This paper reports on the developments for advanced fuel cycles to fully utilize resource materials (e.g., uranium) from used nuclear fuel and de-crease the amount and toxicity of waste requiring geological disposal by the U.S. Department of Energy.

The US High Performance Research Reactor (USHPRR) Conversion program is developing an Al-clad Zr-bonded U-10Mo nuclear fuel foil to enable the conversion of research reactors to high-assay low-enriched uranium (HALEU) alloy fuel. Fabrication scrap from the fuel manufacturing will be electrorefined to recover the HALEU, which will generate waste streams consisting of Zr, Mo, and residual U retained in the stainless steel anode basket used in the electrorefiner. Four alloys were formulated with different relative amounts of Zr, Mo, and 316L-SS to represent the expected range of waste compositions. Laboratory-scale ingots were made and metallurgically characterized to assess differences in the microstructures and distributions of the surrogate waste metals. Different amounts of the same dominant phases with similar compositions were formed in most materials: a γ-austenite based matrix, two ZrFe 2 intermetallic phases that could host residual uranium, an FeCrMo sigma (σ) intermetallic phase, a chi (χ) phase, and a secondary austenite (γ 2 ) phase. Electrochemical tests were conducted to compare the corrosion behaviors in acidic and alkaline brine solutions for a range of simulated environmental redox conditions. Corrosion of the Mo-rich γ and γ 2 solid solutions and the σ and χ intermetallic phases occurred. The multiphase alloy waste form formulated with the lowest molybdenum and highest zirconium contents showed the highest corrosion resistance and has sufficient amounts of ZrFe 2 to immobilize trace amounts of residual uranium in the waste stream. The alloy formulation strategy and testing approach will be presented with key results. Work conducted at Argonne National Laboratory is operated for the U.S. Department of Energy, Office of Science under contact DE-AC02-06CH11357.

Tree, and Piping and Instrumentation Diagram) and a utility module. This reference manual provides a screen-by-screen guide of the entire FEP System.

The electrochemical corrosion behaviors of two multiphase alloys representing waste forms made with 316 L stainless steel and different amounts of surrogate metallic fuel wastes were measured and related to the microstructures. Potentiodynamic (PD) scans were performed in an acid brine solution and the corroded surfaces were characterized with scanning electron microscopy (SEM) to compare the electrochemical responses to the corrosion of specific phases. PD scans for the two multiphase alloys, 316 L stainless steel, and pure palladium were compared to understand the complex corrosion behavior of these multiphase alloys also recently classified as multi principle element alloys (MPEAs) and to determine the effects of alloying elements and noble metals present in constituent phases on the corrosion behavior.

The corrosion rates of alloys that will be present in spent fuel waste packages are being measured to represent the generation rates of H2 in a breached package under a range of Eh-pH conditions. The H2 concentration is a crucial parameter in repository performance analyses because the presence of even small amounts of H2 in a breached waste container can significantly decrease the oxidative-dissolution rate of directly-disposed spent UO2 fuel. Potentiostatic tests are being conducted in pH 4, 7, and 10 buffer solutions spiked with NaCl at several potentials that span the ranges of redox and chemical conditions that could occur in a breached waste package. The corrosion currents are monitored as the alloys corrode and changes in the electrical properties of the metal surfaces are measured using electrochemical impedance spectroscopy. Results of tests with 316L stainless steel, Zircaloy-4, and 4320 low alloy carbon steel conducted at several voltages in pH 10 solutions with added NaCl are presented as examples. The dependencies of alloy corrosion rates on these variables and the sensitivity of the fuel dissolution rate are discussed.

In this report, issues that must be addressed to advance the technology readiness level of phosphate glass waste forms being developed to immobilize high-level radioactive salt waste streams are identified, the states of understanding various technical aspects of formulation, processing, and performance are summarized, and approaches supporting further development are recommended. Processing results in dehalogenation of the waste salt, capture of the gaseous halide-bearing species, and immobilization of the residual salt components in a phosphate glass waste form. The approach is suitable for high-level salt waste from electrochemical reprocessing and molten salt reactors. The technology has been demonstrated for chloride-based salts and may also be suitable for the treatment and immobilization of fluoride-based and iodide-bearing waste salts. Aspects of the process requiring further development are identified and approaches recommended.

