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DOE OSTI2020

Immobilizing Pertechnetate in Ettringite via Sulfate Substitution

Until recently, a nuclear waste form did not exist for the immobilization and long-term storage of the anionic form of radionuclide technetium-99 (99Tc), pertechnetate (TcO 4 - ). Across international nuclear waste sites, the inability to capture TcO 4 - poses a threat to the local environment due to the anion’s long half-life and environmental mobility under environmentally oxic conditions. To combat this challenge, most waste forms would capture 99Tc using materials that reduce TcO 4 - to Tc(IV) species, e.g. 99Tc oxides or 99Tc sulfides. However, this is a short-lived solution unless reducing capacities can be regenerated to last over the long life expectancy of some nuclear waste forms (~10,000 years). In light of recent experimental observations gained from the development of cementitious waste forms and unconfirmed hypotheses from the literature, this work explores how ettringite, a common mineral formed in cementitious materials with demonstrated success in incorporating oxyanions, may be used to immobilize TcO 4 - directly and over the life-time of the waste form. Using batch precipitation experiments, solid phase characterization techniques, and ab initio molecular dynamics simulations, we demonstrate herein successful incorporation of TcO 4 - into ettringite via substitution for SO 4 2- within the ettringite crystal structure. Here, these results are first of their kind and confirm the hypothesis that ettringite acts as an immobilizing mineral within cementitious waste forms for the highly mobile TcO 4 - oxyanion, enhancing the overall performance of these waste forms over time.

36 MATERIALS SCIENCE↗

DOE OSTI2015

Supplemental Immobilization of Hanford Low-Activity Waste: Cast Stone Augmented Formulation Matrix Tests

More than 56 million gallons of radioactive and hazardous waste are stored in 177 underground storage tanks at the U.S. Department of Energy’s (DOE’s) Hanford Site in Washington State. The HLW will be vitrified in the HLW facility for ultimate disposal at an offsite federal repository. A portion (~35%) of the LAW will be vitrified in the LAW vitrification facility for disposal onsite at the Integrated Disposal Facility (IDF). The pretreatment and HLW vitrification facilities will have the capacity to treat and immobilize all of the wastes destined for those facilities. However, a second facility will be needed for the expected volume of LAW requiring immobilization. Cast Stone, a cementitious waste form, is being considered to provide the required additional LAW immobilization capacity. The Cast Stone waste form must be acceptable for disposal in the IDF. The Cast Stone waste form and immobilization process must be tested to demonstrate that the final Cast Stone waste form can comply with the waste acceptance criteria for the disposal facility and that the immobilization processes can be controlled to consistently provide an acceptable waste form product. A testing program was developed in fiscal year (FY) 2012 describing in detail the work needed to develop and qualify Cast Stone as a waste form for the solidification of Hanford LAW. A statistically designed test matrix was used to evaluate the effects of key parameters on the properties of the Cast Stone as it is initially prepared and after curing. For the processing properties, the water-to-dry-blend mix ratio was the most significant parameter in affecting the range of values observed for each property. The single shell tank (SST) Blend simulant also showed differences in measured properties compared to the other three simulants tested. A review of the testing matrix and results indicated that an additional set of tests would be beneficial to improve the understanding of the impacts noted in the Screening Matrix tests. A set of Cast Stone formulations were devised to augment the original screening test matrix and focus on the range of the test conditions. Fly ash and blast furnace slag were limited to either northwest or southeast and the salt solutions were narrowed to the Average and the SST Blend at the 7.8M Na concentration. To fill in the matrix, a mix ratio of 0.5 was added. In addition, two admixtures, Xypex Admix C-500 and Rheomac SF100 (silica fume), were added as an additional dry material binder in select compositions. As in the Screening Matrix, both fresh and cured properties were evaluated for the formulations. In this study, properties that were influenced by the W/DM ratio in the Screening Matrix; flow diameter, plastic viscosity, density, and compressive strength, showed consistent behavior with respect to W/DM. The leach index for highly soluble components, sodium and nitrate, were not influenced by changes in formulation or the admixtures. The leach index for both iodine and Tc-99 show an influence from the addition of the admixture, Xypex Admix C-500. Additional testing should be performed to further evaluate the influence of Xypex Admix C-500 on the leach index over a range of admixture concentrations, Cast Stone formulations, and curing and storage conditions.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2014

Technetium Immobilization Forms Literature Survey

Of the many radionuclides and contaminants in the tank wastes stored at the Hanford site, technetium-99 ( 99 Tc) is one of the most challenging to effectively immobilize in a waste form for ultimate disposal. Within the Hanford Tank Waste Treatment and Immobilization Plant (WTP), the Tc will partition between both the high-level waste (HLW) and low-activity waste (LAW) fractions of the tank waste. The HLW fraction will be converted to a glass waste form in the HLW vitrification facility and the LAW fraction will be converted to another glass waste form in the LAW vitrification facility. In both vitrification facilities, the Tc is incorporated into the glass waste form but a significant fraction of the Tc volatilizes at the high glass-melting temperatures and is captured in the off-gas treatment systems at both facilities. The aqueous off-gas condensate solution containing the volatilized Tc is recycled and is added to the LAW glass melter feed. This recycle process is effective in increasing the loading of Tc in the LAW glass but it also disproportionally increases the sulfur and halides in the LAW melter feed which increases both the amount of LAW glass and either the duration of the LAW vitrification mission or the required supplemental LAW treatment capacity.

