Metal waste forms from the electrometallurgical treatment of spent nuclear fuel

Stainless steel-zirconium alloys are being developed for the disposal of radioactive metal isotopes isolated using an electrometallurgical treatment technique to treat spent nuclear fuel. The nominal waste forms are stainless steel-15 wt% zirconium alloy and zirconium-8 wt% stainless steel alloy. These alloys are generated in yttria crucibles by melting the starting materials at 1,600 C under an argon atmosphere. This paper discusses the microstructures, corrosion and mechanical test results, and thermophysical properties of the metal waste form alloys

Thoria-based cermet nuclear fuel : sintered microsphere fabrication by spray drying.

Cermet nuclear fuels have been demonstrated to have significant potential to enhance fuel performance because of low internal fuel temperatures and low stored energy. The combination of these benefits with the inherent proliferation resistance, high burnup capability, and favorable neutronic properties of the thorium fuel cycle produces intriguing options for advanced nuclear fuel cycles. This paper describes aspects of a Nuclear Energy Research Initiative (NERI) project with two primary goals: (1) Evaluate the feasibility of implementing the thorium fuel cycle in existing or advanced reactors using a zirconium-matrix cermet fuel, and (2) Develop enabling technologies required for the economic application of this new fuel form. Spray drying is a physical process of granulating fine powders that is used widely in the chemical, pharmaceutical, ceramic, and food industries. It is generally used to produce flowable fine powders. Occasionally it is used to fabricate sintered bodies like cemented carbides, but it has not, heretofore, been used to produce sintered microspheres. As a physical process, it can be adapted to many powder types and mixtures and thus, has appeal for nuclear fuels and waste forms of various compositions. It also permits easy recycling of process ''wastes'' and minimal chemical waste streams that can arise in chemical sol/gel processing. On the other hand, for radioactive powders, it presents safety challenges for processing these materials in powder form and in achieving microspheres of high density and perfection

Characterization of Irradiated Metal Waste from the Pyrometallurgical Treatment of Used EBR-II Fuel

As part of the pyrometallurgical treatment of used Experimental Breeder Reactor-II fuel, a metal waste stream is generated consisting primarily of cladding hulls laden with fission products noble to the electrorefining process. Consolidation by melting at high temperature [1873 K (1600 degrees C)] has been developed to sequester the noble metal fission products (Zr, Mo, Tc, Ru, Rh, Te, and Pd) which remain in the iron-based cladding hulls. Zirconium from the uranium fuel alloy (U-10Zr) is also deposited on the hulls and forms Fe-Zr intermetallics which incorporate the noble metals as well as residual actinides during processing. Hence, Zr has been chosen as the primary indicator for consistency of the metal waste. Recently, the first production-scale metal waste ingot was generated and sampled to monitor Zr content for Fe-Zr intermetallic phase formation and validation of processing conditions. Chemical assay of the metal waste ingot revealed a homogeneous distribution of the noble metal fission products as well as the primary fuel constituents U and Zr. Microstructural characterization of the ingot confirmed the immobilization of the noble metals in the Fe-Zr intermetallic phase

Hot-isostatic pressing of U-10Zr by grain boundary diffusion and creep cavitation. Part 2: Theory and data analysis

