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At least 109 records · Page 6
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Closed Fuel Cycle Waste Treatment Strategy

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.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Phase 1 Testing Results of Immobilization of WTP Effluent Management Facility Evaporator Bottoms Core Simulant

The Hanford Waste Treatment and Immobilization Plant (WTP) Low Activity Waste (LAW) vitrification facility will generate an aqueous condensate recycle stream (LAW Melter Off-Gas Condensate) from the off-gas system. The baseline plan for disposition of this stream during full WTP operations is to send it to the WTP Pretreatment Facility, where it will be blended with LAW, concentrated by evaporation and recycled to the LAW vitrification facility. However, during the Direct Feed LAW (DFLAW) scenario, planned disposition of this stream is to evaporate it in a new evaporator in the Effluent Management Facility (EMF) and then return it to the LAW melter. It is important to understand the composition of the effluents from the melter and new evaporator so that the disposition of these streams can be accurately planned and accommodated. Furthermore, alternate disposition of this stream would eliminate recycling of problematic components, and would enable less integrated operation of the LAW melter and the Pretreatment Facilities. Alternate disposition would also eliminate this stream from recycling within WTP when it begins operations and would decrease the LAW vitrification mission duration and quantity of glass waste. This LAW Melter Off-Gas Condensate stream will contain components that are volatile at melter temperatures and are problematic for the glass waste form, such as halides and sulfate, along with entrained, volatile, and semi-volatile metals, such as Hg, As, and Se. Because this stream will recycle within WTP, these components accumulate in the Melter Condensate stream, exacerbating their impact on the number of LAW glass containers that must be produced. Diverting the stream reduces the halides and sulfate that get recycled to the melter, and is a key objective of this work. This overall program examines the potential treatment and immobilization of this stream to enable alternative disposal. The objective of earlier tasks was to formulate and prepare a simulant of the LAW Melter Off-gas Condensate expected during DFLAW operations and use it in evaporator testing to predict the composition of the effluents from the Effluent Management Facility (EMF) evaporator to aid in planning for their disposition. The objective of this task was to test immobilization options for this evaporator bottoms aqueous stream. This document describes the method used to formulate a simulant of this EMF evaporator bottoms stream, immobilize it, and determine if the immobilized waste forms meet disposal criteria.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Investigating the Durability of Iodine Waste Forms in Dilute Conditions

To prevent the release of radioiodine during the reprocessing of used nuclear fuel or in the management of other wastes, many technologies have been developed for iodine capture. The capture is only part of the challenge as a durable waste form is required to ensure safe disposal of the radioiodine. This work presents the first durability studies in dilute conditions of two AgI-containing waste forms: hot-isostatically pressed silver mordenite (AgZ) and spark plasma sintered silver-functionalized silica aerogel (SFA) iodine waste forms (IWF). Using the single-pass flow-through (SPFT) test method, the dissolution rates respective to Si, Al, Ag and I were measured for variants of the IWFs. By combining solution and solid analysis information on the corrosion mechanism neutral-to-alkaline conditions was elucidated. The AgZ samples were observed to have corrosion preferentially occur at secondary phases with higher Al and alkali content. These phases contained a lower proportion of I compared with the matrix. The SFA samples experienced a higher extent of corrosion at Si-rich particles, but an increased addition of Si to the waste led to an improvement in corrosion resistance. The dissolution rates for the IWF types are of similar magnitude to other Si-based waste form materials measured using SPFT.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2012

Redox Control For Hanford HLW Feeds

The principal objectives of this work were to investigate the effects of processing simulated Hanford HLW at the estimated maximum concentrations of nitrates and oxalates and to identify strategies to mitigate any processing issues resulting from high concentrations of nitrates and oxalates. This report provides results for a series of tests that were performed on the DM10 melter system with simulated C-106/AY-102 HLW. The tests employed simulated HLW feeds containing variable amounts of nitrates and waste organic compounds corresponding to maximum concentrations projected for Hanford HLW streams in order to determine their effects on glass production rate, processing characteristics, glass redox conditions, melt pool foaming, and the tendency to form secondary phases. Such melter tests provide information on key process factors such as feed processing behavior, dynamic effects during processing, processing rates, off-gas amounts and compositions, foaming control, etc., that cannot be reliably obtained from crucible melts.