This paper provides microstructural, microchemical, and bulk X-ray diffraction information for a zinc-in-titania waste form for immobilizing rare-earth oxide (REOx) fission products recovered through precipitation from electrochemical salt wastes. This waste form can be used to immobilize REOx at loadings as high as 40 mass% in a chemically durable matrix consisting of glass-bonded minerals including Zn2TiO4 as well as RE-containing phases of REOx, RE2Ti2O7, and REPO4 (monazite). Chemical durability data is also provided, and comparisons are drawn between this waste form and one made using a similar process with half the REOx loading.

This work is being performed as part of the DOE NE Spent Fuel and Waste Science and Technology Campaign, Argillite and Crystalline Rock R&D work packages: SF-20AN01030101 and SF-20AN01030201 respectively. This document is submitted in fulfillment of milestone M4SF-20AN010301013 for Argillite R&D and the milestone M4SF-20AN010302013 for Crystalline R&D

This study demonstrates the conversion of a simulated salt waste from electrochemical fuel reprocessing into a chemically durable iron phosphate glass waste form through reaction of the salt with ammonium hydrogen phosphate precursors [i.e., NH4H2PO4 and (NH4)(2)HPO4] and adding Fe2O3. The reaction of the salt simulant ERV2 with the phosphate precursors resulted in dehalogenation of the salt through the production of gaseous NH4CI and water vapor (plus other minor byproducts) with the residual salt cations incorporating into the iron phosphate glass as oxides. Physical properties and chemical durabilities were measured for iron phosphate glasses made using different salt/phosphate/Fe2O3 ratios. These iron phosphate glasses accommodate high salt cation loadings and have high bulk densities (similar to 3 x 10(3) kg/m(3)), which would provide a high storage volume waste form option (i.e., the volume required to immobilize a given mass of waste). A proof-of-concept demonstration was conducted to show that the NH4CI off-gas generated during this process reacts with uranium dendrites to directly produce UCI3 that can be recycled as a feedstock for electrochemical processing. (C) 2019 Elsevier B.V. All rights reserved.

Borosilicate waste glass degradation models must quantify the effects of the solution composition on the dissolution rate. Here, we present results of modified ASTM C1285 tests conducted at 90 °C with AFCI and LRM glasses to determine whether dependencies of dissolution rates on the pH, Al, and Si concentrations must be included. Solution compositions were modified from those generated by glass dissolution alone by adding small amounts of K4SiO4 glass, Al(OH)3•2H2O, and a concentrated NaOH solution when the tests were initiated. Results show rate laws for the initial and resumption regimes must include pH dependences, but the residual rate can be modeled independent of the pH, Al, and Si concentrations. Triggering the resumption rate probably depends on the pH, Si, and Al concentrations and perhaps other aspects of the glass composition. A waste glass degradation model using is being parameterized using tests with a range of waste glass compositions to quantify these dependencies.

This paper provides an overview of research evaluating the use of lead tellurite glass as a waste form for salt wastes from electrochemical reprocessing of used nuclear fuel. The efficacy of using lead tellurite glass to immobilize three different salt compositions was evaluated: a LiCl-Li2O oxide reduction salt containing fission products from oxide fuel, a LiCl-KCl eutectic salt containing fission products from metallic fuel, and SrCl2. Physical and chemical properties of glasses made with these salts were characterized with X-ray diffraction, bulk density measurements, differential thermal analysis, chemical durability tests, scanning and transmission electron microscopies, and energy-dispersive X-ray spectroscopy. These glasses were found to accommodate high salt concentrations and have high densities, but further development is needed to improve chemical durability.

The interface with reactive transport models used in performance assessment calculations is described to identify aspects of the glass waste form degradation model important to long-term predictions. These are primarily the conditions that trigger the change from the residual rate to the Stage 3 rate and the values of those rates. Although the processes triggering the change and controlling the Stage 3 rate are not yet understood mechanistically, neither appears related to an intrinsic property of the glass. The sudden and usually significant increase in the glass dissolution rate suggests the processes that trigger the increase are different than the processes controlling glass dissolution prior to that change. Application of a simple expression that was derived for mineral transformation to represent the kinetics of coupled glass dissolution and secondary phase precipitation reactions is shown to be consistent with experimental observations of Stage 3 and useful for modeling long-term glass dissolution in a complex disposal environment.

