Oxybelis aeneus

Number of Pages: 4Integrative BiologyGeological Science

Chemical Interactions Between Metallic SFR Fuel and Advanced Claddings

Advanced cladding materials are being considered for high-temperature applications in Generation IV reactors (e.g. T91, ODS steels). These materials provide better mechanical properties (e.g., creep strength) at these higher temperatures than the more commonly employed cladding steels (e.g., HT-9). However, consideration must also be given to other high-temperature phenomenon besides just the mechanical properties of the cladding when using fuels at high temperatures. In particular, the compatibility of the fuel and cladding must be considered. In Sodium Fast Reactors (SFRs) that employ metallic fuel, interactions of fuel and cladding (FCI) during irradiation of a fuel element can result in the formation of strength-reducing zones or possibly the formation of low melting phases [1]. This paper reports the results of diffusion experiments that were conducted with metallic fuel and advanced cladding materials at the relatively high temperature of 700 C

Observations Derived From the Characterization of Monolithic Fuel Plates Irradiated as Part of the RERTR-6 Experiment

Evaluation of the PIE results of the monolithic plates that were irradiated as part of the RERTR-6 experiment has continued. Specifically, comparisons have been made between the microstructures of the fuel plates before and after irradiation. Using the results from the rigorous characterization that was performed on the as-fabricated plates using scanning electron microscopy, it is possible to improve understanding of how monolithic fuel plates perform when they are irradiated. This paper will discuss the changes that occur, if any, in the microstructure of a monolithic fuel plate that is fabricated using techniques like what were employed for fabricating RERTR-6 fuel plates. In addition, the performance of fuel/cladding interaction layers that were present in the fuel plates due to the fabrication process will be discussed, particularly in the context of swelling of these layers and how these layers exhibit different behaviors depending on whether the fuel alloy in the fuel plate is U-7Mo or U-10Mo

Immobilization of Technetium in a Metallic Waste Form

Fission-product technetium accumulated during treatment of spent nuclear fuel will ultimately be disposed of in a geological repository. The exact form of Tc for disposal has yet to be determined; however, a reasonable solution is to incorporate elemental Tc into a metallic waste form similar to the waste form produced during the pyrochemical treatment of spent, sodium-bonded fuel. This metal waste form, produced at the Idaho National Laboratory, has undergone extensive qualification examination and testing for acceptance to the Yucca Mountain geological repository. It is from this extensive qualification effort that the behavior of Tc and other fission products in the waste form has been elucidated, and that the metal waste form is extremely robust in the retention of fission products, such as Tc, in repository like conditions. This manuscript will describe the metal waste form, the behavior of Tc in the waste form; and current research aimed at determining the maximum possible loading of Tc into the metal waste and subsequent determination of the performance of high Tc loaded metal waste forms

SEM and TEM Characterization of As-Fabricated U-7Mo Disperson Fuel Plates

The starting microstructure of a dispersion fuel plate can have a dramatic impact on the overall performance of the plate during irradiation. To improve the understanding of the as-fabricated microstructures of dispersion fuel plates, SEM and TEM analysis have been performed on RERTR-9A archive fuel plates, which went through an additional hot isostatic procsssing (HIP) step during fabrication. The fuel plates had depleted U-7Mo fuel particles dispersed in either Al-2Si or 4043 Al alloy matrix. For the characterized samples, it was observed that a large fraction of the ?-phase U-7Mo alloy particles had decomposed during fabrication, and in areas near the fuel/matrix interface where the transformation products were present significant fuel/matrix interaction had occurred. Relatively thin Si-rich interaction layers were also observed around the U-7Mo particles. In the thick interaction layers, (U)(Al,Si)3 and U6Mo4Al43 were identified, and in the thin interaction layers U(Al,Si)3, U3Si3Al2, U3Si5, and USi1.88-type phases were observed. The U3Si3Al2 phase contained some Mo. Based on the results of this work, exposure of dispersion fuel plates to relatively high temperatures during fabrication impacts the overall microstructure, particularly the nature of the interaction layers around the fuel particles. The time and temperature of fabrication should be carefully controlled in order to produce the most uniform Si-rich layers around the U-7Mo particles

