Materials and Fuels for the Current and Advanced Nuclear Reactors VI: Fuels II
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
Monday 2:00 PM
February 27, 2017
Location: Marriott Marquis Hotel
Session Chair: Yongho Sohn, University of Central Florida; Vineet Joshi, Pacific Northwest National Laboratory
Reduced Modulus and Hardness of Uranium-molybdenum Solid Solution as a Function of Mo Composition and Related Phase Transformations: Ryan Newell1; Youngjoo Park1; Abhihek Mehta1; Dennis Keiser2; Yongho Sohn1; 1University of Central Florida; 2Idaho National Laboratory
The Materials Management and Minimization program is a global initiative to reduce the use of highly enriched uranium in research and test nuclear reactors. The critical parameters of thermo-mechanical fabrication and performance of these low enriched fuels depend strongly on the mechanical properties of the fuel system constituents, wherein the U-10wt.%Mo is the fuel alloy. Through nano-indentation testing, hardness and reduced modulus values were systematically investigated for varying molybdenum composition created by several solid-to-solid diffusion couples in the uranium-molybdenum system. A local minimum in reduced modulus and hardness with respect to Mo composition was observed near 11 at. % molybdenum. Variation in reduced modulus and hardness are presented and discussed with respect to the composition-dependent destabilization of the high temperature body centered cubic γ-phase, and composition-dependent mechanical behavior within the phases observed.
Interdiffusion and Reaction between U and Zr: Youngjoo Park1; Ryan Newell1; Abhishek Mehta1; Dennis Keiser2; Yongho Sohn1; 1University of Central Florida; 2Idaho National Laboratory
The microstructural development and diffusion kinetics in solid-to-solid, U vs. Zr diffusion couples were examined at 580, 650, 680 and 710░C by SEM/XEDS. The interdiffusion and reaction layer consisted of α-U (oC4) containing Zr acicular precipitate, α’ (oC4-variant) and (γU,βZr) (cI2) solid solution at 650, 680 and 710░C. The δ-UZr2 phase, instead of (γU,βZr) solid solution phase, was observed in the couple annealed at 580░C. The interdiffusion fluxes and coefficients were determined using both Sauer-Freise and Boltzmann-Matano analyses, interdiffusion coefficients for the α-U, (γU,βZr) and δ-UZr2 (580░C only). For the α’-phase with negligible concentration gradient, integrated interdiffusion coefficients were determined via Wagner method. Marker plane was found in (γU,βZr) (cI2) solid solution at 650, 680 and 710░C and δ-UZr2 at 580░C, and intrinsic diffusion coefficients were calculated: U intrinsically diffused an order magnitude faster than Zr. Consistency in results were examined by Arrhenius temperature-dependence and comparison to existing literature.
Microstructural Analysis of Electrochemically Formed Zirconium Coatings for Uranium-Molybdenum Nuclear Fuels: Alexander Smirnov1; John Scott O'Dell1; 1Plasma Processes LLC
In the Global Threat Reduction Initiative Conversion Program, there is a need to develop alternative techniques for the zirconium coating of uranium-molybdenum nuclear fuels, which produce less scrap than the baseline cladding method. One of the leading alternative coating methods is electrochemical forming (EL-Form«).The base of EL-Form« processes is the electrolysis of molten salts.Recently, the feasibility of EL-Form« processing has been demonstrated by the production of zirconium coatings on uranium-molybdenum surrogate materials such as molybdenum, hafnium, and zirconium substrates as well as depleted uranium-molybdenum substrates. Analysis of the EL-Form« zirconium deposits has shown dense, well bonded coatings can be produced on a verity of different substrate materials. This paper will detail the characterization of the EL-Form« zirconium coatings on different substrates and the effect of post-deposition heat treatments on microstructure and chemical composition.
Sensitivity Analysis on the Temperature of U–Mo/Al Plate-type Dispersion Fuel: Faris B. Sweidan1; Jeong Sik Yim2; Ho Jin Ryu1; 1Korea Advanced Institute of Science and Technology; 2Korea Atomic Energy Research Institute (KAERI)
U-Mo/Al plate-type dispersion fuel is a promising candidate for the conversion of research reactors from HEU to LEU fuels due to its high density and good irradiation stability. Uncertainties of the critical parameters have a significant impact on the probable fuel temperature ranges since fuel performance, represented by swelling, fission gas release, and interaction layers formation, is affected by fuel temperature. In this paper, the uncertainty ranges of the reactor operation conditions, fuel fabrication, fuel properties, and the dynamic changes of fuel during irradiation, such as the thermal conductivity of irradiated fuel, oxide layer thickness and pH value uncertainties, are used to determine the probable fuel temperature ranges. The combined uncertainty effect of these parameters on the fuel temperature range is also determined using the propagation of uncertainty and probabilistic sensitivity analysis (Monte Carlo simulation). These sensitivity analyses should provide technical advantages in the fuel performance evaluation and safety analyses.
