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 southeastern Washington State. The Hanford Tank Waste Treatment and Immobilization Plant (WTP) is being constructed to treat the wastes and immobilize them in a glass waste form. The WTP includes a pretreatment facility to separate the wastes into a small volume of high-level waste (HLW) containing most of the radioactivity and a larger volume of low-activity waste (LAW) containing most of the nonradioactive chemicals. The HLW will be converted to glass in the HLW vitrification facility for ultimate disposal at an offsite federal repository. At least a portion (~35%) of the LAW will be converted to glass in the LAW vitrification facility and will be disposed of onsite at the Integrated Disposal Facility (IDF). The pretreatment and HLW vitrification facilities will have the capacity to treat and immobilize the wastes destined for each facility. However, a second facility will be needed for the expected volume of additional LAW requiring immobilization. A cementitious waste form known as Cast Stone 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 waste acceptance criteria for the IDF disposal facility and that the immobilization processes can be controlled to consistently provide an acceptable waste form product. Further, the waste form must be tested to provide the technical basis for understanding the long term performance of the waste form in the IDF disposal environment. These waste form performance data are needed to support risk assessment and performance assessment (PA) analyses of the long-term environmental impact of the waste disposal in the IDF. A testing program was developed in fiscal year (FY) 2012 describing in some detail the work needed to develop and qualify Cast Stone as a waste form for the solidification of Hanford LAW (Westsik et al. 2012). Included within Westsik et al. (2012) is a section on the near-term needs to address Tri-Party Agreement Milestone M-062-40ZZ. The objectives of the testing program to be conducted in FY 2013 and FY 2014 are to: • Determine an acceptable formulation for the LAW Cast Stone waste form. • Evaluate sources of dry materials for preparing the LAW Cast Stone. • Demonstrate the robustness of the Cast Stone waste form for a range of LAW compositions. • Demonstrate the robustness of the formulation for variability in the Cast Stone process. • Provide Cast Stone contaminant release data for PA and risk assessment evaluations. The first step in determining an acceptable formulation for the LAW Cast Stone waste form is to conduct screening tests to examine expected ranges in pretreated LAW composition, waste stream concentrations, dry-materials sources, and mix ratios of waste feed to dry blend. A statistically designed test matrix will be used to evaluate the effects of these key parameters on the properties of the Cast Stone as it is initially prepared and after curing. The second phase of testing will focus on selection of a baseline Cast Stone formulation for LAW and demonstrating that Cast Stone can meet expected waste form requirements for disposal in the IDF. It is expected that this testing will use the results of the screening tests to define a smaller suite of tests to refine the composition of the baseline Cast Stone formulation (e.g. waste concentration, water to dry mix ratio, waste loading).

At the Savannah River Plant (SRP), a process has been developed for immobilizing high-level radioactive waste in a borosilicate glass. The waste is currently stored as soluble salts and insoluble solids. Insoluble waste as stored requires further processing before vitrification is possible. The processes required have been developed and demonstrated with actual waste. They include removal of aluminum in some waste, washing soluble salts out of the insoluble waste, and mercury stripping. Each of the processes and the results with actual SRP waste will be discussed. The benefits of each step will also be included.

The effectiveness of three dispersants to modify slurry rheology was examined using rheology measurements and laser scanning confocal microscopy (LSCM) in simulated waste solutions. All of the dispersants lowered the yield stress of the slurries below the baseline samples. The rheology curves were fitted reasonably to a Bingham Plastic model. The three-dimensional LSCM images of simulants showed distinct aggregates were greatly reduced after the addition of dispersants leading to a lowering of the yield stress of the simulated waste slurry solutions.

