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DOE OSTI2025

Thermal properties and Thermodynamic Equilibrium Modeling of Cementitious Waste Forms

INTRODUCTION Cementitious matrices are used in the US DOE complex and worldwide to solidify aqueous radioactive, hazardous, mixed salt solutions, and sludges to meet low-level radioactive waste (LLW) disposal requirements. Blended formulations are used for most waste form mix designs. Substitution of pozzolans for Class F fly ash has the potential to alter the processing properties and stabilization properties of the resulting waste forms because they add chemical and mineralogical complexity to the final material. The Savannah River Site saltstone waste form was selected as the test case material. CemGEMS software was chosen as the model for predicting hydrated cementitious phases assemblages and phase evolution. Isothermal calorimetry was used as the method for evaluating processing and properties of fresh, uncured waste forms. The test cases consisted of reference case saltstone, and five natural pozzolan substituted saltstone mixes.

Bustamante, Michael E. [Savannah River National La↗
DOE OSTI2015

Technetium Incorporation in Glass for the Hanford Tank Waste Treatment and Immobilization Plant

A priority of the United States Department of Energy (U.S. DOE) is to dispose of nuclear wastes accumulated in 177 underground tanks at the Hanford Nuclear Reservation in eastern Washington State. These nuclear wastes date from the Manhattan Project of World War II and from plutonium production during the Cold War. The DOE plans to separate high-level radioactive wastes from low activity wastes and to treat each of the waste streams by vitrification (immobilization of the nuclides in glass) for disposal. The immobilized low-activity waste will be disposed of here at Hanford and the immobilized high-level waste at the national geologic repository. Included in the inventory of highly radioactive wastes is large volumes of 99Tc (~9 × 10E2 TBq or ~2.5 × 104 Ci or ~1500 kg). A problem facing safe disposal of Tc-bearing wastes is the processing of waste feed into in a chemically durable waste form. Technetium incorporates poorly into silicate glass in traditional glass melting. It readily evaporates during melting of glass feeds and out of the molten glass, leading to a spectrum of high-to-low retention (ca. 20 to 80%) in the cooled glass product. DOE-ORP currently has a program at Pacific Northwest National Laboratory (PNNL), in the Department of Materials Science and Engineering at Rutgers University and in the School of Mechanical and Materials Engineering at Washington State University that seeks to understand aspects of Tc retention by means of studying Tc partitioning, molten salt formation, volatilization pathways, and cold cap chemistry. Another problem involves the stability of Tc in glass in both the national geologic repository and on-site disposal after it has been immobilized. The major environmental concern with 99Tc is its high mobility in addition to a long half-life (2.1×105 yrs). The pertechnetate ion (TcO4-) is highly soluble in water and does not adsorb well onto the surface of minerals and so migrates nearly at the same velocity as groundwater. Long-term corrosion of glass waste forms is an area of current interest to the DOE, but attention to the release of Tc from glass has been little explored. It is expected that the release of Tc from glass should be highly dependent on the local glass structure as well as the chemistry of the surrounding environment, including groundwater pH. Though the speciation of Tc in glass has been previously studied, and the Tc species present in waste glass have been previously reported, environmental Tc release mechanisms are poorly understood. The recent advances in Tc chemistry that have given rise to an understanding of incorporation in the glass giving rise to significantly higher single-pass retention during vitrification are presented. Additionally, possible changes to the baseline flowsheet that allow for relatively minor volumes of Tc reporting to secondary waste treatment will be discussed.

Kruger, Albert A.↗
DOE OSTI2016

Enhanced LAW Glass Correlation - Phase 1

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 a 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. Acceptable glass formulations for vitrification of Hanford low activity waste (LAW) must meet a variety of product quality, processability, and waste loading requirements. To this end, The Vitreous State Laboratory (VSL) at The Catholic University of America (CUA) developed and tested a number of glass formulations during Part A, Part B1 and Part B2 of the WTP development program. The testing resulted in the selection of target glass compositions for the processing of eight of the Phase I LAW tanks. The selected glass compositions were tested at the crucible scale to confirm their compliance with ILAW performance requirements. Duramelter 100 (DM100) and LAW Pilot Melter tests were then conducted to demonstrate the viability of these glass compositions for LAW vitrification at high processing rates.

