the results from initial irradiation testing of integral fuel compacts.
4:40 PM
Corrosion of SiC in Molten Salt and Liquid Lead: Huali Wu1; Jinsuo Zhang1; 1Virginia Polytechnic Institute and State University
Ceramic composites based on silicon carbide fibers with a silicon carbide matrix (SiC-SiC) have been studied for use in advanced nuclear reactors such as molten salt reactors and lead-cooled fast reactors and show very promising results in both corrosion and radiation tolerance. The present study tested SiC corrosion by liquid lead and U-bearing molten fluoride salt. All the tests were conducted at 700 oC in the static liquid for 120 hours. The post-test samples were characterized by SEM and XRD. The characterization was focus on the Si depletion layer, the oxidation layer, and the liquid infiltration. Different SiC samples with different fabrication methods were tested, and results showed that fabrication method may influence the corrosion resistance of SiC to both molten salt and liquid lead.