technetium↗

DOE OSTI2013

The Production of Advanced Glass Ceramic HLW Forms using Cold Crucible Induction Melter

Cold Crucible Induction Melters (CCIMs) will favorably change how High-Level radioactive Waste (from nuclear fuel recovery) is treated in the 21st century. Unlike the existing Joule-Heated Melters (JHMs) currently in operation for the glass-based immobilization of High-Level Waste (HLW), CCIMs offer unique material features that will increase melt temperatures, increase throughput, increase mixing, increase loading in the waste form, lower melter foot prints, eliminate melter corrosion and lower costs. These features not only enhance the technology for producing HLW forms, but also provide advantageous attributes to the waste form by allowing more durable alternatives to glass. This paper discusses advantageous features of the CCIM, with emphasis on features that overcome the historical issues with the JHMs presently utilized, as well as the benefits of glass ceramic waste forms over borosilicate glass waste forms. These advantages are then validated based on recent INL testing to demonstrate a first-of-a-kind formulation of a non-radioactive ceramic-based waste form utilizing a CCIM.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

Waste Management Approach for a Potential Future U.S. Closed Fuel Cycle - 16233

The U.S. Department of Energy, Office of Nuclear Energy has chartered an effort to develop technologies to enable safe and cost effective recycle of commercial used nuclear fuel (UNF) in the U.S. Part of this effort includes the evaluation of exiting waste management technologies for effective treatment of wastes in the context of current U.S. regulations and development of waste forms and processes with significant cost and/or performance benefits over those existing. This study summarizes the results of these ongoing efforts with a focus on the highly radioactive primary waste streams. The primary streams considered and the recommended waste forms include: - Tritium separated from either a low volume gas stream or a high volume water stream. The recommended waste form is low-water cement in high integrity containers. - Iodine-129 separated from off-gas streams in aqueous processing. There are a range of potentially suitable waste forms. 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 separated from LWR fuel treatment off-gases and immobilized as a CaCO{sub 3} in a cement waste form. - Krypton-85 separated from LWR and SFR fuel treatment off-gases and stored as a compressed gas. - An aqueous reprocessing high-level waste (HLW) raffinate waste which is immobilized by the vitrification process in one of three forms: a single phase borosilicate glass, a borosilicate based glass ceramic, or a multi-phased titanate ceramic [e.g., synthetic rock (Synroc)]. - An undissolved solids (UDS) fraction from aqueous reprocessing of LWR fuel that is either included in the borosilicate HLW glass or is immobilized in the form of a metal alloy in the case of glass ceramics or titanate ceramics. - Zirconium-based LWR fuel cladding hulls and stainless steel (SS) fuel assembly hardware that are washed and super-compacted for disposal or as an alternative Zr purification and reuse (or disposal as low-level waste, LLW) by reactive gas separations. - Electrochemical process salt HLW which is immobilized in a glass bonded Sodalite waste form known as the ceramic waste form (CWF). - Electrochemical process UDS and SS cladding hulls which are 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. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2024

Bonding Environments and Radiation Stabilities of Phosphate Glasses

This report is a summary of some key take-aways presented during the Phosphate Round Table Workshop held in 2020, which was led by PNNL and ANL and funded by DOE-NE. Phosphate glasses have a wide range of commercial and industrial uses due to their unique optical, chemical, and physical properties. They are a candidate material for use as an advanced waste form matrix for immobilizing radioactive waste due to their unique ability to immobilize high fractions of alkali- and sulfur-rich streams in chemically durable glasses. This literature review provides an overview of the structure of phosphate glasses, waste-form related properties of interest (e.g., chemical durability and radiation stability), a summary of how specific composition ratios affect chemical durability (e.g., [Fe]/[P] ratio, [O]/[P] ratio, Fe 2+ /Fe 3+ ratio), as well different ways that phosphate glasses can be affected by radiation. All of these are of interest to the waste form community as phosphate glasses, primarily Fe-P-O glasses, are investigated for usage in immobilizing various types of nuclear waste including U.S. legacy wastes as well as a method for treating salt-based high-level wastes from molten salt reactors and pyroprocessing of used nuclear fuels.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2016

