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At least 145 records · Page 8
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Apatite and sodalite based glass-bonded waste forms for immobilization of 129I and mixed halide radioactive wastes

The goal of the project was to utilize the knowledge accumulated by the team, in working with minerals for chloride wastes and biological apatites, toward the development of advanced waste forms for immobilizing 129I and mixed-halide wastes. Based on our knowledge, experience, and thorough literature review, we had selected two minerals with different crystal structures and potential for high chemical durability, sodalite and CaP/PbV-apatite, to form the basis of this project. The focus of the proposed effort was towards: (i) low temperature synthesis of proposed minerals (iodine containing sodalite and apatite) leading to the development of monolithic waste forms, (ii) development of a fundamental understanding of the atomic-scale to meso-scale mechanisms of radionuclide incorporation in them, and (iii) understanding of the mechanism of their chemical corrosion, alteration mechanism, and rates. The proposed work was divided into four broad sections. deliverables. 1. Synthesis of materials 2. Materials structural and thermal characterization 3. Design of glass compositions and synthesis glass-bonded minerals, and 4. Chemical durability testing of materials.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Corrosion Behavior of Fe-7Cr-1Mo-12Zr Metal Waste Form Alloy in Simulated Repository Environment

Metal waste forms (MWF) are solid metallic waste categorized as high-level radioactive waste generated during pyro-reprocessing of spent metallic fuels and geologically disposed off. This waste form generated requires an excellent corrosion resistance in a geological repository environment. In the present work, Fe-7Cr-1Mo-12Zr MWF alloy was developed by casting in an arc melting furnace. The x-ray diffraction analysis revealed the formation of stable phases of α-Fe solid solution and the intermetallic phases of Fe{sub 2}Zr and Fe{sub 23}Zr{sub 6}. The corrosion behavior of Fe-7Cr-1Mo-12Zr alloy was investigated using electrochemical techniques in aggressive (highly acidic with Cl{sup −}) and nominal repository pH environment. Cyclic potentiodynamic polarization studies indicated the shift in corrosion potential toward cathodic potential in highly acidic solution with Cl{sup −}. The corrosion mechanism of pitting is associated with acidified pH medium with more Cl{sup −}, whereas uniform corrosion or transpassive dissolution is resultant of mild acidity or mild alkaline medium. The passive region is wider in pH 5 and 8 when compared to pH 1 and 3. Electrochemical impedance spectroscopic (EIS) analysis revealed the lowering of passive film stability in the highly acidic medium in the presence of Cl{sup −}. The role of passive oxide layers on the corrosion resistance of the Fe-7Cr-1Mo-12Zr MWF alloy is discussed.