A high-level waste (HLW) remediation process scheduled to begin in 2007 at the Savannah River Site is the Modular Caustic Side Solvent Extraction (CSSX) Unit (MCU). The MCU will use a hydrocarbon solvent (diluent) containing a cesium extractant, a calix[4]arene compound, to extract radioactive cesium from caustic HLW. The resulting decontaminated HLW waste or raffinate will be processed into grout at the Saltstone Production Facility (SPF). The cesium containing CSSX stream will undergo washing with dilute nitric acid followed by stripping of the cesium nitrate into a very dilute nitric acid or the strip effluent stream and the CSSX solvent will be recycled. The Defense Waste Processing Facility (DWPF) will receive the strip effluent stream and immobilize the cesium into borosilicate glass. Excess CSSX solvent carryover from the MCU creates a potential flammability problem during DWPF processing. Bench-scale DWPF process testing was performed with simulated waste to determine the fate of the CSSX solvent components. A simple high performance liquid chromatography (HPLC) method was developed to identify the modifier (which is used to increase Cs extraction and extractant solubility) and extractant within the DWPF process. The diluent and triocytlamine (which is used to suppress impurity effect and ion-pair disassociation) were determined using gas chromatography mass spectroscopy (GCMS). To close the organic balance, two types of sample preparation methods were needed. One involved extracting aqueous samples with methylene chloride or hexane, and the second was capturing the off gas of the DWPF process using carbon tubes and rinsing the tubes with carbon disulfide for analysis. This paper addresses the development of the analytical methods and the bench-scale simulated waste study results.

Researchers at the Department of Energy's Savannah River Technology Center are using advanced microscopy techniques to understand the effects of trace organic chemical additions on nuclear waste slurry flow properties. Trace organic chemicals, surfactants (rheology modifiers), are being used in all types of industries to modify the flow properties of various commercial chemicals. Nuclear waste treatment at the Department of Energy's weapons production facilities, Savannah River Site and Hanford Reservation, is limited by the viscosity of the nuclear waste slurries as the material is processed through a variety of waste treatment and immobilization processes. The picture was taken using a laser scanning confocal microscope.

One factor limiting the production rate of radioactive waste immobilization processes is the rheological limitations imposed by the design of remotely maintained slurry process equipment (i.e. pumps, piping). Rheology modifiers (dispersants/flocculants) that could potentially decrease the yield stress and/or plastic viscosity of radioactive waste slurries were tested on simulated waste to determine which provided the largest decrease in yield stress and plastic viscosity. The goals of this study were to: 1) determine if trace levels of chemical additives could be used to reduce the rheological characteristics of radioactive waste slurries, 2) identify potential chemical additives for this work and future testing, 3) test a limited set of chemical additive candidates on simulated radioactive wastes, and 4) develop advanced techniques to visualize the internal slurry structure and particle-particle interaction within the slurry. Radioactive wastes slurries generated from the production of plutonium and tritium during the Cold War are being (and will be) immobilized in a borosilicate glass matrix using joule heated glass melters at various Department of Energy (DOE) facilities located across the United States. The maximum insoluble solids content of the waste slurries is limited by the design-basis rheological properties (e.g. the Bingham plastic yield stress and plastic viscosity) used to design the slurry handling systems. It is possible to modify the equipment used to mix, sample, and transport the waste slurry. However, the design and construction cost for any such modifications is very high due to the constraints (radiation, non-visible remote operation) imposed on the design and operation of radioactive waste processes. The rheology of two slurries with various rheology modifiers was evaluated using a conventional concentric cylinder rheometer (Haake Rheometer RS150). Only one rheology modifier of those tested was found to decrease the apparent viscosity of the waste slurry by any significant amount and several of the modifiers tested produced the opposite effect. Duramax D-3005 was found to decrease the Bingham Plastic yield stress of simulated radioactive waste slurries by approximately 18%. Selected slurries were further analyzed by a laser scanning confocal microscope. This technique allows the slurry to be analyzed in an unaltered condition. The microscope has the ability to make both two-dimensional pictures and three-dimensional representations of the slurry’s internal structure. The microscope allows the user to understand how particles are flocculated or dispersed throughout a concentrated suspension of heterogeneous simulated nuclear waste slurries.