Waste Management Approach for a Potential Future U.S. Closed Fuel Cycle - 16233

The U.S. Department of Energy, Office of Nuclear Energy has chartered an effort to develop technologies to enable safe and cost effective recycle of commercial used nuclear fuel (UNF) in the U.S. Part of this effort includes the evaluation of exiting waste management technologies for effective treatment of wastes in the context of current U.S. regulations and development of waste forms and processes with significant cost and/or performance benefits over those existing. This study summarizes the results of these ongoing efforts with a focus on the highly radioactive primary waste streams. The primary streams considered and the recommended waste forms include: •Tritium separated from either a low volume gas stream or a high volume water stream. The recommended waste form is low-water cement in high integrity containers. •Iodine-129 separated from off-gas streams in aqueous processing. There are a range of potentially suitable waste forms. 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 separated from LWR fuel treatment off-gases and immobilized as a CaCO3 in a cement waste form. •Krypton-85 separated from LWR and SFR fuel treatment off-gases and stored as a compressed gas. •An aqueous reprocessing high-level waste (HLW) raffinate waste which is immobilized by the vitrification process in one of three forms: a single phase borosilicate glass, a borosilicate based glass ceramic, or a multi-phased titanate ceramic [e.g., synthetic rock (Synroc)]. •An undissolved solids (UDS) fraction from aqueous reprocessing of LWR fuel that is either included in the borosilicate HLW glass or is immobilized in the form of a metal alloy in the case of glass ceramics or titanate ceramics. •Zirconium-based LWR fuel cladding hulls and stainless steel (SS) fuel assembly hardware that are washed and super-compacted for disposal or as an alternative Zr purification and reuse (or disposal as low-level waste, LLW) by reactive gas separations. •Electrochemical process salt HLW which is immobilized in a glass bonded Sodalite waste form known as the ceramic waste form (CWF). •Electrochemical process UDS and SS cladding hulls which are 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.

Fuel Cycle↗

DOE OSTI2013

Cast Stone Formulation At Higher Sodium Concentrations

A low temperature waste form known as Cast Stone is being considered to provide supplemental Low Activity Waste (LAW) immobilization capacity for the Hanford site. Formulation of Cast Stone at high sodium concentrations is of interest since a significant reduction in the necessary volume of Cast Stone and subsequent disposal costs could be achieved if an acceptable waste form can be produced with a high sodium molarity salt solution combined with a high water to premix (or dry blend) ratio. The objectives of this study were to evaluate the factors involved with increasing the sodium concentration in Cast Stone, including production and performance properties and the retention and release of specific components of interest. Three factors were identified for the experimental matrix: the concentration of sodium in the simulated salt solution, the water to premix ratio, and the blast furnace slag portion of the premix. The salt solution simulants used in this study were formulated to represent the overall average waste composition. The cement, blast furnace slag, and fly ash were sourced from a supplier in the Hanford area in order to be representative. The test mixes were prepared in the laboratory and fresh properties were measured. Fresh density increased with increasing sodium molarity and with decreasing water to premix ratio, as expected given the individual densities of these components. Rheology measurements showed that all of the test mixes produced very fluid slurries. The fresh density and rheology data are of potential value in designing a future Cast Stone production facility. Standing water and density gradient testing showed that settling is not of particular concern for the high sodium compositions studied. Heat of hydration measurements may provide some insight into the reactions that occur within the test mixes, which may in turn be related to the properties and performance of the waste form. These measurements showed that increased sodium concentration in the salt solution reduced the time to peak heat flow, and reducing the amount of slag in the premix increased the time to peak heat flow. These observations may help to describe some of the cured properties of the samples, in particular the differences in compressive strength observed after 28 and 90 days of curing. Samples were cured for at least 28 days at ambient temperature in the laboratory prior to cured properties analyses. The low activity waste form for disposal at the Hanford Site is required to have a compressive strength of at least 500 psi. After 28 days of curing, several of the test mixes had mean compressive strengths that were below the 500 psi requirement. Higher sodium concentrations and higher water to premix ratios led to reduced compressive strength. Higher fly ash concentrations decreased the compressive strength after 28 days of curing. This may be explained in that the cementitious phases matured more quickly in the mixes with higher concentrations of slag, as evidenced by the data for the time to peak heat generation. All of the test mixes exhibited higher mean compressive strengths after 90 days of curing, with only one composition having a mean compressive strength of less than 500 psi. Leach indices were determined for the test mixes for contaminants of interest. The leaching performance of the mixes evaluated in this study was not particularly sensitive to the factors used in the experimental design. This may be beneficial in demonstrating that the performance of the waste form is robust with respect to changes in the mix composition. The results of this study demonstrate the potential to achieve significantly higher waste loadings in Cast Stone and other low temperature, cementitious waste forms. Additional work is needed to elucidate the hydration mechanisms occurring in Cast Stone formulated with highly concentrated salt solutions since these reactions are responsible for determining the performance of the cured waste form. The thermal analyses completed in this study provide some preliminary insight, although the limited range of the factors in the test matrix hindered the identification of individual component effects. Future work should involve broader factor ranges to identify the roles played by each of the components in the mix via thermal analyses, analytical microscopy, and characterization of phase formation.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2013

