Radioactive Tanks Research Articles

Automated SEM analysis of particles in Hanford tank waste

Multiple bench-scale filtration campaigns of Hanford tank waste supernatant on a backpulseable dead-end filtration skid have provided greater insight into the solids that cause fouling and reduce filter performance. The solids collected during each campaign were concentrated from the backpulse solutions and examined using automated particle analysis (APA) methods with scanning electron microscopy and X-ray energy dispersive spectroscopy to categorize particle types and their morphological characteristics. We show that with APA, thousands of particles can be analyzed to provide accurate insight into the phases that may be impacting filter performance.

Long-Term Performance of Ag/AgCl Reference Electrodes for Corrosion Potential Monitoring in Radioactive Tank Waste at the Hanford Site

This work studied the exposure effects of radioactive tank waste on the long-term performance of single junction Ag/AgCl reference electrodes for corrosion potential monitoring at the Hanford Site. Electrodes from three manufacturers with very different designs were studied using open-circuit potential and electrochemical impedance measurements in radioactive tank waste. Post-test analyses were conducted on some failed electrodes using destructive and nondestructive techniques. The intrusion of the aggressive and radioactive chemicals in tank waste through the porous frit materials was the primary mechanism that led to the clogging of frit, physical and chemical degradation of AgCl bonded to Ag wire, and alteration of the internal electrolyte. Radiolytic chemical species such as H2O2 and HNO3 may have also induced the degradation of the Ag wire. The extent of electrode degradation and failure probability highly depended on the electrode design and environmental conditions. Chemicals in tank waste had stronger effects than radiation on the long-term performance of the Ag/AgCl reference electrodes.

Temperature-responsive alkaline aqueous biphasic system for radioactive wastewater treatment

The treatment of anionic 99TcO4− in the waste tank with high alkalinity is still very challenging. In this work, a new temperature-responsive alkaline aqueous biphasic system (ABS) based on (tri-n‑butyl)-n-tetradecyl phosphonium chloride (P44414Cl) was developed to remove radioactive 99TcO4−. The phase transition mechanism was studied by cloud point titration, small-angel X-ray scattering, dynamic light scattering, and molecular dynamic simulations. As the NaOH concentration or temperature increased, the P44414+ micelle could grow and aggregate. This micelle showed a particularly high affinity toward ReO4−/99TcO4− compared to other competing anions and could directly extract more than 98.6% of 99TcO4− from simulated radioactive tank waste supernatant. Furthermore, the loaded 99TcO4− could be easily stripped by using concentrated nitric acid rather than metal salt-based reductants. This work clearly demonstrates that the alkaline ABS is a promising separation system for solving the technetium problem in the alkaline waste tank.

Effect of Cr(III) Adsorption on the Dissolution of Boehmite Nanoparticles in Caustic Solution.

The incorporation of relatively minor impurity metals onto metal (oxy)hydroxides can strongly impact solubility. In complex highly alkaline multicomponent radioactive tank wastes such as those at the Hanford Nuclear Reservation, tests indicate that the surface area-normalized dissolution rate of boehmite (γ-AlOOH) nanomaterials is at least an order of magnitude lower than that predicted for the pure phase. Here, we examine the dissolution kinetics of boehmite coated by adsorbed Cr(III), which adheres at saturation coverages as sparse chemisorbed monolayer clusters. Using 40 nm boehmite nanoplates as a model system, temperature-dependent dissolution rates of pure versus Cr(III)-adsorbed boehmite showed that the initial rate for the latter is consistently several times lower, with an apparent activation energy 16 kJ·mol-1 higher. Although the surface coverage is only around 50%, solution analysis coupled to multimethod solids characterization reveal a phyicochemical armoring effect by adsorbed Cr(III) that substantially reduces the number of dissolution-active sites on particle surfaces. Such findings could help improve kinetics models of boehmite and/or metal ion adsorbed boehmite nanomaterials, ultimately providing a stronger foundation for the development of more robust complex radioactive liquid waste processing strategies.

Numerical comparison of bubbling in a waste glass melter

Radioactive tank waste is scheduled for vitrification at the Waste Treatment and Immobilization Plant (WTP) being constructed at the Hanford Site. Testing of the pilot-scale DuraMelter 1200 at the Vitreous State Laboratory at the Catholic University of America has demonstrated that bubbling increases the melt rate of the batch material, and as a result, melter throughput. Computational fluid dynamics (CFD) models of this pilot-scale waste glass melter are being developed to improve our understanding of the processes that occur within the melter to aid in process optimization and troubleshooting of the WTP melters. Unfortunately, model validation is complicated by the difficulty of obtaining suitable experimental data for operational melters due to the inaccessibility for direct observation and measurements of the high-temperature, opaque fluid through the water-jacketed, refractory-lined steel vessel. This study focuses on assessing the fidelity of the CFD models to accurately predict the bubbling behavior. Because of the paucity of experimental data at the resolution required for CFD validation, a code comparison was used to evaluate two common approaches for simulating flows of two immiscible Newtonian fluids on numerical grids and resolving multiphase interfaces. Here, the volume of fluid and level set methods are used to resolve the dynamically evolving interfaces between the molten glass and the air bubbles. To aid in the validation of the results of these codes, a comparison of the bubble behavior, growth, and frequency of bubble generation are presented and a grid convergence study is performed for the two approaches. The predictions from the two codes are within 6% for the average bubble radius of curvature within the bubble channels, within 2% for average terminal rise velocity of the bubbles, and within 4% for the area mean of the local maxima at the free surface. These parameters are of interest since they affect the convection within the melter and at the interface between the glass and batch layer. Ultimately, the results of this work can assist in confirming the predictive ability of waste glass melter models and provide a better understanding of the flow patterns within the WTP melters.