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DOE OSTI2020

Characterization Strategy of a Limited Number of Radioactive Waste Effluent Tanks in View of Decommissioning a Large Number of Tanks - 20286

In order to organize the D and D of the former effluent treatment facility at the CEA Cadarache Center (France), the key data to gather through the facility characterization are the following:: volumes and masses, liquid/solid distribution and specific activity. By studying the treatment processes and the history of the facility, families of tanks have been defined and a reference tank chosen for investigation in each family. The intrusive investigations (video, residue thickness, dose rate..), the gamma spectrometry measurements carried out on site on the residue samples and radiochemical results collected in laboratories, led us to conclude that there was a relative radiological homogeneity (for specific activity and Am-241/Cs-137 ratio) within each tank. The next step consisted in extrapolating the specific activity evaluated for the reference tanks, to the other tanks of each family by using data such as geometry, dose rate or the function of the tanks within the process. Thus, a cartography of the 130 tanks of the facility is available. Finally, the conclusions drawn from the radiological homogeneity of the tanks and their affiliation to homogeneous, well-identified tank clusters lead to a simple and efficient operational approach for the D and D of the 130 tanks of the facility. Indeed, because of the residue characteristics (homogeneous activity and strong differentiation of Cs/Am ratio), a sample collected when opening the tank, and measured on site by gamma spectrometry will allow us to validate the operational design and the emptying techniques defined during the D and D study phase thanks to the characterization of a limited number of reference tanks. (authors)

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DOE OSTI2024

Value of abstraction in performance assessment – When is a higher level of detail necessary?

In this study, different approaches in performance assessment (PA) of the long-term safety of a repository for radioactive waste were examined. This investigation was carried out as part of the DECOVALEX-2023 project, an international collaborative effort for research and model comparison. One specific task of the DECOVALEX-2023 project was the Salt Performance Assessment Modelling task (Salt PA), which aimed at comparing various models and methods employed in the performance assessment of deep geological repositories in salt. In the context of the Salt PA task, three distinct teams from SNL (United States), Quintessa Ltd (United Kingdom), and GRS (Germany) examined the consequences of employing different levels of abstractions when modelling the repository's geometry and implementing various features and processes, using the example of a simple hypothetical repository structure in domal salt. Each team applied their own tools: PFLOTRAN (SNL), QPAC (Quintessa) and LOPOS (GRS). These differ essentially regarding numerical concept and degree of detail in the representation of the underlying physical processes. The discussion focused on when simplifications can be appropriately applied and what consequences result from them. Furthermore, it was explored when and if a higher level of fidelity in geometry or physical processes is required.

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DOE OSTI2014

FY13 Glycolic-Nitric Acid Flowsheet Demonstrations of the DWPF Chemical Process Cell with Simulants

Savannah River Remediation is evaluating changes to its current Defense Waste Processing Facility flowsheet to replace formic acid with glycolic acid in order to improve processing cycle times and decrease by approximately 100x the production of hydrogen, a potentially flammable gas. Higher throughput is needed in the Chemical Processing Cell since the installation of the bubblers into the melter has increased melt rate. Due to the significant maintenance required for the safety significant gas chromatographs and the potential for production of flammable quantities of hydrogen, eliminating the use of formic acid is highly desirable. Previous testing at the Savannah River National Laboratory has shown that replacing formic acid with glycolic acid allows the reduction and removal of mercury without significant catalytic hydrogen generation. Five back-to-back Sludge Receipt and Adjustment Tank (SRAT) cycles and four back-to-back Slurry Mix Evaporator (SME) cycles were successful in demonstrating the viability of the nitric/glycolic acid flowsheet. The testing was completed in FY13 to determine the impact of process heels (approximately 25% of the material is left behind after transfers). In addition, back-to-back experiments might identify longer-term processing problems. The testing was designed to be prototypic by including sludge simulant, Actinide Removal Product simulant, nitric acid, glycolic acid, and Strip Effluent simulant containing Next Generation Solvent in the SRAT processing and SRAT product simulant, decontamination frit slurry, and process frit slurry in the SME processing. A heel was produced in the first cycle and each subsequent cycle utilized the remaining heel from the previous cycle. Lower SRAT purges were utilized due to the low hydrogen generation. Design basis addition rates and boilup rates were used so the processing time was shorter than current processing rates.