Cast Stone Formulation At Higher Sodium Concentrations

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2014

Cast Stone Formulation At Higher Sodium Concentrations

A low temperature waste form known as Cast Stone is being considered to provide supplemental Low Activity Waste (LAW) immobilization capacity for the Hanford site. Formulation of Cast Stone at high sodium concentrations is of interest since a significant reduction in the necessary volume of Cast Stone and subsequent disposal costs could be achieved if an acceptable waste form can be produced with a high sodium molarity salt solution combined with a high water to premix (or dry blend) ratio. The objectives of this study were to evaluate the factors involved with increasing the sodium concentration in Cast Stone, including production and performance properties and the retention and release of specific components of interest. Three factors were identified for the experimental matrix: the concentration of sodium in the simulated salt solution, the water to premix ratio, and the blast furnace slag portion of the premix. The salt solution simulants used in this study were formulated to represent the overall average waste composition. The cement, blast furnace slag, and fly ash were sourced from a supplier in the Hanford area in order to be representative. The test mixes were prepared in the laboratory and fresh properties were measured. Fresh density increased with increasing sodium molarity and with decreasing water to premix ratio, as expected given the individual densities of these components. Rheology measurements showed that all of the test mixes produced very fluid slurries. The fresh density and rheology data are of potential value in designing a future Cast Stone production facility. Standing water and density gradient testing showed that settling is not of particular concern for the high sodium compositions studied. Heat of hydration measurements may provide some insight into the reactions that occur within the test mixes, which may in turn be related to the properties and performance of the waste form. These measurements showed that increased sodium concentration in the salt solution reduced the time to peak heat flow, and reducing the amount of slag in the premix increased the time to peak heat flow. These observations may help to describe some of the cured properties of the samples, in particular the differences in compressive strength observed after 28 and 90 days of curing. Samples were cured for at least 28 days at ambient temperature in the laboratory prior to cured properties analyses. The low activity waste form for disposal at the Hanford Site is required to have a compressive strength of at least 500 psi. After 28 days of curing, several of the test mixes had mean compressive strengths that were below the 500 psi requirement. Higher sodium concentrations and higher water to premix ratios led to reduced compressive strength. Higher fly ash concentrations decreased the compressive strength after 28 days of curing. This may be explained in that the cementitious phases matured more quickly in the mixes with higher concentrations of slag, as evidenced by the data for the time to peak heat generation. All of the test mixes exhibited higher mean compressive strengths after 90 days of curing, with only one composition having a mean compressive strength of less than 500 psi. Leachability indices were determined for the test mixes for contaminants of interest. The leaching performance of the mixes evaluated in this study was not particularly sensitive to the factors used in the experimental design. This may be beneficial in demonstrating that the performance of the waste form is robust with respect to changes in the mix composition. The results of this study demonstrate the potential to achieve significantly higher waste loadings in Cast Stone and other low temperature, cementitious waste forms. Additional work is needed to elucidate the hydration mechanisms occurring in Cast Stone formulated with highly concentrated salt solutions since these reactions are responsible for determining the performance of the cured waste form. The thermal analyses completed in this study provide some preliminary insight, although the limited range of the factors in the test matrix hindered the identification of individual component effects. Future work should involve broader factor ranges to identify the roles played by each of the components in the mix via thermal analyses, analytical microscopy, and characterization of phase formation.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2015

Evaluation of Technetium Getters to Improve the Performance of Cast Stone

Cast Stone has been selected as the preferred waste form for solidification of aqueous secondary liquid effluents from the Hanford Tank Waste Treatment and Immobilization Plant (WTP) process condensates and low-activity waste (LAW) melter off-gas caustic scrubber effluents. Cast Stone is also being evaluated as a supplemental immobilization technology to provide the necessary LAW treatment capacity to complete the Hanford tank waste cleanup mission in a timely and cost effective manner. One of the major radionuclides that Cast Stone has the potential to immobilize is technetium (Tc). The mechanism for immobilization is through the reduction of the highly mobile Tc(VII) species to the less mobile Tc(IV) species by the blast furnace slag (BFS) used in the Cast Stone formulation. Technetium immobilization through this method would be beneficial because Tc is one of the most difficult contaminants to address at the U.S. Department of Energy (DOE) Hanford Site due to its complex chemical behavior in tank waste, limited incorporation in mid- to high-temperature immobilization processes (vitrification, steam reformation, etc.), and high mobility in subsurface environments. In fact, the Tank Closure and Waste Management Environmental Impact Statement for the Hanford Site, Richland, Washington (TC&WM EIS) identifies technetium-99 ( 99 Tc) as one of the radioactive tank waste components contributing the most to the environmental impact associated with the cleanup of the Hanford Site. The TC&WM EIS, along with an earlier supplemental waste-form risk assessment, used a diffusion-limited release model to estimate the release of different contaminants from the WTP process waste forms. In both of these predictive modeling exercises, where effective diffusivities based on grout performance data available at the time, groundwater at the 100-m down-gradient well exceeded the allowable maximum permissible concentrations for 99 Tc. (900 pCi/L). Recent relatively short-term (63 day) leach tests conducted on both LAW and secondary waste Cast Stone monoliths indicated that 99 Tc diffusivities were at or near diffusivities where the groundwater at the 100-m down-gradient well would exceed the allowable maximum permissible 99 Tc concentrations. There is, therefore, a need and an opportunity to improve the retention of Tc in the Cast Stone waste form. One method to improve the performance of the Cast Stone waste form is through the addition of “getters” that selectively sequester Tc inside Cast Stone.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2013