Uranium-10 wt % zirconium (U-10Zr) is a fuel alloy that has been used in the Experimental Breeder Reactor-II (EBR-II). The high burnup that was desired in this fuel system made high demands on the mechanical compatibility between fuel and cladding both during normal operation and during safety-related transients when rapid differential expansion may cause high stresses. In general, this mechanical stress can be reduced by cladding deformation if the cladding is sufficiently ductile at high burnup, and/or by fuel hot-pressing. Fortunately, the fuel is very porous when it contacts the cladding, but this porosity gradually fills with solid fission products (primarily lanthanides) that may limit the fuel`s compressibility. If the porosity remains open, gaseous fission products are released and the porous fuel creeps rather than hot-presses under contact stresses. If the pores are closed by sintering or by solid fission products, the porous fuel will hot-isostatic press (HIP), as represented by the models to be discussed. HIP experiments performed at 700 C on U-10Zr samples with different impurity phase contents (Part 1) are analyzed in terms of several creep cavitation models. The coupled diffusion/creep cavitation model of Chen and Argon shows good quantitative agreement with measured HIP rates for hydride- and metal-derived U-10Zr materials, assuming that pores are uniformly distributed on grain boundaries and are of modal size, and that far-field strain rates are negligible. The analysis predicts, for the first time, an asymmetry between HIP and swelling at identical pressure-induced driving forces due to differences in grain boundary stresses. The differences in compressibility of hydride- and metal-derived U-10Zr can be partially explained by differences in pore size and spacing. The relevance of the experiments to description of in-reactor densification under external pressure or contact stress due to fuel/cladding mechanical interaction is discussed

Decomposition of the ?? phase in as-cast and quenched U-Zr alloys

An investigation of the decomposition of the high temperature gamma phase in as-cast and quenched U-Zr alloys was conducted. Differential scanning calorimetry data clearly showed delta reversible arrow g transformations in alloys with <10 wt% Zr while XRD data did not contain any peaks which uniquely identify it's presence. Since delta phase forms via omega transformation, a comparison of the theoretical diffraction patterns for omega and delta revealed that the intensities of the peaks which uniquely identify the existence of delta when alpha-U is present, were either very weak, or were zero in omega, suggesting that the ambiguity can be explained if the phase present in these alloys is omega as opposed to delta. Our data are consistent with the presence of delta and omega in as-cast and quenched U-50Zr alloy, respectively, and (alpha+omega) in rest of the as-cast and quenched alloys. Based on the experimental data, the transformation sequence from gamma phase in U-Zr alloys is proposed.clos

Thoria-based cermet nuclear fuel : neutronics fuel design and fuel cycle analysis.

Cermet nuclear fuel has been demonstrated to have significant potential to enhance fuel performance because of low internal fuel temperatures and low stored energy. The combination of these benefits with the inherent proliferation resistance, high burnup capability, and favorable neutronic properties of the thorium fuel cycle produces intriguing options for advanced nuclear fuel cycles. This paper describes aspects of a Nuclear Energy Research Initiative (NERI) project with two primary goals: (1) Evaluate the feasibility of implementing the thorium fuel cycle in existing or advanced reactors using a zirconium-matrix cermet fuel, and (2) Develop enabling technologies required for the economic application of this new fuel form. This paper will first describes the fuel thermal performance model developed for the analysis of dispersion metal matrix fuels. The model is then applied to the design and analysis of thorium/uranium/zirconium metal-matrix fuel pins for light-water reactors using neutronic simulation methods

Thoria-based cermet nuclear fuel : cermet fabrication and behavior estimates.

Cermet nuclear fuels have been demonstrated to have significant potential to enhance fuel performance because of low internal fuel temperatures and low stored energy. The combination of these benefits with the inherent proliferation resistance, high burnup capability, and favorable neutronic properties of the thorium fuel cycle produces intriguing options for advanced nuclear fuel cycles. This paper describes aspects of a Nuclear Energy Research Initiative (NERI) project with two primary goals: (1) evaluate the feasibility of implementing the thorium fuel cycle in existing or advanced reactors using a zirconium-matrix cermet fuel, and (2) develop enabling technologies required for the economic application of this new fuel form. Critical elements in the demonstration of this new fuel form include developing low-cost fabrication methods and characterizing the cermet properties and important performance parameters. A powder-in-tube drawing and heat treatment process is being evaluated as an alternative to hot extrusion. In this method, zirconium metal and ceramic microspheres are mixed, poured into a Zircaloy shell, and compacted into simulated fuel pins. Important processing variables being evaluated include the amount of compaction required to achieve a desired matrix density and the inter-drawing thermal treatment temperature required to achieve adequate matrix fusion and grain growth