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

The potential for New Mexico basalts to sequester CO2 with a focus on critical element mobility

This study investigates the potential for New Mexico basalts to serve as geologic CO₂ sequestration media through mineral carbonation. Batch-type experiments were conducted using crushed basalts from the Carrizozo and Taos regions at low PCO₂ (0.7–2 bar) and temperatures up to 40°C over 180 days. The experiments evaluated the kinetics of CO₂-water-rock interaction, cation release, and critical element mobility (Cu, Ni, Co, Zn). Solid and fluid samples were analyzed using SEM, EMPA, ICP-OES, and ICP-MS, and geochemical modeling was performed with GEM-Selektor and PHREEQC. Results indicated that up to 80% of the injected CO₂ was mineralized within 100 days, forming secondary carbonates such as siderite and ankerite. Additionally, significant mobilization of critical metals was observed, which has implications for resource recovery and mine-waste valorization in future CO₂ sequestration projects.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
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 OSTI2014

The Performance of Underground Radioactive Waste Storage Tanks at the Savannah River Site: A 60-Year Historical Perspective

The Savannah River Site produced weapons-grade materials for nearly 35 years between 1953 and 1988. The legacy of this production is nearly 37 million gallons of radioactive waste. Since the 1950s, the liquid waste has been stored in large, underground carbon steel waste tanks. During the past 20 years, the site has begun to process the waste so that it may be stored in vitrified and grout forms, which are more suitable for long-term storage. Over the history of the site, some tanks have experienced leakage of the waste to the secondary containment. This article is a review of the instances of leakage and corrosion degradation that the tanks and associated equipment have experienced since the first tanks were built. Furthermore, the activities that the site has taken to mitigate the degradation and manage the service life of the tank for its anticipated lifetime are reviewed.

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

The Reactor Graphite Handling. Problems and Solutions - 17208

The problem of industrial uranium-graphite reactors and NPP units decommissioning is one of the urgent tasks that require efficient and cost-effective solutions. To date, only Russia has already shut down and subject to decommissioning more than 120 objects. World-wide, more than 260,000 ton of irradiated graphite have arisen from the commercial nuclear power operations, research reactors, and defense production reactors. Under the circumstances the problem to develop cost-effective methods of the irradiated graphite management has appeared. Alternative disposal options for the irradiated graphite were discussed for at least 25-30 years. To date, various methods of irradiated graphite reprocessing have been proposed. They are based on the application of different physical and chemical processes, most of which involve incineration of the entire mass of graphite. However, they all have disadvantages, and do not solve the whole complex of problems. The primary goal of the research presented here was to carry out a study aimed at the development of a more simple technology (as compared to the incineration of the entire mass of graphite) that would provide for the graphite purification and allow the possibility of its transfer to the less hazardous waste category followed by the disposal of it in a near-surface repository. To achieve it is proposed to consider two directions, each of them reducing the cost of graphite reprocessing and the waste volume disposed of in the repository. The first alternative is based on the principle of partitioning radionuclides of different nature from the graphite, thus allowing to transfer it into the LLW category and minimize the volume of secondary waste products. The second option to change the reprocessing technology is the idea to use reactor graphite as the starting material to produce stable final waste form compositions intended for the long-term storage of radioactive waste including transuranic elements. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Experimental and Modeling of Low Pressure Plasma as a Complementary Process for Thermal Waste Treatment - 19270

Polymeric waste contaminated with radionuclides is a challenging wasteform in its raw state. This waste is characterized by its high volume to weight ratio, together with the hazard of its radiolysis and consequently, the generation of flammable and explosive gases (e.g. H2 and light hydrocarbons). Thus, minimizing waste volumes together with stabilizing the wasteform are of high significance in the treatment of contaminated polymers. Thermal decomposition of polymeric waste results in gas products, mainly hydrocarbon molecules (C2-C6), which cannot be released to the atmosphere for environmental reasons. Here, we study a secondary gas treatment reactor, using plasma formed by inductive radio-frequency (13.56 MHz RF), in order to decompose the hydrocarbon gas products into smaller molecules. As a case study, we compare the plasma treatment of gases from two sources: a mixed gaseous product collected from the pyrolytic decomposition of high-density polyethylene (HDPE), and industrial butane gas. We also examine the effects of systematic addition of Ar, N{sub 2} and Ar+N{sub 2} mixtures to the treated gases. In this work, we utilize optical emission spectroscopy (OES) to obtain in situ data from the region in the center of the inductive RF coil. The main emissions from discharges in the gases or their mixtures occur in the 200-700 nm spectral region. Their relative intensities depend strongly on the plasma excitation mode, N{sub 2} flow rate, and gas feed composition. The addition of N{sub 2} to the input gas enhanced hydrocarbon decomposition as compared to the addition of pure Ar. This can be attributed to decomposition by free-radical reactions rather than by particles collisions processes. We performed preliminary modeling of the plasma behavior from ignition to steady state, which takes less than 0.01 s under these conditions. These simulations will help us to better understand and optimize a variety of decomposition reactions. The results of this study suggest that coupling low pressure plasma to conventional thermal decomposition techniques can provide an effective treatment process for contaminated solid plastic waste. (authors)

