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This report summarizes the 2011 fiscal+calendar year efforts for developing waste forms for a spent salt generated in reprocessing nuclear fuel with an electrochemical separations process. The two waste forms are tellurite (TeO 2 -based) glasses and sol-gel-derived high-halide mineral analogs to stable minerals found in nature.
Epsilon metal (ε-metal), an alloy of Mo, Pd, Rh, Ru, and Tc, is being developed as a waste form to treat and immobilize the undissolved solids and dissolved noble metals from aqueous reprocessing of commercial used nuclear fuel. Epsilon metal is an attractive waste form for several reasons: increased durability relative to borosilicate glass, it can be fabricated without additives (100% waste loading), and in addition it also benefits borosilicate glass waste loading by eliminating noble metals from the glass and thus the processing problems related there insolubility in glass. This work focused on the processing aspects of the epsilon metal waste form development. Epsilon metal is comprised of refractory metals resulting in high reaction temperatures to form the alloy, expected to be 1500 - 2000°C making it a non-trivial phase to fabricate by traditional methods. Three commercially available advanced technologies were identified: spark-plasma sintering, microwave sintering, and hot isostatic pressing, and investigated as potential methods to fabricate this waste form. Results of these investigations are reported and compared in terms of bulk density, phase assemblage (X-ray diffraction and elemental analysis), and microstructure (scanning electron microscopy).
A review of the most up-to-date and relevant data currently available was conducted to develop a set of recommended values for use in the Integrated Disposal Facility (IDF) performance assessment (PA) to model contaminant release from a cementitious waste form for aqueous wastes treated at the Hanford Effluent Treatment Facility (ETF). This data package relies primarily upon recent data collected on Cast Stone formulations fabricated with simulants of low-activity waste (LAW) and liquid secondary wastes expected to be produced at Hanford. These data were supplemented, when necessary, with data developed for saltstone (a similar grout waste form used at the Savannah River Site). Work is currently underway to collect data on cementitious waste forms that are similar to Cast Stone and saltstone but are tailored to the characteristics of ETF-treated liquid secondary wastes. Recommended values for key parameters to conduct PA modeling of contaminant release from ETF-treated liquid waste are provided.
Reference alloy waste forms (RAW) were fabricated and underwent hybrid corrosion/immersion testing to parameterize the ANL analytical oxidative-dissolution model to enable the calculation of fractional release rates and to determine the effectiveness of Ni and Cr trim additions in reducing release rates of radionuclide surrogates. Figure 1 shows the prototypical multiphase microstructure of the alloys with each phase type contributing about equally to the exposed surface area. The waste forms tested at SRNL were variations of the RAW-6 formulation that uses HT9 as the main alloy component, and are meant to enable evaluation of the impact of Ni and Cr trim additions on the release rates of actinides and Tc-99. The test solutions were deaerated alkaline and acidic brines, ranging in pH 3 to pH 10, representing potential repositories with those conditions. The testing approach consisted of 4 major steps; 1) bare surface corrosion measurements at pH values of 3, 5, 8, and 10, 2) hybrid potentiostatic hold/exposure measurements at pH 3, 3) measurement of radionuclide concentrations and relations to anodic current from potentiostatic holds, and 4) identification of corroding phases using SEM/EDS of electrodes.
This project aimed to gain detailed and fundamental understanding of the stability of glass waste forms to be used for safe and reliable encapsulation of nuclear waste. The focus was to understand thermodynamic and structural properties involved in crystallization and phase separation in melt-derived waste form glasses and glass-ceramics by studying a baseline glass composition with varying additives. This was achieved by combining neutron total scattering, advanced calorimetric measurements, and extensive modeling study. This project harnessed diverse and complementary approaches and expertise from four collaborating interdisciplinary institutions (two universities, one national laboratory, and one UK university partner). Samples were prepared at the national laboratory and exchanged with the two university institutions. The thermodynamic data obtained by various advanced calorimetry techniques were linked to the underlying short-range structure of waste forms measured by state-of-the-art neutron total scattering experiments. An important element of this project was extensive modeling to interpret and fundamentally understand the experimental results.
This document meets Level 4 Milestone: Corrosion mechanisms for metal alloy waste forms - experiment and theory. A multiphysics model is introduces that will provide the framework for the quantitative prediction of corrosion rates of metallic waste forms incorporating the fission product Tc. The model requires a knowledge of the properties of not only the metallic waste form, but also the passive oxide films that will be generated on the waste form, and the chemistry of the metal/oxide and oxide/environment interfaces. in collaboration with experimental work, the focus of this work is on obtaining these properties from fundamental atomistic models. herein we describe the overall multiphysics model, which is based on MacDonald's point-defect model for passivity. We then present the results of detailed electronic-structure calculations for the determination of the compatibility and properties of Tc when incorporated into intermetallic oxide phases. This work is relevant to the formation of multi-component oxides on metal surfaces that will incorporate Tc, and provide a kinetic barrier to corrosion (i.e. the release of Tc to the environment). Atomistic models that build upon the electronic structure calculations are then described using the modified embedded atom method to simulate metallic dissolution, and Buckingham potentials to perform classical molecular dynamics and statics simulations of the technetium (and, later, iron-technetium) oxide phases. Electrochemical methods were then applied to provide some benchmark information of the corrosion and electrochemical properties of Technetium metal. The results indicate that published information on Tc passivity is not complete and that further investigation is warranted.