The Fuel Matrix Dissolution Model (FMDM) is an electrochemical reaction/diffusion model for the dissolution of spent uranium oxide fuel. The model was developed to provide radionuclide source terms for use in performance assessment calculations for various types of geologic repositories. It is based on mixed potential theory and consists of a two-phase fuel surface made up of UO2 and a noble metal bearing fission product phase in contact with groundwater. The corrosion potential at the surface of the dissolving fuel is calculated by balancing cathodic and anodic reactions occurring at the solution interfaces with UO2 and NMP surfaces. Dissolved oxygen and hydrogen peroxide generated by radiolysis of the groundwater are the major oxidizing agents that promote fuel dissolution. Several reactions occurring on noble metal alloy surfaces are electrically coupled to the UO2 and can catalyze or inhibit oxidative dissolution of the fuel. The most important of these is the oxidation of hydrogen, which counteracts the effects of oxidants (primarily H2O2 and O-2). Inclusion of this reaction greatly decreases the oxidation of U(IV) and slows fuel dissolution significantly. In addition to radiolytic hydrogen, large quantities of hydrogen can be produced by the anoxic corrosion of steel structures within and near the fuel waste package.The model accurately predicts key experimental trends seen in literature data, the most important being the dramatic depression of the fuel dissolution rate by the presence of dissolved hydrogen at even relatively low concentrations (e.g., less than 1 mM). This hydrogen effect counteracts oxidation reactions and can limit fuel degradation to chemical dissolution, which results in radionuclide source term values that are four or five orders of magnitude lower than when oxidative dissolution processes are operative. This paper presents the scientific basis of the model, the approach for modeling used fuel in a disposal system, and preliminary calculations to demonstrate the application and value of the model. (C) 2015 Elsevier B.V. All rights reserved.

Series of 7-day Product Consistency Tests (PCTs) were conducted with ARM-1 glass using the -100+200 mesh size fraction and several sub-fractions to measure the sensitivity of the test response to the distribution of particle sizes. Separate samples were prepared for testing by dry sieving and wet sieving, and the particle size distributions and PCT responses were measured for each fraction. Triplicate tests were conducted at 90 °C using a water/glass mass ratio of 10.0 with each size fraction. Test results are evaluated regarding the sensitivity of the test response to the particle size distributions and, conversely, the uncertainty due to calculating the surface areas (and dissolution rates) by modeling the particles as spheres. These analyses show the solution feedback effects of dissolved glass constituents (i.e., the reaction affinity) counteract the effects of the glass surface areas provided by different particle size distributions on the test response. The opposing effects of the surface area on the amount of glass dissolved and on the glass dissolution rate moderate the sensitivity of the PCT response to the particle size distribution.