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

Near-Complete Retention of Technetium During Vitrification of Hanford Low Activity Waste With Recycle - 17315

Technetium is one of the more volatile radionuclides for which retention in low activity waste (LAW) glass under the high temperature conditions during vitrification can vary depending on feed composition, feed chemistry, and melter operating parameters. High retention of technetium in the glass is desirable in order to minimize the fraction of technetium that is directed to secondary waste streams, which include liquid off-gas treatment system effluents. Consequently, test data are crucial to assess and underpin the assumptions underlying such mass balance projections, the compositions of secondary waste streams, and, in particular, the effects of recycle on increasing the incorporation of technetium in the glass products and minimizing the fraction in non-glass waste forms. Previous testing performed on a scaled melter system that incorporates prototypical off-gas treatment components and real-time recycle of the liquid effluents back to the melter feed demonstrated that recycle resulted in significant increases of the incorporation of technetium into the LAW glass product and behavior that was in good agreement with process models for the system. However, the maximum technetium retention achieved in those tests was less than predicted. Tests conducted in the present work demonstrated, for the first time, near-complete retention of technetium in the LAW glass product with recycle. Holdup of technetium in the vacuum evaporator, which concentrates the liquid effluents prior to their recycle to the melter feed, was identified as the primary reason for the previously observed shortfall. The fate of other volatile constituents was also assessed. In particular, in addition to technetium, rhenium, which is frequently used as a non-radioactive surrogate for technetium, was included in these tests. This allows for a direct comparison of the behavior of rhenium and technetium during vitrification and in the down-stream processes. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Enhanced LAW Glass Formulation Testing (Final Report)

The principal objective of this work was to extend the glass formulation methodology developed in the earlier work for Envelope A, B and C waste compositions for development of compliant glass compositions targeting five high sodium-sulfur waste loading regions. This was accomplished through a combination of crucible-scale tests, and tests on the DM10 melter system. The DM10 was used for several previous tests on LAW compositions to determine the maximum feed sulfur concentrations that can be processed without forming secondary sulfate phases on the surface of the melt pool. This melter is the most efficient melter platform for screening glass compositions over a wide range of sulfate concentrations and therefore was selected for the present tests. The tests were conducted to provide information on melter processing characteristics and off-gas data, including sulfur incorporation and partitioning. As described above, the main objective was to identify the limits of waste loading in compliant glass formulations spanning the range of expected Na 2 O and SO 3 concentrations in the LAW glasses.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Technitium Management at the Hanford Site