3:20 PM Cancelled
Characterization of Metallic Ffuel Slugs Fabricated by Injection Casting: Jeong-Yong Park1; Jong-Hwan Kim1; Ki-Hwan Kim1; Hoon Song1; Jung-Won Lee1; Seok-Jin Oh1; Seoung-Woo Kuk1; Young-Mo Ko1; Yoon-Myung Woo1; Chan-Bock Lee1; 1Korea Atomic Energy Research Institute
U-Zr-RE(Nd-Ce-Pr-La) fuel slugs were fabricated by injection casting to establish the fabrication process of metallic fuels for sodium-cooled fast reactors. Density deviation in the fabricated fuel slugs was less than 0.3g/cm3. Precipitates were shown to be uniformly distributed over the fuel slug. The fuel weight loss measured after the injection casting was about 1.5%. The reaction between the melt and the crucible was found to be significant in the fabrication of RE-containing fuel slugs compared to U-Zr fuel slugs. U-Zr-Mn fuel slugs were fabricated as a surrogate for MA(minor actinide)-bearing metallic fuels by the injection casting method. Three different pressure conditions (vacuum, 400 torr Ar and 600 torr Ar) were applied during the melting process. Mn was found to be evaporated up to 68% in the vacuum injection casing whereas no evaporation of Mn was detected when melted under Ar atmosphere.
3:40 PM Break
Characterization of Nuclear Fuels by Neutron Diffraction and Energy-resolved Neutron Imaging: Sven Vogel1; 1Los Alamos National Laboratory
The unique advantages of neutrons for characterization of nuclear fuel materials are applied to accelerate the development and ultimately licensing of new nuclear fuel forms. Neutrons allow to characterize the crystallography of phases consisting of heavy elements (e.g. uranium) and light elements (e.g. oxygen, nitrogen, or silicon). The penetration ability in combination with comparably large (e.g. cm sized) beam spots provide microstructural characterization of typical fuel geometries for phase composition, strains, and textures from neutron diffraction. In parallel, we are developing energy-resolved neutron imaging and tomography with which we can complement diffraction characterization. This unique approach not only allows to visualize cracks, arrangement of fuel pellets in rodlets etc., but also characterization of isotope or element densities by means of neutron absorption resonance analysis. In this presentation, we provide an overview of our recent accomplishments in fuel characterization for accident-tolerant fuel consisting of uranium nitride/uranium silicide composite fuels.
Microstructure Evolution during Spark Plasma Sintering of Nuclear Fuel Pellets and Their Large-scale Manufacturability: Ghatu Subhash1; James Tulenko1; 1University of Florida
Spark plasma sintering (SPS) has been used to fabricate UO2, UO2-Diamond and UO2-SiC composite nuclear fuels. The influence of process parameters and their relationship to microstructure evolution are investigated. Master sintering curves (MSC) were developed in order to model and control the sintering process. The activation energy for UO2 and UO2-SiC composite were found to be 140 KJ/mol and 420 KJ/mol, respectively. The ability and accuracy of the constructed master sintering curves for density prediction and sintering profile design have also been demonstrated. The stability of 2nd phases (diamond and SiC) during sintering has been investigated and level of graphitization of diamond has been quantified. Finally, the ability of SPS process for large scale manufacture of hundreds of millions of pellets per day have been explored.
Fabrication and Characterization of TRISO Particles Using 800Ám Uranium Nitride and Surrogate ZrO2 Kernels: Brian Jolly1; Grant Helmreich1; Kevin Cooley1; John Dyer1; Kurt Terrani1; 1Oak Ridge National Laboratory
In support of fully ceramic microencapsulated (FCM) fuel development, coating development work is ongoing at the Oak Ridge National Laboratory (ORNL) to produce tri-structural isotropic (TRISO) coated fuel particles with both UN kernels and surrogate (uranium-free) kernels. The nitride kernels are used to increase fissile density in these SiC-matrix fuel pellets which have the potential to be an accident tolerant fuel form. The surrogate TRISO particles are necessary for separate effects testing and for utilization in consolidation process development. The advanced gas reactor (AGR) program at ORNL used fluidized bed chemical vapor deposition (FBCVD) techniques for TRISO coating of UCO (two phase mixture of UO2 and UCx) kernels. Similar techniques were employed for coating the UN and ZrO2 kernels, however significant changes in processing conditions were required to maintain acceptable coating properties due to physical property and dimensional differences between the previous 350Ám UCO and the current 800Ám UN/ZrO2 kernels.
Fission Product Electron Microscopy Analysis of Post Irradiated TRISO-coated Particles from the Second Advanced Gas Reactor Experiment: Clemente Parga1; Jeffery Aguiar1; Isabella van Rooyen1; 1Idaho National Laboratory
Correlative electron microscopy combining precession electron diffraction and point-resolved chemical imaging have been used to provide the material structure and chemical information on the evolution and performance of irradiated uranium oxy-carbide (14.0 wt% enriched) and uranium oxide (9.6 wt% enriched) kernelled tristructural isotropic-coated (TRISO) coated particle fuels. Particles from the Advanced Gas Reactor (AGR)-2 experiment from the Advanced Test Reactor Program at Idaho National Laboratory are compared before and after irradiation to explore differences resulting from different kernels and potential effects on fission product release through intact SiC layers. Particles from AGR-2 were irradiated to an average burn-up of 10.8% fissions per initial metal atom, time-averaged, volume-averaged temperature of 1199░C, a time-averaged, peak temperature of 1360░C, and an average fast peak fluence of 2.99x1025 n/cm2. The resulting microstructure, grain orientation distribution, grain boundary chemistry, and fission products present in the kernel, graphite, and SiC before and after irradiation were identified.