36 MATERIALS SCIENCE↗
DOE OSTI2015

X-Ray Diffraction of Slag-Based Sodium Salt Waste Forms - 15513

Cementitious materials are used to solidify and stabilize aqueous based radioactive waste containing sodium salts. The types and proportions of cementitious ingredients used to treat aqueous radioactive waste streams containing sodium salts depend on the performance objectives for the waste forms and the compositions of the waste streams. This paper documents sample preparation and x-ray diffraction results for a series of materials made with water or highly alkaline sodium salt simulated waste water and cementitious binders. The objective of this study was to: 1) generate a base line for the evolution of the waste form as a function of time and conditions, 2) design new binders based on mineralogy of the binder, 3) understand and predict anion and cation leaching behavior of contaminants of concern, and 4) predict performance of the waste forms for which phase solubility and thermodynamic data are available. Characterization of the mineralogy is also important for understanding the buffering effects that the waste form has on infiltrating water / leachates. In summary, mixtures of Type II portland cement, Grade 100 ground granulated blast furnace slag (GGBFS) and carbon burn-out (CBO) Class F Fly ash which were hydrated with water contained hydrated phase assemblages typical of those reported in the literature. The calcium silicate hydrate phase assemblage in samples hydrated with the alkaline 4.4 M sodium salt simulated waste solution was found to be a function of the (CaO + MgO) / (SiO{sub 2} + Al{sub 2}O{sub 3}) ratio of the samples characterized. No significant differences were detected in samples cured 2 months and 14 months in sealed containers at ambient indoor temperatures. Slag and a blend of slag and cement hydrated with caustic 4.4 M Na salt solution resulted in the most crystalline matrix. In addition to poorly ordered C-S-H, these samples contained fairly well ordered C-S-H I (a precursor of 14A tobermorite) and 11 A Al-substituted tobermorite. These crystalline C-S-H phases did not form or were present in only trace amounts in slag blends containing about 45 to 62 mass percent fly ash. These slag-Class F fly ash blends had a higher silica plus alumina content relative to lime and magnesia than the blends that produced C-S-H I and Al-substituted tobermorite. The calcium silicate binder in the 10:45:45 mixture of cement: slag: fly ash was made up of poorly ordered C-S-H. The sample cured for 14 months may contain a small amount of the more crystalline calcium silicate hydrate phases. Layered double hydroxides in the hydrotalcite (magnesium-aluminum carbonate hydroxide) and hydrocalumite / AFm phases (calcium aluminum hydroxide) were present in mixtures containing slag. The specific phase(s) were not identified because these phases form solid solutions and have a considerable amount of overlap in their x-ray patterns. Sodium nitrate was the only sodium salt phase identified in x-ray diffraction patterns of the samples hydrated with salt solution. Drying during x-ray diffraction sample preparation may have resulted in precipitation of the sodium nitrate or it may have been present in the samples prior to x-ray sample preparation. Sodium sulfate, aluminate, and carbonate may have been incorporated in the structures of the layered double hydroxide (AFm) type phases. These mixed metal layered double hydroxides make up an important fraction of the matrix in the slag containing blends hydrated with caustic salt solution. They are among the few oxide-based phases that exhibit substantial, permanent anion exchange capacity [Kirkpatrick, et al. 1999, Plamer, et al., 2009, and Zhang and Reardon, 2003]. They also contribute to the structural properties of cementitious matrices [Taylor, 1997]. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2011

Computational Efficient Upscaling Methodology for Predicting Thermal Conductivity of Nuclear Waste forms

This study evaluated different upscaling methods to predict thermal conductivity in loaded nuclear waste form, a heterogeneous material system. The efficiency and accuracy of these methods were compared. Thermal conductivity in loaded nuclear waste form is an important property specific to scientific researchers, in waste form Integrated performance and safety code (IPSC). The effective thermal conductivity obtained from microstructure information and local thermal conductivity of different components is critical in predicting the life and performance of waste form during storage. How the heat generated during storage is directly related to thermal conductivity, which in turn determining the mechanical deformation behavior, corrosion resistance and aging performance. Several methods, including the Taylor model, Sachs model, self-consistent model, and statistical upscaling models were developed and implemented. Due to the absence of experimental data, prediction results from finite element method (FEM) were used as reference to determine the accuracy of different upscaling models. Micrographs from different loading of nuclear waste were used in the prediction of thermal conductivity. Prediction results demonstrated that in term of efficiency, boundary models (Taylor and Sachs model) are better than self consistent model, statistical upscaling method and FEM. Balancing the computation resource and accuracy, statistical upscaling is a computational efficient method in predicting effective thermal conductivity for nuclear waste form.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2014

Radioactive Demonstration Of Mineralized Waste Forms Made From Hanford Low Activity Waste (Tank SX-105 And AN-103) By Fluidized Bed Steam Reformation

One of the immobilization technologies under consideration as a Supplemental Treatment for Hanford’s Low Activity Waste (LAW) is Fluidized Bed Steam Reforming (FBSR). The FBSR technology forms a mineral waste form at moderate processing temperatures thus retaining and atomically bonding the halides, sulfates, and technetium in the mineral phases (nepheline, sodalite, nosean, carnegieite). Additions of kaolin clay are used instead of glass formers and the minerals formed by the FBSR technology offers (1) atomic bonding of the radionuclides and constituents of concern (COC) comparable to glass, (2) short and long term durability comparable to glass, (3) disposal volumes comparable to glass, and (4) higher Na2O and SO{sub 4} waste loadings than glass. The higher FBSR Na{sub 2}O and SO{sub 4} waste loadings contribute to the low disposal volumes but also provide for more rapid processing of the LAW. Recent FBSR processing and testing of Hanford radioactive LAW (Tank SX-105 and AN-103) waste is reported and compared to previous radioactive and non-radioactive LAW processing and testing.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Reduction of Porosity in Cementitious Waste Forms to Reduce Effective Diffusivity