Results of Recent Microstructural Characterization of Irradiated U-Mo Dispersion Fuels with Al Alloy Matrices that Contain Si

RERTR U-Mo dispersion fuel plates are being developed for application in research reactors throughout the world. As part of this development, reactor experiments are being conducted in the Advanced Test Reactor to determine the irradiation performance of different dispersion fuels that contain U-Mo alloys with different Mo contents and Al alloy matrices with different Si contents. Of particular interest is the performance of the dispersion fuels depending on the Si content of the Al alloy matrix, since the addition of Si is being looked to for improving the performance of these dispersion fuels. This paper will describe the results of recent microstructural examinations that have been performed using optical metallography and scanning electron microscopy on as-fabricated and as-irradiated dispersion fuels with different amounts of Si added to the Al matrix. Differences in the microstructural development during irradiation as a function of the Si content in the Al matrix will be discussed, and comments will be made about the development and stability of the fuel/matrix interaction layers that are commonly present in irradiated dispersion fuels

Stainless steel-zirconium alloy waste forms

An electrometallurgical treatment process has been developed by Argonne National Laboratory to convert various types of spent nuclear fuels into stable storage forms and waste forms for repository disposal. The first application of this process will be to treat spent fuel alloys from the Experimental Breeder Reactor-II. Three distinct product streams emanate from the electrorefining process: (1) refined uranium; (2) fission products and actinides extracted from the electrolyte salt that are processed into a mineral waste form; and (3) metallic wastes left behind at the completion of the electrorefining step. The third product stream (i.e., the metal waste stream) is the subject of this paper. The metal waste stream contains components of the chopped spent fuel that are unaffected by the electrorefining process because of their electrochemically ``noble`` nature; this includes the cladding hulls, noble metal fission products (NMFP), and, in specific cases, zirconium from metal fuel alloys. The selected method for the consolidation and stabilization of the metal waste stream is melting and casting into a uniform, corrosion-resistant alloy. The waste form casting process will be carried out in a controlled-atmosphere furnace at high temperatures with a molten salt flux. Spent fuels with both stainless steel and Zircaloy cladding are being evaluated for treatment; thus, stainless steel-rich and Zircaloy-rich waste forms are being developed. Although the primary disposition option for the actinides is the mineral waste form, the concept of incorporating the TRU-bearing product into the metal waste form has enough potential to warrant investigation

Radiation Effect on Microstructural Stability of RERTR Fuel

Three depleted uranium alloys are successfully cast for the radiation stability studies of the fuel-cladding interaction product using proton irradiation. SEM analysis indicates the presence of the phases of interest: U(Si,Al)3, (U,Mo)(Si,Al)3 and a mixture of UMo2Al20, U6Mo4Al43, and UAl4. Irradiation with 2.6 MeV protons at 200ºC to the doses of 0.1, 1.0 and 3.0 dpa are carried out

Analysis of irradiated U-7wt%Mo dispersion fuel microstructures using automated image processing

The High Performance Research Reactor Fuel Development (HPPRFD) program is responsible for developing low enriched uranium (LEU) fuel substitutes for high performance reactors fueled with highly enriched uranium (HEU) that have not yet been converted to LEU. The uranium-molybdenum (U-Mo) fuel system was selected for this effort. In this study, fission gas pore segmentation was performed on U-7wt%Mo dispersion fuel samples at three separate fission densities using an automated image processing interface developed in MATLAB. Pore size distributions were attained that showed both expected and unexpected fission gas behavior. In general, it proved challenging to identify any dominant trends when comparing fission bubble data across samples from different fuel plates due to varying compositions and fabrication techniques. The results exhibited fair agreement with the fission density vs. porosity correlation developed by the Russian reactor conversion program.Keywords: Automated image analysis, Fission bubbles, Porosity, Nuclear fuel, MATLAB, Fission densit