Materials and Fuels for the Current and Advanced Nuclear Reactors VI: Fuels III
Sponsored by: TMS Structural Materials Division, TMS: Corrosion and Environmental Effects Committee, TMS: Nuclear Materials Committee, TMS: Mechanical Behavior of Materials Committee
Program Organizers: Ramprashad Prabhakaran, Pacific Northwest National Laboratory; Dennis Keiser, Idaho National Laboratory; Raul Rebak, GE Global Research; Clarissa Yablinsky, Los Alamos National Laboratory
Tuesday 8:30 AM
February 28, 2017
Location: Marriott Marquis Hotel
Session Chair: Kurt Terrani, Oak Ridge National Laboratory; Isabella van Rooyen, Idaho National Laboratory
Production of Fully Ceramic Microencapsulated Fuel for Test Reactor Irradiation: Kurt Terrani1; James Kiggans1; Michael Trammell1; Wilson Cowherd2; Gregory Core3; 1Oak Ridge National Laboratory; 2Idaho National Laboratory ; 3Idaho National Laboratory
The fully ceramic microencapsulated fuel (FCM) consists of tri-structural isotropic (TRISO) fuel particles embedded inside a dense SiC matrix. Utilization of this fuel in light water reactors (LWRs) as an accident tolerant fuel form has the potential to greatly improve safety given multiple inherent barriers to fission product release and the high-temperature steam oxidation resistance of the SiC matrix. This specific manifestation of this fuel form differs in two ways from the well-established high temperature gas reactor (HTGR) fuel: the fuel kernel consists of a high density uranium nitride-carbide solid solution instead of oxide or oxide carbide two phase mixture. The SiC matrix of FCM fuel replaces the graphite matrix used in HTGR compacts. This study reports the various steps in the LEU (7.4% enriched) fuel production campaign at ORNL to provide feedstock for first irradiation testing of FCM fuel concept at INL’s Advanced Test Reactor.
Microstructural Characterization and Thermal Properties of Metallic Pu-Zr Systems: Assel Aitkaliyeva1; Cynthia Papesch1; 1Idaho National Laboratory
This contribution reports the results of transmission electron microscopy (TEM) investigation of the microstructure and phases formed in the Pu-Zr systems since the complete understanding of the irradiation behavior of transmutation fuels cannot be achieved without this fundamental knowledge. The high spatial resolution of TEM allowed characterization of individual matrix phases and selective area diffraction analysis provided important space-group information and lattice parameters of the phases formed in Pu-Zr system. Microstructural characterization is linked to thermal properties analysis. Transition temperatures, enthalpies of transition, and heat capacities will be determined employing a Differential Scanning Calorimeter (DSC). Thermal diffusivities and conductivities of the fuels will be determined using laser flash analyzers (LFA). Phase transitions observed during DSC measurements are correlated with the phases observed in TEM. This work was supported by the LDRD program at INL and FCRD program of US Department of Energy.
Post Irradiation Electron Microscopy Examination of UCO Fuel Kernels from TRISO Coated Particles: Terry Holesinger1; Isabella van Rooyen2; Weicheng Zhong2; 1Los Alamos National Laboratory; 2Idaho National Laboratory
High resolution electron microscopy investigations of selected coated particles from the first advanced gas reactor experiment (AGR-1) at Idaho National Laboratory have provided important information on the evolution of the UCO fuel kernel during irradiation. TRISO coated particles from AGR-1 were irradiated to an average burn-up of 15.3% fissions per initial metal atom, time-averaged, volume-averaged temperature of 1092°C, a time-averaged, peak temperature of 1166°C, and an average fast fluence of 3.22x1025 n/cm2. Resulting kernel phases and fission products accumulations present in the kernel and adjoining buffer layer after irradiation were identified. For particle AGR1-131-066 it was found that the U-C phase contained Zr, Mo, and/or Ru with trace amounts of Nd and Pr found in the oxide phase.
Preliminary Post Irradiation Examination SEM Analysis of AGR 2 UO2 and UCO TRISO Fuel Particles: Tyler Gerczak1; John Hunn1; Charles Baldwin1; Robert Morris1; Fred Montgomery1; 1Oak Ridge National Laboratory
Compacts containing tristructural isotropic (TRISO)-coated kernels of uranium carbide/oxide (UCO) or uranium oxide (UO2) have been irradiated as part of the US Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. Preliminary scanning electron microscopy and energy dispersive x-ray spectroscopy results of irradiated UCO-TRISO and UO2-TRISO AGR‑2 particle cross‑sections will be presented. The analysis includes deconsolidated particles from various compacts which experienced a range in irradiation conditions and post‑irradiation safety testing temperatures. Deconsolidated particles for analysis were selected based on their fission product inventory. Particles with standard fission product inventories of well retained isotopes and a range in fission product inventories of isotopes with variable release behaviors (e.g. Ag-110 and Eu‑154) were specifically targeted for analysis. Analysis of these particles allowed for correlation of fission product retention behavior with TRISO layer integrity and fission product distributions in the TRISO layer components.