Evidence of technetium and iodine release from a sodalite-bearing ceramic waste form

Sodalites have been proposed as a possible host of certain radioactive species, specifically 99Tc and 129I, which may be encapsulated into the cage structure of the mineral. To demonstrate the ability of this framework silicate mineral to encapsulate and immobilize 99Tc and 129I, single-pass flow-through (SPFT) tests were conducted on a sodalite-bearing multi-phase ceramic waste form produced through a steam reforming process. Two samples made using a steam reformer samples were produced using non-radioactive I and Re (as a surrogate for Tc), while a third sample was produced using actual radioactive tank waste containing Tc and added Re. One of the non-radioactive samples was produced with an engineering-scale steam reformer while the other non-radioactive sample and the radioactive sample were produced using a bench-scale steam reformer. For all three steam reformer products, the similar steady-state dilute-solution release rates for Re, I, and Tc at pH (25 °C) = 9 and 40 °C were measured. However, it was found that the Re, I, and Tc releases were equal or up to 4.5x higher compared to the release rates of the network-forming elements, Na, Al, and Si. The similar releases of Re and Tc in the SPFT test, and the similar time-dependent shapes of the release curves for samples containing I, suggest that Re, Tc, and I partition to the sodalite minerals during the steam reforming process.

Technetium Getters to Improve Cast Stone Performance

ABSTRACTA cementitious waste form known as Cast Stone is the baseline waste form for solidification of aqueous secondary wastes, including Hanford Tank Waste Treatment and Immobilization Plant (WTP) secondary liquid effluents to be treated and solidified at the Hanford Site Effluent Treatment Facility. Cast Stone is also being evaluated as a possible supplemental immobilization technology to provide the necessary low activity waste (LAW) treatment capacity to complete the Hanford tank waste cleanup mission in a timely and cost effective manner. Two radionuclides of particular concern in these waste streams are technetium-99 (99Tc) and iodine-129 (129I). These radioactive tank waste components are predicted to contribute the most risk to groundwater – the most probable pathway for future environmental impacts associated with the cleanup of the Hanford site. A recent environmental assessment of Cast Stone performance, which assumes a diffusion controlled release of contaminants from the waste form, calculates groundwater in excess of the allowable maximum permissible concentrations for both contaminants. There is, therefore, a need and an opportunity to improve the retention of both 99Tc and 129I in Cast Stone. One method to improve the performance of Cast Stone is through the addition of “getters” that selectively sequester Tc and I, therefore reducing their diffusion out of Cast Stone. In this paper, we present results of Tc and I removal from solution with various getters. Batch sorption experiments were conducted with deionized water (DIW) and a highly caustic LAW simulant with a 7.8 M average Na concentration. In general, the data show that the selected getters are effective in DIW but their performance is compromised when experiments are performed with the 7.8 M Na Ave LAW simulant. The diminished performance in the LAW simulant may be due to competition with Cr present in the 7.8 M Na Ave LAW simulant and to a pH effect that may create a negatively charged surface that can repel negatively charged species.

Application of Metallography for the Slovak Nuclear Power Plants

The austenitic cladding of liquid radioactive waste underground storage tanks was recently required to be under continuous corrosion surveillance with aim to monitor its corrosion stability until all radioactive waste will be processed and safely stored. Tanks are intended primarily to be used for long-term storage of radioactive waste water solutions and nascent sediments. Tanks are components of technology designed for post-processing of radioactive sediments by means of cement matrix fixation. This paper is dealing with an example of metallografical methods focused on corrosion processes monitoring using the special surveillance program. The brief information is presented for the preparation of specific long-run corrosion programme and experimental material samples evaluation. General view of the monitoring system is shown in Fig. 1.