Although current practices meet regulations for grout waste forms, development of sequestration materials and techniques for contaminants of concern could expand the use of grout waste forms and reduce risk to established processes from future compositional changes. There are contaminants in waste streams that are not permanently and completely sequestered in current grout waste forms under certain conditions. These contaminants include methyl mercury, 129 I, 99 Tc, and nitrate. Each of these species have minor fractions that are not permanently and completely isolated from the environment and therefore potentially contribute to long-term environmental impacts. An approach of interest includes the technique of reducing porosity to result in reduced leachability of all contaminants simultaneously. The current Savannah River Site (SRS) saltstone formulation was used to evaluate the relationship between leachability and porosity reduced waste forms by altering the water-to-binder ratio. Results indicated decreasing the water-to-binder directly reduced porosity and resulted in the waste form’s ability to more effectively immobilize nitrate. This work was funded by the Department of Energy Office of Environmental Management (DOE-EM) Technology Operations Office (TOO) Project # HQ221818.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Ion Exchange of LiCl-KCl in H-Y Zeolite for Dechlorination of Electro-refiner Salt Waste Forms

The present study focuses on the aim to significantly reduce the volume of electro-refiner waste salt (primarily LiCl-KCl) via dechlorination during waste form production. The target waste form will be zeolite loaded with metallic elements from the salt bound together by glass. Since the presence of I-129 and possibly Cl-36 will have a significant effect on long-term dose associated with the waste form, the ability to separate these halides and either recycle or dispose of them in special engineered waste forms is deemed attractive. A new method has been proposed to achieve this dechlorination involving the ion-exchange reaction between molten chloride ER salt and the protonated Y zeolite. Halides such as Cl and I should off-gas as HCl or HI, which can then be separated and decomposed into Cl{sub 2} and I{sub 2}, respectively. The Cl{sub 2} gas can either be captured and disposed of, or recycled to the electrometallurgical process. Initial experiments have been completed to determine the optimal zeolite drying temperature and to demonstrate ion exchange with non-optimized conditions. Further study in the future will strive to improve the kinetics of the ion-exchange reaction and to maximize the removal of the chlorine.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Ceramic waste form production from used nuclear fuel

According to one aspect of the invention, a method to create a ceramic waste form from used nuclear fuel. An active metal salt waste, a rare earth metal waste, and raw materials are received. The active metal salt waste is combined with the rare earth metal waste, forming a waste salt. The waste salt is then heated to approximately 500° C. The raw materials are also heated to approximately 500° C. The waste salt and raw materials are then blended to form a homogenous waste mixture. The homogenous waste mixture is heated to a first predetermined temperature for a predetermined amount of time, creating a ceramic waste form. The ceramic waste form is cooled to a second predetermined temperature.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2011

Evaluation of THOR TM Mineralized Waste Forms (Granular and Monolith) for the DOE Advanced Remediation Technologies (ART) Phase 2 Project