Corrosion tests are in progress to assess the dissolution behavior of developmental iron phosphate waste forms for salt waste from electrochemical separation operations. Specimens cut from materials made at PNNL to represent a range of waste salt loadings in prototype composition DPF-5 and a range of iron oxide loadings in prototype composition DPF-3 were subjected to modified ASTM C1308 tests to assess the compositional effects on the relative durabilities and retention of immobilized salt cations. The materials were composed of different amounts of glass and crystalline phases. The test results show both the intrinsic durability of the material and the salt cation retention decrease as the total amount of salt cations added to the material is increased. The results of ASTM C1220 and ASTM C1285 tests conducted with materials DPF5-280 and DPF5-336, which contained about 28% and 34% salt cations (on a mass oxide basis) and bounded the durabilities of the other materials, show that solution feedback does not attenuate the dissolution rate. This suggests the degradation of iron phosphate waste forms can be modelled by using a constant rate that probably depends on the composition and volume fraction of the glass phase. A reference composition should be formulated and used in tests to determine those dependencies. The ASTM C1220 and ASTM C1285 tests with DPF5-280 and DPF5-336 remain in progress to corrode the specimens enough to distinguish the relative extents of glass and crystalline phase corrosion by microscopic examinations. Those results are expected to provide insights regarding the relative dissolution rates of crystalline and glass phases and impact on durability to inform the formulation of the reference composition.
2 prototype alloys: RAW-1(Tc) and RAW-2(UTc) suitable for a wide range of waste stream compositions are being evaluated to support development of a waste form degradation model that can be used to calculate radionuclide source terms for a range of waste form compositions and disposal environments. Tests and analyses to support formulation of waste forms and development of the degradation model include detailed characterizations of the constituent phases using SEM/EDS and TEM, electrochemical tests to quantify the oxidation behavior and kinetics of the individual and coupled phases under a wide range of environmental conditions, and corrosion tests to measure the gross release kinetics of radionuclides under aggressive test conditions.
In this study, multi-phase silicate-based glass-ceramics were investigated as an alternate waste form for immobilizing non-fissionable products from used nuclear fuel. Currently, borosilicate glass is the waste form selected for immobilization of this waste stream, however, the low thermal stability and solubility of MoO{sub 3} in borosilicate glass translates into a maximum waste loading in the range of 15-20 mass%. Glass-ceramics provide the opportunity to target durable crystalline phases, e.g., powellite, oxyapatite, celsian, and pollucite, that will incorporate MoO{sub 3} as well as other waste components such as lanthanides, alkalis, and alkaline earths at levels 2X the solubility limits of a single-phase glass. In addition a glass-ceramic could provide higher thermal stability, depending upon the properties of the crystalline and amorphous phases. Glass-ceramics were successfully synthesized at waste loadings of 42, 45, and 50 mass% with the following glass additives: B{sub 2}O{sub 3}, Al{sub 2}O{sub 3}, CaO and SiO{sub 2} by slow cooling form from a glass melt. Glass-ceramics were characterized in terms of phase assemblage, morphology, and thermal stability. The targeted phases: powellite and oxyapatite were observed in all of the compositions along with a lanthanide borosilicate, and cerianite. Results of this initial investigation of glass-ceramics show promise as a potential waste form to replace single-phase borosilicate glass.
The Savannah River National Laboratory (SRNL) conducted research from FY2011 – FY2020 to develop ceramic waste forms to treat High-Level Waste (HLW) resulting from reprocessing of commercial Used Nuclear Fuel (UNF) as part of the Department of Energy (DOE) office of Nuclear Energy (NE) Fuel Cycle Technologies (FCT) research program. The objective of the research was to develop a reference ceramic waste form and process technology needed to immobilize the combined HLW raffinate stream, including undissolved solids and soluble technetium, and potentially the TRU waste stream resulting from reprocessing processes. This report compiles the relevant literature produced under, and pertinent to, the program objectives and provides a perspective on the technology development as well as prospects for future.
Materials and methods of making low-sintering-temperature glass waste forms that sequester radioactive iodine in a strong and durable structure. First, the iodine is captured by an adsorbant, which forms an iodine-loaded material, e.g., AgI, AgI-zeolite, AgI-mordenite, Ag-silica aerogel, ZnI.sub.2, CuI, or Bi.sub.5O.sub.7I. Next, particles of the iodine-loaded material are mixed with powdered frits of low-sintering-temperature glasses (comprising various oxides of Si, B, Bi, Pb, and Zn), and then sintered at a relatively low temperature, ranging from 425.degree. C. to 550.degree. C. The sintering converts the mixed powders into a solid block of a glassy waste form, having low iodine leaching rates. The vitrified glassy waste form can contain as much as 60 wt % AgI. A preferred glass, having a sintering temperature of 500.degree. C. (below the silver iodide sublimation temperature of 500.degree. C.) was identified that contains oxides of boron, bismuth, and zinc, while containing essentially no lead or silicon.