Radiation Stability of Benzyl Tributyl Ammonium Chloride Towards Technetium-99 Extraction

A closed nuclear fuel cycle combining new separation technologies along with generation III and generation IV reactors is a promising way to achieve a sustainable energy supply. But it is important to keep in mind that future recycling processes of used nuclear fuel (UNF) must minimize wastes, improve partitioning process, and integrate waste considerations into processes. New separation processes are being developed worldwide to complement the actual industrialized PUREX process which selectively separates U(VI) and Pu(IV) from the raffinate. As an example, low nitric acid concentration in the aqueous phase of a UREX based process will co-extract U(VI) and Tc(VII) by tri-n-butyl phosphate (TBP). Technetium (Tc-99) is recognized to be one of the most abundant, long-lived radiotoxic isotopes in UNF (half-life, t1/2 = 2.13 × 105 years), and as such, it is targeted in UNF separation strategies for isolation and encapsulation in solid waste forms for final disposal in a nuclear waste repository. Immobilization of Tc-99 by a durable solid waste form is a challenge, and its fate in new advanced technology processes is of importance. It is essential to be able to quantify and locate 1) its occurrence in any new developed flow sheets, 2) its chemical form in the individual phases of a process, 3) its potential quantitative transfer in any waste streams, and consequently, 4) its quantitative separation for either potential transmutation to Ru-100 or isolation and encapsulation in solid waste forms for ultimate disposal. Furthermore, as a result of an U(VI)-Tc(VII) co-extraction in a UREX-based process, Tc(VII) could be found in low level waste (LLW) streams. There is a need for the development of new extraction systems that would selectively extract Tc-99 from LLW streams and concentrate it for feed into high level waste (HLW) for either Tc-99 immobilization in metallic waste forms (Tc-Zr alloys), and/or borosilicate-based waste glass. Studies have been launched to investigate the suitability of new macrocompounds such as crown-ethers, aza-crown ethers, and resorcinarenes for the selective extraction of Tc-99 from nitric acid solutions. The selectivity of the ligand is important in evaluating potential separation processes and also the radiation stability of the molecule is essential for minimization of waste and radiolysis products. In this paper, we are reporting the extraction of TcO4- by benzyltributyl ammonium chloride (BTBA). Experimental efforts were focused on determining the best extraction conditions by varying the ligand’s matrix conditions and concentration, as well as varying the organic phase composition (i.e., diluent variation). Furthermore, the ligand has been investigated for radiation stability. The ?-irradiation was performed on the neat organic phases containing the ligand at different absorbed doses to a maximum of 200 kGy using external Co-60 source. Post-irradiation solvent extraction measurements will be discussed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2018

Development of Hot-Isostatic Pressing for Treatment of UK ILW Radioactive Wastes - 18280

Hot-Isostatic Pressing (HIPing) has been proposed as an improved method for the treatment of radioactive wastes in the UK. Sludges and slurries housed on the Sellafield site are a prime candidate for HIP treatment due to their large volumes and issues associated with the use of the baseline cementation strategy. Options have been developed by GeoRoc Ltd to co-process these wastes, including a detailed conceptual plant design. This paper will describe research and development aimed at demonstrating the viability of the concept presented in this design. This paper summarizes two studies performed to address the knowledge gaps identified for treatment of these wastes using HIP technology. The first study related to the poorly characterized nature of the wastes, which requires any treatment option to be flexible in terms of the chemical and physical properties of the waste feed. The second study involved demonstrating that HIPing can be effectively utilized at a scale applicable to treat a waste inventory estimated to be in excess of 5,000 m{sup 3}. The first section of this paper presents a summary of research carried out to determine where variation in the waste stream composition may affect the processing and wasteform produced when HIPing Magnox sludges. This study aimed to qualify where restrictions on the process envelope may exist due to poorly defined waste-stream characteristics. A baseline wasteform was selected with the aim of achieving significant waste volume reduction, while producing a physically and chemically stable wasteform. A glass-bonded magnesium-titanate/magnesium-silicate wasteform was selected as the preferred option for the Magnox sludge wastes. A series of studies investigated the densification and phase formation of waste-forms by altering key variables in the Magnox sludge simulant. To test the extremes of the processable waste loading, samples were produced and characterized at 100 wt% waste-loading for Magnox sludge, Site-Ion Exchange Plant (SIXEP) sand/clinoptilolite and a 50:50 mix of the two waste-streams. Further studies involved a) altering the waste loading to additive mass ratio (to simulate changes in the waste loading/sludge solids content), b) varying the particle size distribution from the datum to include only small or large particulates in the sludge feed; and c) introducing up to 20 wt% metallic Magnox alloy, organic species or carbonate materials (to determine how their inclusion may affect the processing and densification of the materials). A U{sub 3}O{sub 8} loaded sample was also produced using the baseline Magnox sludge formulation. All the tests performed were shown to produce monolithic, well densified waste-forms with substantial volume reductions. Minimal variations from the baseline were noted but the waste-forms produced with these deviations was determined to be of comparable quality to of the baseline product. U{sub 3}O{sub 8} was shown to behave in a similar fashion to the ZrO{sub 2} surrogate used to simulate it in the process envelope studies and was encapsulated in the magnesium titanate/magnesium silicate matrix. This work demonstrates that HIP technology can successfully produce waste-forms across the broad process envelope expected in the Magnox sludge inventory. It is also possible to produce monolithic waste-forms at 100 wt% waste loading from Magnox sludges, SIXEP sand clinoptilolite slurries or a co-processed mixture of the two. The second part of this paper summarizes a commercially relevant demonstration of HIP technology for the treatment of sludge and slurry wastes housed at the Sellafield site (UK). This study culminated in production of the largest HIPed wasteform successfully produced to date. This study was performed in a series of stages. Investigations into formulation and properties of waste-forms were performed at a lab scale to evaluate suitable wasteform systems and identify suitable combinations of additives and waste loadings. Scale up demonstrations were then performed in stages with custom-designed HIP canisters. Scale-up and characterization of key wasteform properties were successfully performed with initial canister volumes of 50 ml, 400 ml and 8 L. Suitable front-end unit operations to produce a granular, free-flowing calcined simulant were demonstrated at an engineering scale. The demonstration was successful in processing a HIP canister with an initial starting volume in excess of 100 L. A 75 % volume reduction from the raw sludge was achieved at scale while demonstrating that HIP is applicable at suitably large scale for immobilization of ILW waste. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2018