The U.S. Department of Energy's (DOE) Office of River Protection (ORP) is responsible for the retrieval, treatment, immobilization, and disposal of Hanford's tank waste. Currently there are approximately 56 million gallons of highly radioactive mixed wastes awaiting treatment. A key aspect of the River Protection Project (RPP) cleanup mission is to construct and operate the Waste Treatment and Immobilization Plant (WTP). The WTP will separate the tank waste into high-level and low-activity waste (LAW) fractions, both of which will subsequently be vitrified. The projected throughput capacity of the WTP LAW Vitrification Facility is insufficient to complete the RPP mission in the time frame required by the Hanford Federal Facility Agreement and Consent Order, also known as the Tri-Party Agreement (TPA), i.e. December 31, 2047. Therefore, Supplemental Treatment is required both to meet the TPA treatment requirements as well as to more cost effectively complete the tank waste treatment mission. The Supplemental Treatment chosen will immobilize that portion of the retrieved LAW that is not sent to the WTP's LAW Vitrification facility into a solidified waste form. The solidified waste will then be disposed on the Hanford site in the Integrated Disposal Facility (IDF). In addition, the WTP LAW Vitrification facility off-gas condensate known as WTP Secondary Waste (WTP-SW) will be generated and enriched in volatile components such as Cs-137, I-129, Tc-99, Cl, F, and SO 4 that volatilize at the vitrification temperature of 1150°C in the absence of a continuous cold cap. The current waste disposal path for the WTP-SW is to recycle it to the supplemental LAW treatment to avoid a large steady state accumulation in the pretreatment-vitrification loop. Fluidized Bed Steam Reforming (FBSR) offers a moderate temperature (700-750°C) continuous method by which LAW and/or WTP-SW wastes can be processed irrespective of whether they contain organics, nitrates, sulfates/sulfides, chlorides, fluorides, volatile radionuclides or other aqueous components. The FBSR technology can process these wastes into a crystalline ceramic (mineral) waste form. The mineral waste form that is produced by co-processing waste with kaolin clay in an FBSR process has been shown to be as durable as LAW glass. Monolithing of the granular FBSR product, which is one of the objectives of this current study, is being investigated to prevent dispersion during transport or burial/storage but is not necessary for performance. FBSR testing of a Hanford LAW simulant and a WTP-SW simulant at the pilot scale was performed by THOR Treatment Technologies, LLC at Hazen Research Inc. in April/May 2008. The Hanford LAW simulant was the Rassat 68 tank blend and the target concentrations for the LAW was increased by a factor of 10 for Sb, As, Ag, Cd, and Tl; 100 for Ba and Re (Tc surrogate); 1,000 for I; and 254,902 for Cs based on discussions with the DOE field office and the environmental regulators and an evaluation of the Hanford Tank Waste Envelopes A, B, and C. It was determined through the evaluation of the actual tank waste metals concentrations that some metal levels were not sufficient to achieve reliable detection in the off-gas sampling. Therefore, the identified metals concentrations were increased in the Rassat simulant processed by TTT at HRI to ensure detection and enable calculation of system removal efficiencies, product retention efficiencies, and mass balance closure without regard to potential results of those determinations or impacts on product durability response such as Toxicity Characteristic Leach Procedure (TCLP). A WTP-SW simulant based on melter off-gas analyses from Vitreous State Laboratory (VSL) was also tested at HRI in the 15-inch diameter Engineering Scale Test Demonstration (ESTD) dual reformer at HRI in 2008. The target concentrations for the Resource Conservation and Recovery Act (RCRA) metals were increased by 16X for Se, 29X for Tl, 42X for Ba, 48X for Sb, by 100X for Pb and Ni, 1000X for Ag, and 1297X for Cd to ensure detection by the analytic laboratory used for the demonstration and, as with the LAW simulant, without regard to potential impacts on product durability response such as TCLP. Details of the ESTD testing that was performed in April and May of 2008 are shown in Table 1 as the Operational Summary. The first eight tests were performed on LAW using a Denitration and Mineralization Reformer (DMR) temperature of 725°C and different ratios of clay:simulant of 675 g or 640 g of OptiKast clay per liter of simulant. The last four tests were performed on WTP-SW (originally referred to as LAW-Recycle) using a DMR temperature in the range of 680°C to 700°C with a mixture of 45/55 Sagger/OptiKast clay with 307 g clay/L simulant. The coal feed rates used in the FBSR processing are also shown as noted at the bottom of Table 1. It should be noted that the carbon feed rates are averaged over those periods.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Final Project Report CFA-14-6357: A New Paradigm for Understanding Multiphase Ceramic Waste Form Performance

This project fabricated model multiphase ceramic waste forms with processing-controlled microstructures followed by advanced characterization with synchrotron and electron microscopy-based 3D tomography to provide elemental and chemical state-specific information resulting in compositional phase maps of ceramic composites. Details of 3D microstructural features were incorporated into computer-based simulations using durability data for individual constituent phases as inputs in order to predict the performance of multiphase waste forms with varying microstructure and phase connectivity.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Crystalline Ceramic Waste Forms: Report Detailing Data Collection In Support Of Potential FY13 Pilot Scale Melter Test