The Separations and Waste Form Campaign (SWFC) under the U.S. Department of Energy (DOE) Fuel Cycle Research and Development Program (FCRD) is responsible for developing advanced separation and waste form technologies to support the various fuel cycle options defined in the DOE Nuclear Energy Research and Development Roadmap, Report to Congress, April 2010. The fiscal year (FY) 2013 accomplishments report provides a highlight of the results of the research and development (R&D) efforts performed within SWFC in FY 2013. Each section contains a high-level overview of the activities, results, technical point of contact, applicable references, and documents produced during the fiscal year. This report briefly outlines campaign management and integration activities, but the intent of the report is to highlight the many technical accomplishments made during FY 2013.
INTRODUCTION Cementitious matrices are used in the US DOE complex and worldwide to solidify aqueous radioactive, hazardous, mixed salt solutions, and sludges to meet low-level radioactive waste (LLW) disposal requirements. Blended formulations are used for most waste form mix designs. Substitution of pozzolans for Class F fly ash has the potential to alter the processing properties and stabilization properties of the resulting waste forms because they add chemical and mineralogical complexity to the final material. The Savannah River Site saltstone waste form was selected as the test case material. CemGEMS software was chosen as the model for predicting hydrated cementitious phases assemblages and phase evolution. Isothermal calorimetry was used as the method for evaluating processing and properties of fresh, uncured waste forms. The test cases consisted of reference case saltstone, and five natural pozzolan substituted saltstone mixes.
This report represents the milestone deliverable M4SF-21SN010309021 “Modeling Activities Related to Waste Form Degradation: Progress Report” that describes the progress of R&D activities of ongoing modeling investigations specifically on nuclear waste glass degradation, Density Functional Theory (DFT) studies on clarkeite structure and stability, and electrochemical modeling of spent nuclear fuel (SNF). These activities are part of the newly-created Waste form Testing, Modeling, and Performance work package at Sandia National Laboratories (SNL). This work package is part of the “Inventory and Waste Form Characteristics and Performance” control account that includes various experimental and modeling activities on nuclear waste degradation conducted at Oak Ridge National Laboratory (ORNL), SNL, Argonne National Laboratory (ANL), and Pacific Northwest National Laboratory (PNNL).
This report represents the milestone deliverable M4SF-22SN010309092 “Modeling Activities Related to Waste Form Degradation: Progress Report” that describes the progress of R&D activities of ongoing modeling investigations specifically on nuclear waste glass degradation, Density Functional Theory (DFT) studies on clarkeite structure and stability, and electrochemical model development of spent nuclear fuel (SNF). These activities are part of the Waste Form Testing, Modeling, and Performance work package at Sandia National Laboratories (SNL). This work package is part of the “Inventory and Waste Form Characteristics and Performance” control account that includes various experimental and modeling activities on nuclear waste degradation conducted at Oak Ridge National Laboratory (ORNL), SNL, Argonne National Laboratory (ANL), and Pacific Northwest National Laboratory (PNNL).
In Corrosion studies were undertaken on alloy waste forms that can result from advanced electrometallurgical processing techniques to better classify their durability and degradation mechanisms. The waste forms were based on the RAW3-(URe) composition, consisting primarily of HT9 steel and other elemental additions to simulate nuclear fuel reprocessing byproducts. The solution conditions of the corrosion studies were taken from an electrochemical testing protocol, and meant to simulate conditions in a repository. The alloys durability was examined in alkaline and acidic brines.
The Department of Energy (DOE) is currently responsible for treating radioactive and mixed waste, performing environmental restoration, and closing contaminated tanks and facilities resulting from nuclear weapons production during the Cold War. Cementitious reagents are the most widely used materials for (1) chemically stabilizing and encapsulating radionuclides and hazardous metals and (2) solidifying radioactive wastewater. Cementitious grouts and flowable concretes are also the most widely used materials for tank and facility closures and are used extensively for physical as well as chemical stabilization in environmental restoration projects. Ambient temperature radioactive waste cementation is a widely used technology for producing waste forms for final disposal. The current practice of designing and testing waste forms is based on a mid to late 20th century technology approach (i.e., materials, characterization, and test methods) for generating parameters for risk assessments. DOE technology development is needed to address both current, long-term, and emerging issues in this area as new waste streams come online and as regulations, performance knowledge, and risk assessment methodology continue to evolve. Consequently, the use of cementation as a means of treating chemically challenging radioactive liquid waste streams and reactive debris requires an enduring effort.
The mission of the Waste Isolation Pilot Plant in New Mexico is the safe and permanent disposition of transuranic radioactive waste from atomic energy defense activities of the United States. The plant consists of an underground repository in a salt formation that is free of flowing water, easily mined, impermeable, and geologically stable. The salt rock naturally seals fractures and closes openings over time. The repository is now anticipated to receive a greater amount of plutonium materials than originally envisioned. An engineered waste form for safe and efficient transport and long-term disposal of plutonium materials is an urgent need. An optimal waste form could also reduce the disposition life cycle and its associated costs to the taxpayer.