DFLAW Glass and Feed Qualifications for AP-107 to Support WTP Start-Up and Flow-Sheet Development (VSL-18R4500-1, Rev 0)

About 50 million gallons of high-level mixed waste is currently stored in underground tanks at The United States Department of Energy’s (DOE’s) Hanford site in the State of Washington. The Hanford Tank Waste Treatment and Immobilization Plant (WTP) will provide DOE’s Office of River Protection (ORP) with a means of treating this waste by vitrification for subsequent disposal. The tank waste will be separated into low- and high-activity waste fractions, which will then be vitrified respectively into Immobilized Low Activity Waste (ILAW) and Immobilized High Level Waste (IHLW) products. The ILAW product will be disposed in an engineered facility on the Hanford site while the IHLW product is designed for acceptance into the national deep geological disposal facility for high-level nuclear waste. The ILAW and IHLW products must meet a variety of requirements with respect to protection of the environment before they can be accepted for disposal. The Low Activity Waste (LAW) at Hanford will be vitrified in a Joule Heated Ceramic Melter (JHCM) to produce a stable product for safe disposal. Although present in relatively small amounts, a few key constituents in the LAW streams play particularly important roles. Sulfur, chlorine, and fluorine can have significant impacts on LAW glass formulation and waste loadings, while technetium and iodine are important for waste form performance, with technetium being the constituent of most concern with respect to waste disposal in the Integrated Disposal Facility (IDF). The IDF is designed to accept ILAW, secondary waste associated with ILAW production, and other on-site Hanford Non-Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) waste. As a result of volatilization of these species during the vitrification process, they are also present in significant amounts in the liquid effluents from the off-gas treatment system. Understanding the fraction of each constituent incorporated in the LAW glass or in WTP liquid effluents is critical for assessing the performance of waste forms to be disposed in the IDF. Liquid effluents from the LAW vitrification process that are not recycled will likely be solidified to produce a secondary waste form that is not as chemically durable as glass. Therefore, understanding partitioning of the key constituents between the glass and non-glass waste forms is a significant factor in the Performance Assessment (PA) for the Hanford IDF. Clearly, it is preferable to incorporate as much of the technetium into the glass as possible because of the superior chemical durability of that waste form. The objective of the present work is to improve the understanding of the partitioning of technetium and other key waste constituents between the glass and non-glass waste forms.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2020

Performance Criteria for Capture and/or Immobilization Technologies (Revision 1)