This study is aimed at evaluating the existing waste management approaches for nuclear fuel cycle facilities in comparison to the objectives of implementing an advanced fuel cycle in the U.S. under current legal, regulatory, and logistical constructs. The study begins with the Global Nuclear Energy Partnership (GNEP) Integrated Waste Management Strategy (IWMS) (Gombert et al. 2008) as a general strategy and associated Waste Treatment Baseline Study (WTBS) (Gombert et al. 2007). The tenets of the IWMS are equally valid to the current waste management study. However, the flowsheet details have changed significantly from those considered under GNEP. In addition, significant additional waste management technology development has occurred since the GNEP waste management studies were performed. This study updates the information found in the WTBS, summarizes the results of more recent technology development efforts, and describes waste management approaches as they apply to a representative full recycle reprocessing flowsheet. Many of the waste management technologies discussed also apply to other potential flowsheets that involve reprocessing. These applications are occasionally discussed where the data are more readily available. The report summarizes the waste arising from aqueous reprocessing of a typical light-water reactor (LWR) fuel to separate actinides for use in fabricating metal sodium fast reactor (SFR) fuel and from electrochemical reprocessing of the metal SFR fuel to separate actinides for recycle back into the SFR in the form of metal fuel. The primary streams considered and the recommended waste forms include; Tritium in low-water cement in high integrity containers (HICs); Iodine-129: As a reference case, a glass composite material (GCM) formed by the encapsulation of the silver Mordenite (AgZ) getter material in a low-temperature glass is assumed. A number of alternatives with distinct advantages are also considered including a fused silica waste form with encapsulated nano-sized AgI crystals; Carbon-14 immobilized as a CaCO3 in a cement waste form; Krypton-85 stored as a compressed gas; An aqueous reprocessing high-level waste (HLW) raffinate waste immobilized by the vitrification process; An undissolved solids (UDS) fraction from aqueous reprocessing of LWR fuel either included in the borosilicate HLW glass or immobilized in the form of a metal alloy or titanate ceramics; Zirconium-based LWR fuel cladding hulls and stainless steel (SS) fuel assembly hardware super-compacted for disposal or purified for reuse (or disposal as low-level waste, LLW) of Zr by reactive gas separations; Electrochemical process salt HLW incorporated into a glass bonded Sodalite waste form; and Electrochemical process UDS and SS cladding hulls melted into an iron based alloy waste form. Mass and volume estimates for each of the recommended waste forms based on the source terms from a representative flowsheet are reported. In addition to the above listed primary waste streams, a range of secondary process wastes are generated by aqueous reprocessing of LWR fuel, metal SFR fuel fabrication, and electrochemical reprocessing of SFR fuel. These secondary wastes have been summarized and volumes estimated by type and classification. The important waste management data gaps and research needs have been summarized for each primary waste stream and selected waste process.

The R&D program from the DOE Used Fuel Disposition Campaign (UFDC) has documented key advances in coupled Thermal-Hydrological-Mechanical-Chemical (THMC) modeling of clay to simulate its complex dynamic behavior in response to thermal and hydrochemical feedbacks. These efforts have been harnessed to assess the isolation performance of heat-generating nuclear waste in a deep geological repository in clay/shale/argillaceous rock formations. This report describes the ongoing disposal R&D efforts on the advancement and refinement of coupled THMC process models, hydrothermal experiments on barrier clay interactions, used fuel and canister material degradation, thermodynamic database development, and reactive transport modeling of the near-field under non-isothermal conditions. These play an important role to the evaluation of sacrificial zones as part of the EBS exposure to thermally-driven chemical and transport processes. Thermal inducement of chemical interactions at EBS domains enhances mineral dissolution/precipitation but also generates mineralogical changes that result in mineral H2O uptake/removal (hydration/dehydration reactions). These processes can result in volume changes that can affect the interface / bulk phase porosities and the mechanical (stress) state of the bentonite barrier. Characterization studies on bentonite barrier samples from the FEBEX-DP international activity have provided important insight on clay barrier microstructures (e.g., microcracks) and interactions at EBS interfaces. Enhancements to the used fuel degradation model outlines the need to include the effects of canister corrosion due the strong influence of H2 generation on the source term.

Optional grouted waste forms were formulated for waste streams generated during the production of 99Mo to be compliant with low-level radioactive waste regulations. The amounts and dose rates of the various waste form materials that would be generated annually were estimated and used to determine the effects of various waste processing options, such as the of number irradiation cycles between uranium recovery operations, different combinations of waste streams, and removal of Pu, Cs, and Sr from waste streams for separate disposition (which is not evaluated in this report). These calculations indicate that Class C-compliant grouted waste forms can be produced for all waste streams. More frequent uranium recovery results in the generation of more chemical waste, but this is balanced by the fact that waste forms for those waste streams can accommodate higher waste loadings, such that similar amounts of grouted waste forms are required regardless of the recovery schedule. Similar amounts of grouted waste form are likewise needed for the individual and combined waste streams. Removing Pu, Cs, and Sr from waste streams lowers the waste form dose significantly at times beyond about 1 year after irradiation, which may benefit handling and transport. Although these calculations should be revised after experimentally optimizing the grout formulations and waste loadings, they provide initial guidance for process development.