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DOE OSTI2017

Experiments and Modeling to Support Field Test Design

Disposition of heat-generating nuclear waste (HGNW) remains a continuing technical and sociopolitical challenge. We define HGNW as the combination of both heat generating defense high level waste (DHLW) and civilian spent nuclear fuel (SNF). Numerous concepts for HGNW management have been proposed and examined internationally, including an extensive focus on geologic disposal (c.f. Brunnengräber et al., 2013). One type of proposed geologic material is salt, so chosen because of its viscoplastic deformation that causes self-repair of damage or deformation induced in the salt by waste emplacement activities (Hansen and Leigh, 2011). Salt as a repository material has been tested at several sites around the world, notably the Morsleben facility in Germany (c.f. Fahland and Heusermann, 2013; Wollrath et al., 2014; Fahland et al., 2015) and at the Waste Isolation Pilot Plant (WIPP) near Carlsbad, NM. Evaluating the technical feasibility of a HGNW repository in salt is an ongoing process involving experiments and numerical modeling of many processes at many facilities.

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DOE OSTI2015

Advances in Online Monitoring for Measurement within Nuclear Fuel Reprocessing Streams

Abstract – Significant effort has been focused on the back end of the nuclear fuel cycle including treatment and reuse of irradiated fuel. Solvent extraction reprocessing methods that have been developed often contain various steps, with each one tailored to the separation of specific radionuclides, which are highly dependent upon solution properties. The instrumentation used to monitor these processes must be robust, require little or no maintenance, and be able to withstand harsh environments such as high radiation fields and aggressive chemical matrices. Our group has been investigating the use of optical spectroscopy for the on-line monitoring of actinides, lanthanides, and acid strength within fuel reprocessing streams. This paper describes the application of the electronic and vibrational spectroscopic techniques for radiochemical process monitoring. In this context, our team experimentally assessed the potential of Raman and UV-vis spectroscopic techniques for online real-time monitoring of the U(VI)/nitrate ion/nitric acid and Pu(VI/IV)/Np(VI/V/IV)/Nd(III) in solutions relevant to used nuclear fuel reprocessing. These techniques demonstrated robust performance in the repetitive batch measurements of each analyte in a wide concentration range using simulant and commercial dissolved used fuel solutions. Currently this has been demonstrated on the macroscopic scale, using sample probes requiring large solution volumes. In an effort to minimize waste and reduce dose to personnel, we have modified this technique to allow measurement at the microfluidic scale using a Raman microprobe. Under the current sampling environment, Raman samples typically require upwards of 10 mL and larger. Using the new sampling system, we can sample volumes at 10 µL or less, which is a scale reduction of over 1,000 fold in sample size. This will significantly reduce the requirements for handling hazardous materials, and greatly reduce the costs associated with worker protection from exposure to radioactive and hazardous materials, and costs due to waste disposal.

process monitoring, U, nitric acid, microfluidic, ↗
DOE OSTI2015

Saltstone 4QCY14 TCLP Toxicity and UTS Results

A Saltstone Disposal Facility (SDF) waste form was prepared in the Savannah River National Laboratory (SRNL) from a Tank 50H sample and Z-Area premix material for the fourth quarter of calendar year 2014 (4QCY14). After a 47 day cure, a sample of the SDF waste form was collected, and shipped to a certified laboratory for Toxic Characteristic and Universal Treatment Standards (UTS) analysis. The metals analysis is performed using the Toxic Characteristic Leaching Procedure (TCLP) . The 4QCY14 saltstone sample results show that the saltstone is Resource Conservation Recovery Act (RCRA) nonhazardous, but is greater than the universal treatment standard for land disposal. The Saltstone Production Facility (SPF) and SDF were in a maintenance outage during the 4QCY14. Thus no processing or disposal of saltstone, as characterized by this 4QCY14 sample, occurred.