The research conducted in this work package is aimed at taking advantage of the long term thermodynamic stability of crystalline ceramics to create more durable waste forms (as compared to high level waste glass) in order to reduce the reliance on engineered and natural barrier systems. Durable ceramic waste forms that incorporate a wide range of radionuclides have the potential to broaden the available disposal options and to lower the storage and disposal costs associated with advanced fuel cycles. Assemblages of several titanate phases have been successfully demonstrated to incorporate radioactive waste elements, and the multiphase nature of these materials allows them to accommodate variation in the waste composition. Recent work has shown that they can be successfully produced from a melting and crystallization process. The objective of this report is to summarize the data collection in support of future melter demonstration testing for crystalline ceramic waste forms. The waste stream used as the basis for the development and testing is a combination of the projected Cs/Sr separated stream, the Trivalent Actinide - Lanthanide Separation by Phosphorous reagent Extraction from Aqueous Komplexes (TALSPEAK) waste stream consisting of lanthanide fission products, the transition metal fission product waste stream resulting from the transuranic extraction (TRUEX) process, and a high molybdenum concentration with relatively low noble metal concentrations. The principal difficulties encountered during processing of the “reference ceramic” waste form by a melt and crystallization process were the incomplete incorporation of Cs into the hollandite phase and the presence of secondary Cs-Mo non-durable phases. In the single phase hollandite system, these issues were addressed in this study by refining the compositions to include Cr as a transition metal element and the use of Ti/TiO 2 buffer to maintain reducing conditions. Initial viscosity studies of ceramic waste forms indicated that the pour spout must be maintained above 1400°C to avoid flow blockages due to crystallization. In-situ electron irradiations simulate radiolysis effects indicated hollandite undergoes a crystalline to amorphous transition after a radiation dose of 10 13 Gy which corresponds to approximately 1000 years at anticipated doses (2 10 10 -2 10 11 Gy). Dual-beam ion irradiations employing light ion beam (such as 5 MeV alpha) and heavy ion beam (such as 100 keV Kr) studies indicate that reference ceramic waste forms are radiation tolerant to the α-particles and β-particles, but are susceptible to a crystalline to amorphous transition under recoil nuclei effects. A path forward for refining the processing steps needed to form the targeted phase assemblages is outlined in this report. Processing modifications including melting in a reducing atmosphere with the use of Ti/TiO 2 buffers, and the addition of Cr to the transition metal additives to facilitate Cs-incorporation in the hollandite phase. In addition to melt processing, alternative fabrication routes are being considered including Spark Plasma Sintering (SPS) and Hot Isostatic Pressing (HIP).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

Control of the Reduction/Oxidation State of the High Level Waste Form in the Defense Waste Processing Facility at the Savannah River Site - 12451

The Defense Waste Processing Facility at the Savannah River Site treats and immobilizes High Level Waste into a durable borosilicate glass for safe, permanent storage. Among the parameters important to the overall process is the reduction/oxidation state of the glass. Control strategies for the reduction/oxidation state of the glass have been developed and evolved over the last 20 years to provide a current strategy which is both accurate and robust. A description of the methodologies implemented to control, evaluate, and validate the reduction/oxidation state of the glass is discussed along with the impact of future facility changes (made to pursue a more aggressive program to empty High Level Waste tanks at the Savannah River Site) on the reduction/oxidation state of glass. The reduction/oxidation state of the glass is an important parameter to the operation of the Defense Waste Processing Facility. Control, evaluation, and validation of the reduction/oxidation state of the glass is of particular importance for melter operation. Melter feed which is too oxidizing can negatively impact production, while feed which is too reducing can result in catastrophic failure of the melter. Control strategies for the reduction/oxidation state of the glass have been developed and evolved over the last 20 years to provide a current strategy which is both accurate and robust. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Corrosion Behavior of Developmental Iron Phosphate Waste Forms: FY20 Status Report

Corrosion tests were performed to assess the dissolution behaviors of developmental iron phosphate materials that are being evaluated as an alternative to the glass-bonded sodalite ceramic waste form to immobilize salt wastes from electrochemical separation operations. Production of the iron phosphate waste form results in dehalogenation of the waste salt that increases potential waste loading. Ferric oxide is added to the waste form to increase the chemical durability. Tests conducted to optimize the amount of ferric oxide used to make the waste form are summarized in this report. Specimens prepared from materials made with between 17 and 34 mole % ferric oxide were subjected to two test methods to assess the effects of the iron content on the relative durability and retention of immobilized salt cations including cesium, strontium, and neodymium. Modified ASTM C1308 test results show the intrinsic durability was significantly higher for materials made with 27 mol % ferric oxide. Further additions generated iron-rich inclusion phases and did not improve durability. Modified ASTM C1285 test results show that solution feedback attenuates the dissolution rate similar to what is observed for borosilicate glasses within seven days. Analyses of test specimens after 42 days show no evidence that surface layers formed. Combined with previous studies addressing the salt loading, a composition region of the ammonium (di)hydrogen phosphate-waste salt-ferric oxide ternary has been defined for durable waste form compositions. Materials with that composition can be used to determine optimal processing conditions, assess long-term performance, and parameterize a degradation model.