Grain Boundary Complexions in SiC and Their Relevance in Silver Diffusion in TRISO Particles: Felix Cancino Trejo1; Eddie Lopez-Honorato1; 1CINVESTAV
The TRISO fuel particles is made of a uranium kernel coated with three layers of pyrolytic carbon and one of silicon carbide. After more than forty years of development one of the main challenges that remain unsolved on the development of this fuel particle is the diffusion and release of the fission product silver. In this work we propose that the existence of grain boundary complexions in SiC is responsible for the complex diffusion observed for silver. The description of grain boundary diffusion through the existence of these complexions could lead to a general explanation of the diffusion of silver through SiC in TRISO fuel particles, as it can integrate the effect of SiC microstructure and composition, temperature, irradiation and presence of secondary elements such as Pd and Si.
10:10 AM Break
On Silver Transport in 3C-SiC: Johannes Neethling1; Jacques O'Connel1; 1Nelson Mandela Metropolitan University
Current high temperature gas reactor designs use TRISO-coated particles as fuel. The TRISO-coated particle consists of a fuel kernel and layers of porous pyrolytic carbon and silicon carbide (SiC) which serves as the main barrier to fission product release. However, the release of radioactive silver (Ag) from intact TRISO-coated particles has been observed which indicates that silver can migrate through silicon carbide. Numerous earlier investigations of the migration of Ag in 3C-SiC did not reveal any significant Ag movement. By using high resolution transmission electron microscopy techniques, the authors and co-workers identified a palladium assisted silver transport mechanism in SiC as the possible main mechanism leading to the release of Ag by intact TRISO particles. New results on the temperature dependence of the Pd-Ag transport rate in neutron irradiated SiC will be presented and compared to earlier diffusion coefficients derived from fractional release data of fission products from TRISO particles.
High Temperature Fuel Cladding Chemical Interactions between Unirradiated TRIGA Fuels and 304 Stainless Steel: Emmanuel Perez1; Bryan Forsmann2; Jatuporn Burns2; Michael Heighes1; Dawn Janney1; Dennis Keiser1; Steven Cook3; Jody Henley1; Eric Woolstenhulme1; 1Idaho National Laboratory; 2Boise State University; 3Center for Space Nuclear Research
High temperature fuel cladding chemical interactions (FCCI) between Training-Research, Isotopes-General-Atomics (TRIGA) fuel elements and their 304 stainless steel (304SS) cladding are of interest to develop an understanding of the fuel-behavior during transient reactor scenarios. TRIGA fuels are composed of uranium particles dispersed in a zirconium hydride (Zr-H) matrix encased in a 304SS cladding. At high temperatures, the fuel and cladding have the potential experience FCCI. A large number of interactions can take place in this multi-component fuel system. Emphasis is given to the development of low melting point phases. Diffusion couples of TRIGA fuel vs. 304SS were annealed at 1000°C for 5 and 24 hours. Characterization was then carried out via scanning electron microscopy and transmission electron microscopy. The as-received and annealed microstructures of the fuel and the interactions regions between the fuel and the 304SS will be discussed.
Small Scale Mechanical Testing of UO2 at Elevated Temperatures: David Frazer1; Benjamin Shaffer2; Kitt Roney2; Harn Lim2; Perdo Peralta2; Peter Hosemann1; 1University of California, Berkeley; 2Arizona State University
Mechanical properties of UO2 are of great importance to the nuclear engineering community for modeling of oxide fuel performance during operation and accident scenarios. There exists a need for precise data describing the deformation of single crystals at both room and elevated temperature. Small-scale testing techniques like microcantilever bending and nanoindentation have the ability to measure mechanical properties in-situ allowing observation of the deformation of the material. These techniques can also be used to study mechanical behavior of UO2 at elevated temperatures, which can then be subsequently investigated with other more conventional techniques. In our work we perform microcantilever and nanoindentation testing on UO2 at room and elevated temperatures, with emphasis on creep behavior, with subsequent TEM investigation to evaluate the deformation mechanisms. In addition we utilize an innovative technique to study the effect of helium bubbles on the UO2 deformation and explore in-situ TEM microcantilever and elevated temperature testing.
Model of Thermal Conductivity Reduction Due to Point Defect Accumulation in Ion Irradiated UO2: M Faisal Riyad1; Vinay Chauhan1; Yuzhou Wang1; Marat Khafizov1; 1The Ohio State University
We present a model to interpret thermal conductivity reduction in light ion irradiated uranium dioxide (UO2). In UO2, thermal conductivity is governed by phonons. A low dose and low temperature light ion irradiation is most likely to induce only point defects. These point defects act as strong scattering centers for phonons and limit UO2’s ability to conduct heat. Point defect’s strength to scatter phonon is defined by an effective mass difference and strain due to the ionic radius mismatch. Defect concentration is estimated based on rate theory equations. Structures of the defects suitable for implementation in the classical model of phonon mediated thermal transport are based on the atomic level defect simulations in UO2 reported in the literature. We compare the predictions of model to the results of recent reports on conductivity reduction and lattice expansion in light ion irradiated UO2.