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DOE OSTI2017

Support for DF LAW Flowsheet Development Report [Final Report]

The primary objectives of this work were to develop a glass formulation for the direct feed LAW from Tank AP-105, collect data on primary off-gas system effluents to support the evaluation of potential direct feed LAW flowsheets, and determine the effect of the dilute direct feed LAW stream on processing rates and partitioning of waste components to the exhaust system. Development of a suitable glass formulation for the LAW Tank AP-105 composition made use of information from the ORP Enhanced LAW Glass Correlation that is currently under development. Subsequently, melter feeds at low solids contents indicative of direct feed LAW waste streams were processed on the DM100 to determine glass production rates and bubbling rates required to process the high water content feed. Liquid effluent solutions generated from prototypical primary off-gas system components (SBS and WESP) were collected while processing the LAW Tank AP-105 composition at nominal conditions through the DM10. These solutions were analyzed to determine their steady state compositions and the effects of feed composition and feed water content at nominal processing conditions.

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DOE OSTI2013

Results from One- and Two- Phase Fluid Flow Calculations within the Preliminary Safety Analysis of the Gorleben Site - 13310

Rock salt is one of the possible host rock formations for the disposal of high-level radioactive wastes in Germany. The Preliminary Safety Analysis of the Gorleben Site (Vorlaeufige Sicherheitsanalyse Gorleben, VSG) evaluates the long-term safety of a hypothetical repository in the salt dome of Gorleben, Germany. A mature repository concept and detailed knowledge of the site allowed a detailed process analysis within the project by numerical modeling of single-phase and two-phase flow. The possibility of liquid transport from the shafts to the emplacement drifts is one objective of the present study. Also, the implications of brine inflow on radionuclide transport and gas generation are investigated. Pressure build-up due to rock convergence and gas generation, release of volatile radionuclides from the waste and pressure-driven contaminant transport were considered, too. The study confirms that the compaction behavior of salt grit backfill is one of the most relevant factors for the hydrodynamic evolution of the repository and the transport of contaminants. Due to the interaction between compaction, saturation and pore pressure, complex flow patterns evolve. Emplacement drifts serve as gas sinks or sources at different times. In most calculation cases, the backfill reaches its final porosity after a few hundred years. The repository is then sealed and radionuclides can only be transported by diffusion in the liquid phase. Estimates for the final porosity of compacted backfill range between 0 % and 2 %. The exact properties of the backfill regarding single- and two-phase flow are not well known for this porosity range. The study highlights that this uncertainty has a profound impact on flow and transport processes over long time-scales. Therefore, more research is needed to characterize the properties of crushed salt grit at low porosities or to reduce the adverse effects of possible higher porosities by repository optimization. (authors)

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DOE OSTI2018

Proven Safe and Effective Process for the Treatment of Bulk Sodium at Shutdown Reactors and Research Facilities - 18601

Elemental sodium is an alkaline metal that reacts when exposed to air or water. This reaction is exothermic in nature and presents the risk of attendant fire as well as hydrogen generation. Within the United States sodium was used in liquid metal fast breeder reactors (LMFBR) during the 1960's and 70's as a heat transfer medium for commercial, experimental and prototype reactors as well as for nuclear energy related research and development experiments due to the element's good thermal conductivity. Shutdown of these reactors and the end of these experiments present a challenge to decommissioning teams as the wastes associated with each project have different characteristics affecting treatment based on the sodium's origin as well as different volumes requiring treatment. French CEA (Commissariat a l'Energie Atomique) and FRAMATOME worked together in the 1980's to develop a treatment process called the NOAH (NaOH - sodium hydroxide) process for the disposal of large volumes of bulk sodium. This paper provides an overview of the treatment process, details past uses of the technology and discusses the process' applications within the United States Department of Energy complex. (authors)

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DOE OSTI2015

Identification and prioritization of high risk deteriorating support infrastructure - 15475