The capture and subsequent immobilization of regulated volatile radionuclides from the off-gas streams of a used nuclear fuel (UNF) reprocessing facility has been a topic of substantial research interest for the US Department of Energy and its international counterparts. Removal of specific radionuclides from the plant effluent streams before discharge to the environment is required to meet regulations set forth by the US Environmental Protection Agency. Upon removal, the radionuclides, as well as associated sorbents that cannot be regenerated in a cost-effective manner, are destined for conversion to a waste form. Research in separation and capture methodologies has included a wide range of technology types, and studies of waste forms are correspondingly diverse. In considering the future development and implementation of both sorbents and waste forms, it is necessary to identify benchmark measures of performance to objectively evaluate each sorbent system or waste form. Sets of performance criteria and associated metrics have been developed for sorbent and waste form evaluation. These criteria address physical, radiological, and chemical characteristics, technical practicality, technical maturity, cost, and, for sorbents, system performance. The criteria and metrics appear to be robust and should be applicable despite the eventual waste classification (as either high- or low-level waste). They are flexible enough to address both aqueous reprocessing and electrochemical reprocessing of UNF. These criteria sets can serve as tools to evaluate performance at multiple stages within the development process, and in this revision (Revision 1) they have been used to assess technologies relating to krypton/xenon separations and iodine capture from off-gas streams arising from UNF reprocessing. Assessment of krypton/xenon separations using engineered forms of two zeolite minerals (silver mordenite and hydrogen mordenite in a polyacrylonitrile-based binder [AgZ-PAN/HZ-PAN]) found that the zeolite-based separation is relatively advanced in its development, but several key issues require resolution. First, desorption processes for both krypton and xenon require refinement to provide an understanding of the product purity that can be achieved. Second, adsorption rate data is needed in order to calculate the bed depth required for effective separation. Finally, it is strongly recommended that a technical review of krypton/xenon separation by AgZ-PAN/HZ-PAN be performed to synergize available data and assess the cost savings and operational benefits that may be realized from implementation of this technology. Assessment of metal organic frameworks (MOFs) for their use in the separation of krypton/xenon found that the ideal separation would be performed using a single-column system with a MOF selective for krypton over xenon. A robust research effort should work to identify a krypton-selective MOF designed to operate at temperatures of approximately 0°C or higher, which could be preferred over cryogenic krypton/xenon capture for used fuel reprocessing off-gas streams. In the case of the CaSDB-MOF (the most well-understood xenon sorbent to date), two issues are judged of high importance. First, xenon breakthrough capacity for the CaSDB-MOF in prototypical conditions should be determined. Preliminary research indicates that breakthrough may be near immediate, presenting a substantial obstacle in separative system design. Second, development of desorption methodology should be performed to determine regeneration time, energy requirements, and the product stream composition. Silver-based sorbents (AgZ and AgAero) for use in iodine capture from the dissolver off-gas were evaluated against the established criteria. These sorbents are significantly better understood for this application as a result of research efforts over the past decade. The potential implementation of AgAero at a large scale is hindered by its physical degradation by components of the dissolver off-gas stream. Less is known about the adsorption of iodine by these sorbents from other off-gas streams in the plant. Initial experimental efforts have been closely coordinated in an effort to understand organic iodine (such as would be found in the vessel off-gas) adsorption by AgZ and AgAero. Future work should expand this experimental program, and analysis of other reprocessing facility off-gas streams such as the vitrification off-gas stream should be conducted to better understand other potential applications for iodine sorbents. A review of iodine waste form development shows that this area is diverse and that multiple promising waste forms have been identified for the immobilization of radioactive iodine. Efforts related to the direct conversion of iodine sorbents (including AgZ and AgAero) should be continued because of the advantages of direct conversion in a waste management strategy and other sorbents should continue to be advanced as merited.

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DOE OSTI2020

Nuclear Criticality Safety Assessment of Criticality Control Containers without Moderation Control at the Waste Isolation Pilot Plant