Stariha, S. A.↗
DOE OSTI2013

Separations and Waste Forms Research and Development: FY 2012 Accomplishments Report

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) 2012 accomplishments report highlights the results of the research and development (R&D) efforts performed within SWFC in FY 2012. Each section contains a high level overview of the activities, results, technical point of contact and applicable references. 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 2012. The campaign continued to maintain a balanced effort between scientific investigations to underpin advanced separations and waste forms technologies and engineering advancement to maintain relevance and focus on implementable technologies. There was an increased focus on development of new tools and methods to develop a better fundamental understanding of separation processes and waste form behavior and this effort is closely tied to modeling and simulation activities.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Alternative Electrochemical Salt Waste Forms, Summary of FY11-FY12 Results

The Fuel Cycle Research and Development Program, sponsored by the U.S. Department of Energy Office of Nuclear Energy, is currently investigating alternative waste forms for wastes generated from nuclear fuel processing. One such waste results from an electrochemical separations process, called the “Echem” process. The Echem process utilizes a molten KCl-LiCl salt to dissolve the fuel. This process results in a spent salt containing alkali, alkaline earth, lanthanide halides and small quantities of actinide halides, where the primary halide is chloride with a minor iodide fraction. Pacific Northwest National Laboratory (PNNL) is concurrently investigating two candidate waste forms for the Echem spent-salt: high-halide minerals (i.e., sodalite and cancrinite) and tellurite (TeO2)-based glasses. Both of these candidates showed promise in fiscal year (FY) 2009 and FY2010 with a simplified nonradioactive simulant of the Echem waste. Further testing was performed on these waste forms in FY2011 and FY2012 to assess the possibility of their use in a sustainable fuel cycle. This report summarizes the combined results from FY2011 and FY2012 efforts.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Material Recover and Waste Form Development--2016 Accomplishments

The Material Recovery and Waste Form Development (MRWFD) Campaign under the U.S. Department of Energy (DOE) Fuel Cycle Technologies (FCT) Program 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). This MRWFD accomplishments report summarizes the results of the research and development (R&D) efforts performed within MRWFD in Fiscal Year (FY) 2016. Each section of the report contains an overview of the activities, results, technical point of contact, applicable references, and documents produced during the FY. This report briefly outlines campaign management and integration activities but primarily focuses on the many technical accomplishments of FY 2016. The campaign continued to use an engineering-driven, science-based approach to maintain relevance and focus.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Apatite Based Ceramic Waste Forms for Immobilization of Radioactive Iodine, An Overview - 16155

Apatite based ceramic waste forms are the most promising candidates for immobilization of radioactive iodine (I-129). While lead-vanadium based apatites have been studied for a long time, significant progress has been made in recent years in the domain of calcium-phosphate based apatites for iodine immobilization. This article is an attempt to present an overview of the progress that has been made in terms of developing novel synthesis and processing routes to contain iodine in apatite based ceramic waste forms. The challenges associated with current state-of-the art waste form synthesis techniques along with novel routes to synthesize these waste forms at low temperatures have also been discussed. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2014

Waste Stream Treatment and Waste Form Fabrication for Pyroprocessing of Used Nuclear Fuel

This report summarizes treatment and waste form options being evaluated for waste streams resulting from the electrochemical/pyrometallurgical (pyro ) processing of used oxide nuclear fuel. The technologies that are described are South Korean (Republic of Korea – ROK) and United States of America (US) ‘centric’ in the approach to treating pyroprocessing wastes and are based on the decade long collaborations between US and ROK researchers. Some of the general and advanced technologies described in this report will be demonstrated during the Integrated Recycle Test (IRT) to be conducted as a part of the Joint Fuel Cycle Study (JFCS) collaboration between US Department of Energy (DOE) and ROK national laboratories. The JFCS means to specifically address and evaluated the technological, economic, and safe guard issues associated with the treatment of used nuclear fuel by pyroprocessing. The IRT will involve the processing of commercial, used oxide fuel to recover uranium and transuranics. The recovered transuranics will then be fabricated into metallic fuel and irradiated to transmutate, or burn the transuranic elements to shorter lived radionuclides. In addition, the various process streams will be evaluated and tested for fission product removal, electrolytic salt recycle, minimization of actinide loss to waste streams and waste form fabrication and characterization. This report specifically addresses the production and testing of those waste forms to demonstrate their compatibility with treatment options and suitability for disposal.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