Identification and prioritization for the re-capitalization of deteriorating waste processing and storage support infrastructure and information technology (IT) is essential for a safe, cost effective, and sustainable waste management program. Funding restraints have limited programs to performing only maintenance on aging infrastructure support systems whose operational costs, technological obsolescence, and reliability risks continue to grow without a disciplined and balanced re-capitalization program. Savannah River Site (SRS) comprises both Department of Energy Environmental Management and NNSA missions, and includes a lab, Savannah River National Laboratory. SRS has developed a crosscutting identification and prioritization tool and process that provides integration of all Site mission support infrastructure re-capitalization requirements and prioritizes into a single database that can be readily accessed for infrastructure investment decisions. This database, referred to as the SRS Critical Integrated Infrastructure Priority List (CIIPL) captures risk data that includes impacts toward safety, regulatory compliance, mission, and cost for inclusion in its prioritization process. Further, the CIIPL provides a status of each entered project's readiness for execution that includes funding source, reliability of current cost estimate, and an out year budget profile. The SRS CIIPL, created in 2009, was initially comprised of re-capitalization projects for the Site's common infrastructure (i.e. shared roads and utilities) that supported all Site missions and activities. As mission program funding became constrained, the value for a Site wide integrated prioritization tool became more apparent and the CIIPL grew to include all Site mission and tenant activities which included safeguards and security requirements. The SRS CIIPL has become a highly effective tool for achieving senior management situational awareness of the Site's most pressing support infrastructure and IT needs. The mission programs have used the CIIPL to prioritize and fund high risk projects within their own program budgets. This past year, the CIIPL was used as a budgeting and decision making tool to allocate funding for two common infrastructures high risk projects for the FY16 Site budget submission and has received laudatory comments from DOE headquarters review teams. The SRS CIIPL and its process can be a benefit to other waste management organizations by demonstrating a proven approach toward the integration of support infrastructure and IT re-capitalization requirements across diverse mission and tenant boundaries. The SRS CIIPL can provide senior management with an objective decision making tool that delivers a balanced evaluation of project requirements in terms of multiple risk criteria; safety, regulatory, mission, and cost impacts. (authors)

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DOE OSTI2011

Surface Geophysical Exploration – Compendium Document

This report documents the evolution of the surface geophysical exploration (SGE) program and highlights some of the most recent successes in imaging conductive targets related to past leaks within and around Hanford's tank farms. While it is noted that the SGE program consists of multiple geophysical techniques designed to (1) locate near surface infrastructure that may interfere with (2) subsurface plume mapping, the report will focus primarily on electrical resistivity acquisition and processing for plume mapping. Due to the interferences from the near surface piping network, tanks, fences, wells, etc., the results of the three-dimensional (3D) reconstruction of electrical resistivity was more representative of metal than the high ionic strength plumes. Since the first deployment, the focus of the SGE program has been to acquire and model the best electrical resistivity data that minimizes the influence of buried metal objects. Toward that goal, two significant advances have occurred: (1) using the infrastructure directly in the acquisition campaign and (2) placement of electrodes beneath the infrastructure. The direct use of infrastructure was successfully demonstrated at T farm by using wells as long electrodes (Rucker et al., 2010, 'Electrical-Resistivity Characterization of an Industrial Site Using Long Electrodes'). While the method was capable of finding targets related to past releases, a loss of vertical resolution was the trade-off. The burying of electrodes below the infrastructure helped to increase the vertical resolution, as long as a sufficient number of electrodes are available for the acquisition campaign.

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DOE OSTI2019

Characterization of Tc-99 Speciation and Chemical Composition of 241-AN-102 Tank Waste Supernatant - 19201

Technetium (Tc) in its isotopic form {sup 99}Tc is generated in high yields from the fission of {sup 235}U and {sup 239}Pu. A significant inventory of {sup 99}Tc remains at the U.S. Department of Energy's Hanford Site, and it is one of the most difficult contaminants to address. In the strongly alkaline solutions that are typical of Hanford tank waste, Tc exists as both pertechnetate (TcO{sub 4}{sup -} in oxidation state VII) and in reduced forms (oxidation state <VII) collectively known as non-pertechnetate species. Pertechnetate is a well-characterized anionic Tc species that can be removed from low-activity waste by anion exchange or other methods. There is no definitive information on the origin or a comprehensive description of the non-pertechnetate species in Hanford tanks. Designing strategies for effective management of Tc, including separation and immobilization, requires an understanding of the molecular structure of the non-pertechnetate species and their identification in actual tank waste samples. Developing this advanced understanding would facilitate development of new technologies for treating dissimilar Tc species. Toward this objective, a spectroscopic library of the Tc(I) [fac-Tc(CO){sub 3}]{sup +} and Tc(IV, VI, VII) compounds was generated using a range of techniques and applied to the characterization of actual 241-AN-102 tank waste supernatant. A sample of the 241-AN-102 tank waste (collected in 2013) was processed to remove {sup 137}Cs by spherical resorcinol-formaldehyde (sRF) ion exchange resin after the sodium concentration was adjusted to 5.6 M. The Cs-loaded sRF column was eluted with 0.45 M HNO{sub 3}. As-received AN-102, Cs-depleted AN-102 effluent, and sRF eluate fractions were comprehensively characterized for chemical composition and speciation of Tc using {sup 99}Tc nuclear magnetic resonance (NMR) spectroscopy and X-ray absorption spectroscopy. The results of these analyses demonstrated for the first time that non-pertechnetate Tc present in the 241-AN-102 tank waste is composed of several low-valent Tc species, including the Tc(I) [fac-Tc(CO){sub 3}L{sub 3}]{sup +}, Tc(IV), and possibly Tc(VI) compounds. This is the second experimental observation of the [fac-Tc(CO){sub 3}L{sub 3}]{sup +} species in the Hanford tank waste that highlights their importance for waste processing. In addition, the chemical composition of the 241-AN-102 tank waste was characterized with the focus on small chelators. (authors)