The Waste Isolation Pilot Plant (WIPP) provides for safe, permanent disposal of government-owned transuranic (TRU) and TRU mixed wastes. Receipt and disposal of waste at the WIPP site began in March 1999. The Sandia report, Consideration of Nuclear Criticality When Disposing of Transuranic Waste at the Waste Isolation Pilot Plant, addressed potential nuclear criticality safety issues based on the projected inventory characteristics known at the time [1]. As designs for inventory, waste forms, and disposal packages have changed, new analyses have been performed, and updates have been made to address any potential effects to the WIPP safety basis. New analyses performed include Saylor 2017 [2] and Brickner 2019 [3], which address certain waste containers with specified loadings under post-closure conditions. Both examined several hypothetical scenarios and included analyses to bound (from a criticality potential standpoint) credible configurations that could occur at WIPP during the repository regulatory post-closure disposal time period for feature, event, and process (FEP) considerations—10,000 years. During this post-closure period at WIPP, the screening of FEPs is governed by the risk-based standards and implementing regulations of the US Environmental Protection Agency (EPA) (i.e., 40 CFR 191 and 40 CFR 194, respectively) [4,5]. An FEP screening can be based on either a low-consequence or low-probability rationale. A low-probability rationale includes either (a) a qualitative rationale that the FEP is not credible or (b) a quantitative demonstration that the probability is less than 10-4 in 104 years. In this evaluation, a qualitative lowprobability rationale of not credible is used by demonstrating that bounding configurations of the waste are not critical. The demonstration of subcriticality is through quantitative calculations, but a probability of criticality is not evaluated. Rather, the rationale for this evaluation is that bounding configurations with an effective neutron multiplication factor (keff) well below the upper subcriticality limit (USL) make criticality incredible. Reference [2] documented a nuclear criticality assessment of the WIPP repository for disposal of dilute surplus plutonium materials using the Dilute and Dispose Approach and packaging in criticality control overpacks (CCOs). The CCO is the waste disposal container recently designed to allow for up to 380 fissile gram equivalent (FGE) 239 Pu per drum, which is a higher fissile loading than typical waste containers. The CCO consists of a criticality control container (CCC) positioned by upper and lower plywood spacers within a standard 55 gal drum. The CCC is used to establish a geometry control for fissile materials during transportation and WIPP emplacement operations. The current WIPP waste acceptance criteria for CCO payloads limit beryllium to less than or equal to 1% by weight of the waste contents and require the waste form to be non-machine compacted. Reference [2] considered two scenario progressions—room closure from salt creep, hereafter referred to as the reconfigured dry scenario, and flooding with brine, hereafter referred to as the reconfigured wet scenario. The subsequent drying out of the reconfigured wet scenarios was also considered. For all scenarios, subcriticality was maintained when 50 g of B 4 C (acting as a neutron absorber) per CCC was intermixed within the plutonium disposition waste form. The analysis used a waste form description that limits the amount of moderation that could be present within the waste form (i.e., it limits the amount of water and polyethylene that could be present based on planned processing conditions). This analysis to evaluate increased limits on the amount of moderation that could be present was performed as a companion to Reference [2] to address concerns associated with verifying moisture and/or plastic contents of waste materials following packaging of dilute surplus plutonium in the CCO. To that end, this analysis used the models and methods from Reference [2] to evaluate a more generic base waste form consisting of water and polyethylene that is more similar (and nearly identical) to the generic waste forms utilized in other models/analyses supporting the TRU Package Transporter Model II (TRUPACTII) safety analysis [6] (all are without moderation controls). The waste form in this analysis uses a base mixture of 75% water and 25% polyethylene, the total amount of which is varied to determine the optimum moderation to fissile material (H/Pu) ratio. The fissile loading is maintained at up to 380 FGE 239 Pu (modeled as PuO 2 ) per CCO with an additional 545 g of beryllium (to bound the 1% by weight contents restriction) and 50 g of B 4 C intermixed per CCO. The beryllium content (1% by weight) is based on the total allowed waste weight (this does not include packaging and container weights). Figures ES-1 and ES-2 display summary results, showing that with this model including 50 g of B 4 C per CCO, the system keff remains under 0.85 for all moderator amounts and provides a significant margin against post-closure criticality under postulated bounding conditions for compaction. Figure ES-1 compares an infinite model with a room model at the initial emplacement spacing and under full radial compaction. Full radial compaction places each CCC in direct contact and does not credit any anticipated spacing associated with current post-closure geomechanical modeling of the repository [7]. The effects of variations in the H/Pu ratio were evaluated by varying the amount of the water/polyethylene component of the waste model, with fissile loading maintained at 380 239 Pu FGE. Similarly, Figure ES-2 illustrates how various amounts of B 4 C per CCO influence k eff at different radial compactions, all at the H/Pu ratio of 200 (in the room array model). Therefore, while the results from Saylor 2017 [2] modeled more realistic process limits associated with packaging of dilute surplus plutonium, this analysis demonstrates that limits on moderation (plastic and water content) are not necessary to ensure subcriticality in the WIPP repository, provided the requisite B 4 C absorber is present.

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DOE OSTI2013

Optimization of sintered AgI-Mordenite composites for {sup 129}I storage

The thermal processing of a proposed durable waste form for {sup 129}I was investigated. The waste form is a composite with a matrix of low-temperature sintering glass that encapsulates particles of AgI-mordenite. Ag-mordenite, an ion-exchanged zeolite, is being considered as a capture medium for gaseous {sup 129}I{sub 2} as part of a spent nuclear fuel reprocessing scheme under development by the US Department of Energy/Nuclear Energy (NE). The thermal processing of the waste form is necessary to increase the density of the glass matrix by viscous sintering so that the final waste form does not have any open porosity. Other processes that can also occur during the thermal treatment include desorption of chemisorbed I{sub 2}, volatilization of AgI and crystallization of the glass matrix. We have optimized the thermal processing to achieve the desired high density with higher AgI-mordenite loading levels and with minimal loss of iodine. Using these conditions, 625 C. degrees for 20 minutes, the matrix crystallizes to form a eulytite phase. Results of durability tests indicate that the matrix crystallization does not significantly decrease the durability in aqueous environments. (authors)

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DOE OSTI2015

Secondary Waste Simulant Development for Cast Stone Formulation Testing

Washington River Protection Solutions, LLC (WRPS) funded Pacific Northwest National Laboratory (PNNL) to conduct a waste form testing program to implement aspects of the Secondary Liquid Waste Treatment Cast Stone Technology Development Plan (Ashley 2012) and the Hanford Site Secondary Waste Roadmap (PNNL 2009) related to the development and qualification of Cast Stone as a potential waste form for the solidification of aqueous wastes from the Hanford Site after the aqueous wastes are treated at the Effluent Treatment Facility (ETF). The current baseline is that the resultant Cast Stone (or grout) solid waste forms would be disposed at the Integrated Disposal Facility (IDF). Data and results of this testing program will be used in the upcoming performance assessment of the IDF and in the design and operation of a solidification treatment unit planned to be added to the ETF. The purpose of the work described in this report is to 1) develop simulants for the waste streams that are currently being fed and future WTP secondary waste streams also to be fed into the ETF and 2) prepare simulants to use for preparation of grout or Cast Stone solid waste forms for testing.

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