Long Abstract. Full Text. The Hanford tank waste contains approx 26,000 Ci of technetium-99 (Tc-99), the majority of which is in the supernate fraction. Tc-99 is a long-lived radionuclide with a half-life of approx 212,000 years and, in its predominant pertechnetate (TcO{sub 4}) form, is highly soluble and very mobile in the vadose zone and ultimately the groundwater. Tc-99 is identified as the major dose contributor (in groundwater) by past Hanford site performance assessments and therefore considered a key radionuclide of concern at Hanford. The United States Department of Energy (DOE) River Protection Project's (RPP) long-term Tc-99 management strategy is to immobilize the Tc-99 in a waste form that will retain the Tc-99 for many thousands of years. To achieve this, the RPP flowsheet will immobilize the majority of the Tc-99 as a vitrified low-activity waste product that will be ultimately disposed on site in the Integrated Disposal Facility. The Tc-99 will be released gradually from the glass at very low rates such that the groundwater concentrations at any point in time would be substantially below regulatory limits.The liquid secondary waste will be immobilized in a low-temperature matrix (cast stone) and the solid secondary waste will be stabilized using grout. Although the Tc-99 that is immobilized in glass will meet the release rate for disposal in IDF, a proportion is driven into the secondary waste stream that will not be vitrified and therefore presents a disposal risk. If a portion of the Tc-99 were to be removed from the Hanford waste inventory and disposed off-site, (e.g., as HLW), it could lessen a major constraint on LAW waste form performance, i.e., the requirement to retain Tc-99 over thousands of years and have a positive impact on the IDF Performance Assessment. There are several technologies available at various stages of technical maturity that can be employed for Tc-99 removal. The choice of technology and the associated efficacy of the technology are dependent on the chemical fonn of the technetium in the waste, the removal location in the tlowsheet. and the ultimate disposition path chosen for the technetium product. This paper will discuss the current plans for the management of the technetium present in the Hanford tank waste. It will present the risks associated with processing technetium in the current treatment flowsheet and present potential mitigation opportunities, the status of available technetium removal technologies, the chemical speciation of technetium in the tank waste, and the available disposition paths and waste forms for technetium containing streams.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Improving in situ liquid SEM imaging of particles

This work presents in situ imaging of synthesized boehmite (γ-AlOOH) particles ranging from 20 to 100 nm, suspended in liquid, in a vacuum compatible microfluidic sample holder using a scanning electron microscopy (SEM) under the high vacuum mode and highlights the advantage of in situ liquid imaging of colloids. Nanometer-sized boehmite particles in high-level radioactive wastes at the Hanford site are known to be difficult to dissolve and cause rheological problems for processing in the nuclear waste treatment plant. Therefore, it is important to characterize boehmite particles and understand how they form aggregates in the liquid state. Several technical advancements are made to optimize in situ liquid SEM chemical imaging resulting in the improved ability to obtain secondary electron (SE), backscattered electron (BSE) images, and energy dispersive X-ray spectroscopy (EDX) spectra. Moreover, our results show mixed particles could be studied and identified based on the particle shape and elemental composition using in situ SEM imaging and EDX. Thus, we provide a new and improved approach to observe the evolution of particle dispersion and stability in liquid under conditions similar to those in the waste tank.

47 OTHER INSTRUMENTATION↗
DOE OSTI2013

Behaviour of IPG waste forms bearing BaSO{sub 4} as the dominant sludge constituent generated from the treatment of water used for cooling the stricken power reactors

The great amount of water used for cooling the stricken power reactors at Fukushima Dai-ichi following the earthquake and tsunami of 11 March 2011 has resulted in accumulation of cooling water so-called the remaining water in some buildings. As the cooling water is subsequently contaminated by fission products (FPs) and some other radioactive substances, it is necessary to decontaminate this 'cooling water' to reduce the volume of liquid radioactive waste and to reuse it again for cooling the affected reactors. Some methods are applied to remove the radioactive substances from the cooling water. However, after treatments of water, there arises a secondary radioactive waste, the sludge. Steps are now taken to immobilize this sludge. In this paper, BaSO{sub 4}, as one of main constituents of the sludge, was chosen as an immobilizing target substance. The appropriate manufacturing condition of glass waste form for loading the sludge (BaSO{sub 4}) was studied and the chemical durability was evaluated by measuring the dissolution rate. For this experiment, the iron phosphate glass (IPG), which is known to possess a large loading capacity for a variety of chemical substances, was employed as the glass medium. Based on experimental results, it is evident that BaSO{sub 4} can be loaded into the IPG medium when it possesses the appropriate composition and melting temperature. During loading, BaSO{sub 4} converted into BaO, acting as a network modifier, which leads to enhanced stability of IPG. (authors)

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

Cesium incorporation in hollandite-rich multiphasic ceramic waste forms

Hollandite-rich multiphase waste form compositions processed by melt-solidification and spark plasma sintering (SPS) were characterized, compared, and validated for nuclear waste incorporation. Phase identification by x-ray diffraction (XRD) and electron back-scattered diffraction (EBSD) confirmed hollandite as the major phase present in these samples along with perovskite, pyrochlore and zirconolite. Distribution of selected elements observed by wavelength dispersive spectroscopy (WDS) maps indicated that Cs formed a secondary phase during SPS processing, which was considered undesirable. On the other hand, Cs partitioned into the hollandite phase in melt-processed samples. Further analysis of hollandite structure in melt-processed composition by selected area electron diffraction (SAED) revealed ordered arrangement of tunnel ions (Ba/Cs) and vacancies, suggesting efficient Cs incorporation into the lattice.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2011