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DOE OSTI2020

Study on the Strategy for Final Disposal and Reuse of Decontamination Wastes after the Fukushima Daiichi NPP Accident - 20426

After the Fukushima Daiichi NPP accident, a large area of East Japan was contaminated by radio-Cs. To discuss the final disposal of the contaminated wastes off-site and the required development of related technologies, it is important to understand the volume and radiation concentration of the wastes after various treatments. For a scientific investigation purpose, a research committee has been formed by the Society for the Remediation of Radioactive Contamination in the Environment, which is independent from the Ministry of Environment. This committee's work is focused on discussing various scenarios for the final disposal considering a mass balance evaluation of the entire process, economic evaluations, and stakeholder consultation for final consensus. The current discussions of this committee are herein reported. (authors)

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DOE OSTI2017

Gas generation during waste treatment of acidic solutions from the dissolution of irradiated LEU targets for 99 Mo production

The goal of the Reduced Enrichment for Research and Test Reactors Program is to limit the use of high-enriched uranium (HEU) in research and test reactors by substituting low-enriched uranium (LEU) wherever possible. The work reported here documents our work to develop the calcining technologies and processes that will be needed for 99 Mo production using LEU foil targets and the Modified Cintichem Process. The primary concern with the conversion to LEU from HEU targets is that it would result in a five- to six-fold increase in the total uranium. This increase results in more liquid waste from the process. We have been working to minimize the increase in liquid waste and to minimize the impact of any change in liquid waste. Direct calcination of uranium-rich nitric acid solutions generates NO 2 gas and UO 3 solid. We have proposed two processes for treating the liquid waste from a Modified Cintichem Process with a LEU foil. One is an optimized direct calcination process that is similar to the process currently in use. The other is a uranyl oxalate precipitation process. The specific goal of the work reported here was to characterize and compare the chemical reactions that occur during these two processes. In particular, the amounts and compositions of the gaseous and solid products were of interest. A series of experiments was carried out to show the effects of temperature and the redox potential of the reaction atmosphere. The primary products of the direct calcination process were mixtures of U 3 O 8 and UO 3 solids and NO 2 gas. The primary products of the oxalate precipitation process were mixtures of U 3 O 8 and UO 2 solid and CO 2 gas. Higher temperature and a reducing atmosphere tended to favor quadrivalent over hexavalent uranium in the solid product. These data will help producers to decide between the two processes. In addition, the data can be used to design appropriate off-gas systems for pilot and production facilities.

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DOE OSTI2011

Iron Phosphate Glass-Containing Hanford Waste Simulant

Resolution of the nation’s high level tank waste legacy requires the design, construction, and operation of large and technically complex one-of-a-kind processing waste treatment and vitrification facilities. While the ultimate limits for waste loading and melter efficiency have yet to be defined or realized, significant reductions in glass volumes for disposal and mission life may be possible with advancements in melter technologies and/or glass formulations. This test report describes the experimental results from a small-scale test using the research scale melter (RSM) at Pacific Northwest National Laboratory (PNNL) to demonstrate the viability of iron phosphate-based glass with a selected waste composition that is high in sulfates (4.37 wt% SO3). The primary objective of the test was to develop data to support a cost-benefit analysis as related to the implementation of phosphate-based glasses for Hanford low activity waste (LAW) and/or other high-level waste streams within the U.S. Department of Energy complex. The testing was performed by PNNL and supported by Idaho National Laboratory, Savannah River National Laboratory, and Mo-Sci Corporation.

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