Evaporation of High Level and Low Activity (HLW & LAW) radioactive wastes for the purposes of radionuclide separation and volume reduction has been conducted at the Savannah River and Hanford Sites for more than forty years. Additionally, the Savannah River Site (SRS) has used evaporators in preparing HLW for immobilization into a borosilicate glass matrix. The Hanford River Protection Project (RPP) is in the process of building the world’s largest radioactive waste treatment facility, Waste Treatment Plant (WTP), which will use evaporators to concentrate the liquid waste and plant recycles prior to immobilization into a borosilicate glass matrix. Radioactive waste is evaporated at each site using various evaporator designs (e.g., forced circulation, horizontal bent tube). While the equipment used to evaporate radioactive waste is relatively simple in design, the complexity in the evaporator processes in current service and in those currently in the design stages stems from the heterogeneous nature of the waste and the effects of seemingly minor components (e.g., Si) on the process. Aqueous electrolyte thermodynamic modeling and experiments have been conducted by the SRS Savannah River Technology Center (SRTC) in support of the SRS HLW and Defense Waste Processing Facility (DWPF) Evaporators and the Hanford RPP WTP. After 40 years of successful operation, accumulation of two solid phases (a nitrated aluminosilicate, Na8AL6Si6O24(NO3)2•4H2O and sodium diuranate, Na2U2O7) developed as an insoluble phase in the Savannah River Site (SRS) 2H evaporator in 1996. The aluminosilicate scale deposit caused the SRS 2-H evaporator to become completely inoperable by October 1999. Accumulation of the sodium diuranate phase on the aluminosilicate scale has caused criticality concerns. Modeling and experiments were conducted to develop a method to control the process chemistry in order to prevent the formation of aluminosilicate deposits in the future. The lessons learned from the development, design, and operation of the SRS waste treatment facilities and the currently operating 242-A Hanford HLW evaporators were applied by SRTC in support of the development and design of the Hanford WTP evaporators. Thermodynamic equilibrium modeling along with solubility and physical property experiments are being conducted to develop process control and flow sheet models. Additionally, lessons learned from the development of an advanced antifoam agent for the SRS vitrification process evaporators are being applied to the testing and development of an antifoam agent for the Hanford WTP evaporators. This paper will discuss the methodologies, results, and achievements of the SRTC evaporator development program that was conducted in support of the SRS and Hanford WTP evaporator processes. The “cross-pollination” and application of waste treatment technologies and methods between the Savannah River and Hanford Sites will be highlighted. The “cross-pollination” of technologies and methods is expected to benefit the Department of Energy’s Mission Acceleration efforts by reducing the overall cost and time for the development of the baseline waste treatment processes.

The beta-emitting radionuclide 137 Cs is precipitated from the Savannah River Site (SRS) high-level waste by adding sodium tetraphenylborate (TPB). The concentrated and washed precipitate slurry containing CsTPB is transferred to Defense Waste Processing Facility (DWPF) Salt Processing Cell and hydrolyzed to yield an aqueous product (PHA) free of benzene. The PHA is highly radioactive (1E+11 dpm) and contains some water-soluble organic compounds that need to be analyzed for process control and process history. High performance liquid chromatography (HPLC) methods were developed to analyze these compounds with instruments contained in unshielded fume hoods. Removing 137 Cs from PHA in a shielded cell prior to HPLC analyses is essential to minimize personnel radiation exposure. In this study ammonium molybdophosphate (Bio-Rad AMP-1) was used for removing 137 Cs in PHA (pH 3.8). With a two-step application of AMP slurry, a decontamination factor of 1E+4 was achieved for Cs removal. Two separate HPLC methods were used to analyze four polar organic compounds. The recovery of all the organic compounds were above 85%.