Immobilization of 99-Technetium (VII) by Fe(II)-Goethite and Limited Reoxidation

During the nuclear waste vitrification process volatilized {sup 99}Tc will be trapped by melter off-gas scrubbers and then washed out into caustic solutions, and plans are currently being contemplated for the disposal of such secondary waste. Solutions containing pertechnetate [{sup 99}Tc(VII)O{sub 4}{sup -}] were mixed with precipitating goethite and dissolved Fe(II) to determine if an iron (oxy)hydroxide-based waste form can reduce Tc(VII) and isolate Tc(IV) from oxygen. The results of these experiments demonstrate that Fe(II) with goethite efficiently catalyzes the reduction of technetium in deionized water and complex solutions that mimic the chemical composition of caustic waste scrubber media. Identification of the phases, goethite + magnetite, was performed using XRD, SEM and TEM methods. Analyses of the Tc-bearing solid products by XAFS indicate that all of the Tc(VII) was reduced to Tc(IV) and that the latter is incorporated into goethite or magnetite as octahedral Tc(IV). Batch dissolution experiments, conducted under ambient oxidizing conditions for more than 180 days, demonstrated a very limited release of Tc to solution (2-7 {micro}g Tc/g solid). Incorporation of Tc(IV) into the goethite lattice thus provides significant advantages for limiting reoxidation and curtailing release of Tc disposed in nuclear waste repositories.

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

Shipping Grossly Dewatered resins in the RT-100 Type B Transportation Cask - 16068

Ion-exchange resins are used in the purification of coolant water in boiling water reactor nuclear power plants. The spent resins are a form of low-level or intermediate-level radioactive waste. These resins may have a relatively high specific activity level and are normally transported to a disposal facility within a secondary container placed in a Type B transportation cask. During transportation, hydrogen gas may build up in the cask due to radiolysis of the hydrogenous material in the contents. In the licensing process, a Type B cask is carefully analyzed to ensure that the maximum amount of hydrogen gas generated does not exceed 5%. The calculation to determine the mole fraction of hydrogen gas generated is a function of many variables, including the properties of the contents and the volume ratios of the contents. Because water generates hydrogen gas when it undergoes radiolysis, the amount of water present in the waste is normally limited. Typically, Type B transportation casks are allowed to carry dewatered resins, meaning less than 1% free-standing water by volume. To ensure compliance with this limitation, the dewatering process requires specialized equipment to pump the water from the secondary container. In an effort to enhance the usability of the RT-100 Type B transportation cask, Robatel Technologies re-analyzed the hydrogen gas generation calculation for contents including 'grossly dewatered' resins. Grossly dewatered resins are defined as resins containing less than 20% free-standing water by volume. This higher limit is made possible by a more thorough hydrogen gas generation calculation, which ensures that the flammability limit is not reached during shipment. In July 2015, The USNRC issued Revision 1 of the Certificate of Compliance for the Robatel Technologies RT-100 Type B transportation cask. This revision allows the cask to be used for grossly dewatered resins. With the revised Certificate of Compliance, the RT-100 is the first of its kind whose authorized contents include greater than 1% free liquids by volume. As a result of this change, the cask user can meet the 20% maximum free-standing water volume with a much simpler gravity-driven dewatering process. It is no longer necessary to have a complex dewatering system to ensure compliance with regulatory requirements. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2026

Effect of the form of chromium on HLW vitrification

The primary objective of the work described herein was to determine the effect of two forms of chromium reported to be abundant in the waste (Cr(OH) 3 and chromite) on spinel formation in the glass, feed properties, and feed processing characteristics during the vitrification process. A secondary objective of the work was to evaluate oxidants that influence the redox state of chromium in the glass, which has the potential to reduce spinel crystallization thereby allowing potential increases in waste loadings in HLW glasses with high chromium content. This was accomplished through a combination of crucible-scale and DM100 melter system tests.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