A team led by BNFL, Inc. was awarded the contract to remediate and immobilize the Hanford radioactive tank waste in support of the Hanford Tank Waste Remediation System program (TWRS). BNFL, Inc. is teamed with BNFL Engineering, LTD., Bechtel National, GTS Duratek, and Science Application International Corporation to develop and design integrated facilities for pretreatment and vitrification in support of the TWRS mission. This facility will pretreat and immobilize approximately 0.375 MT/day of high level waste (HLW, producing 1.5 MT/day of HLW glass) and approximately 4.5 MT/day of low activity waste (LAW, producing 30 MT/day of LAW glass) during Part B2 of the TWRS project. During the initial phase of Part B (FY98 ‐ FY00, B1), the technology is expected to be optimized to the point that it can be used as the basis for final design, construction, and operation of a vitrification facility in Hanford, Washington. As part of the overall Hanford Tank Waste Remediation System (TWRS) Project, BNFL, Inc. has contracted DOE-Westinghouse Savannah River Company’s Savannah River Technology Center (SRTC) to provide research and development services in characterization, pretreatment, and immobilization of actual Hanford tank wastes. Additionally, SRTC is developing design basis data using simulants of Hanford tank wastes in areas of ion exchange, filtration, precipitation, glass former blending, evaporation, and slurry mixing. This paper will provide an overview of the SRTC TWRS development program.

The original goal of this scoping study was to determine if the solubility of sodium and potassium tetraphenylborates in benzene was sufficiently large to justify designing and performing kinetic studies on a benzene-phase catalytic reaction.

This document is referenced in WSRC-TR-94-0556. This memorandum summarizes results of experimental work performed on the original IDMS PX6 melter feed, the remediated IDMS PX6 melter feed, and melter feeds produced in a laboratory simulation to duplicate the IDMS remediation as well as the experimental results on the caustic treatment to enhance the rheology. Characterization of the products of excess caustic addition and what steps to take if excess caustic is inadvertently added to the IDMS PX6 melter feed are also discussed

Removing aromatic carbon from an aqueous slurry of cesium-137 and other alkali tetraphenylborates by acid hydrolysis will be an important step in preparing high-level radioactive waste for vitrification at the Savannah River Site's Defense Waste Processing Facility (DWPF). Kinetic data obtained in bench-scale precipitate hydrolysis reactors suggest changes in operating parameters to improve product quality in the future plant-scale radioactive operation. The rate-determining step is the removal of the fourth phenyl group, i.e. hydrolysis of phenylboronic acid. Efforts to maximize this rate have established the importance of several factors in the system, including the ratio of copper(II) catalyst to formic acid, the presence of nitrite ion, reactions of diphenylmercury, and the purge gas employed in the system.

The first step towards immobilization of the soluble radioactive species in borosilicate glass is the addition of sodium tetraphenylborate (TPB) and sodium titanate to the radioactive aqueous solution. Initial studies of the TPB hydrolysis process have found that some component of the radiolysis mixture inactivates the Cu catalyst. The interaction of organic materials with the catalyst, and the subsequent interference with the hydrolysis process, would have presented problems with the use of the vitrification process. Prevention of the catalyst deactivation is obtained by washing the irradiated TPB precipitate in the Late Wash Facility prior to hydrolysis to remove the soluble radiolysis products. Identification of the organic radiolysis products, their distribution in the Late Wash Facility, and their interactions with the Cu catalyst has become an important analytical issue. To further investigate the reaction products of the TPB precipitation process, a simulated feedstock was created from compounds known to be present in the starting materials. This simulated feedstock was precipitated with sodium TPB and then exposed to Co-60 gamma radiation to simulate two years of additional storage time prior to the hydrolysis process. The irradiated product was divided into two parts, the filtered supernatant liquid and the precipitate slurry, which contains the TPB and the solid sodium titanate. Using gas chromatography/mass spectrometry, liquid secondary ion mass spectrometry, inductively coupled plasma/mass spectrometry, ion chromatography, and high performance liquid chromatography, over 50 organic and inorganic species have been identified in the aqueous portion of a simulated feedstock for TPB hydrolysis. The major organic species present are benzene, phenol, benzamide and a variety of substituted phenylphenols. The major inorganic species present are sodium, nitrite, and oxalate ions.

The addition of the Nitric Acid Flowsheet to the DWPF feed preparation process introduces nitric acid into a vessel which will subsequently receive a formic acid solution. The combination of these two acids suggests that a denitration reaction might occur. This memorandum reviews the conditions under which a denitration reaction is possible and compares these conditions to DWPF operating conditions.