Insights into Stabilization of the 99 Tc V O Core for Synthesis of 99 Tc V O Compounds

Synthesis of technetium-99 (99Tc; t1/2: 2.1 105 years, max: 253 keV) materials is of importance in studies of the nuclear fuel cycle where Tc is a major fission product (6percent thermal yield from 235U and 239Pu), in understanding radioactive tank waste composition, and in identifying 99mTc compounds for nuclear medicine imaging. One of the most useful synthetic starting materials, (NBu4)TcOCl4, is susceptible to disproportionation in water to form TcO4 and TcIV species, especially TcO2 2H2O. This unwanted reaction is especially problematic when working with ligands bearing hard donor atoms, such as oxygen, where the stability with the soft TcV=O3+ core may be low. Polyoxometalates (POMs) are such ligands. They possess defect sites with four hard oxygen atoms and show low (ca. 108) stability constants with transition metals. Tc complexes of POMs are molecular-level models for Tc metal oxide solid-state materials and can provide information on coordination and redox environments of metal oxides that stabilize low-valent Tc. In order to synthesize pure Tc POM complexes [TcVO(1-P2W17O61)]7 (TcVO-1) and [TcVO(2-P2W17O61)]7 (TcVO-2) from (NBu4)TcOCl4, we have identified strategies that minimize formation of TcIV species and optimize the formation of pure TcV species. The parameters that we consider are the amount of ethylene glycol,more » which is employed as a transfer ligand to prevent hydrolysis of (NBu4)TcOCl4, and the precipitating agent. The TcIV species that contaminates the non-optimized syntheses is likely a TcIV -oxido-bridged dimer [TcIV-(-O)2-TcIV]. We also employ a novel procedure where the 2 ligand is photoactivated and reduced (in the presence of a sacrificial electron donor) to subsequently reduce TcVIIO4 to an isolatable TcVO-2 product that is remarkably free of TcIV.« less

Changes in the pore network structure of Hanford sediment after reaction with caustic tank wastes

At the former nuclear weapon production site in Hanford, WA, caustic radioactive tank waste leaks into subsurface sediments and causes dissolution of quartz and aluminosilicate minerals, and precipitation of sodalite and cancrinite. This work examines changes in pore structure due to these reactions in a previously-conducted column experiment. The column was sectioned and 2D images of the pore space were generated using backscattered electron microscopy and energy dispersive X-ray spectroscopy. A pre-precipitation scenario was created by digitally removing mineral matter identified as secondary precipitates. Porosity, determined by segmenting the images to distinguish pore space from mineral matter, was up to 0.11 less after reaction. Erosion-dilation analysis was used to compute pore and throat size distributions. Images with precipitation had more small and fewer large pores. Precipitation decreased throat sizes and the abundance of large throats. These findings agree with previous findings based on 3D X-ray CMT imaging, observing decreased porosity, clogging of small throats, and little change in large throats. However, 2D imaging found an increase in small pores, mainly in intragranular regions or below the resolution of the 3D images. Also, an increase in large pores observed via 3D imaging was not observed in the 2D analysis. Changes in flow conducting throats that are the key permeability-controlling features were observed in both methods.

Nuclear Energy: A Perspective on Recent Results on the Removal of Strontium from Waste

Nuclear Energy: A Perspective on Recent Results on the Removal of Strontium from Waste

Review: Waste-Pretreatment Technologies for Remediation of Legacy Defense Nuclear Wastes

The U.S. Department of Energy (DOE) is responsible for retrieving, immobilizing, and disposing of radioactive waste generated during the production of nuclear weapons in the United States. The general strategy for treating the radioactive tank waste consists of pretreating the wastes by separating them into high-level and low-activity fractions. The high-level fraction will be immobilized in a glass form suitable for disposal in a geologic repository. The low-activity waste will be immobilized in a waste form suitable for on-site. This paper reviews recent developments in the application of pretreatment technologies to the processing of the DOE radioactive tank wastes.

The application of in situ formed mixed iron oxides in the removal of strontium and actinides from nuclear tank waste

AbstractThis article discusses the experimental program conducted at Argonne National Laboratory to study the performance of in situ formed mixed iron oxides (IS‐MIO) for the removal of strontium (Sr), plutonium (Pu), neptunium (Np), uranium (U), and americium (Am) from the Savannah River Site (SRS) radioactive tank waste. The boundaries for the experimental work were defined in collaboration with SRS. IS‐MIO was actually found to be a mixture of Fe(II) and Fe(III) oxides and hydroxides, including magnetite. Decontamination factor (DF) values were measured for both IS‐MIO and monosodium titanate (MST), the baseline sorbent used by SRS. DF values for IS‐MIO were found to be superior to MST for all isotopes studied. DF values for Pu, Np, and Sr, achieved within 30 min of IS‐MIO formation were orders of magnitude larger than the needed values. DF values for U and Am were less than the former three but still acceptable, and greater than MST. © 2010 American Institute of Chemical Engineers AIChE J, 2010

Characterization of High Phosphate Radioactive Tank Waste and Simulant Development

A sample of high-level radioactive tank waste was characterized to provide a basis for developing a waste simulant. The simulant is required for pilot-scale testing of pretreatment processes in a nonradiological facility. The waste material examined was derived from the bismuth phosphate process, which was the first industrial process implemented to separate plutonium from irradiated nuclear fuel. The bismuth phosphate process sludge is a complex mixture rich in bismuth, iron, sodium, phosphorus,silicon, and uranium.The form of phosphorus in this particular tank waste material is of specific importance because that is the primary component (other than water-soluble sodium salts) that must be removed from the high-level waste solids by pretreatment. This work shows unequivocally that the phosphorus in this waste material is not present as bismuth phosphate. Rather, the phosphorus appears to be incorporated mostly into an amorphous iron(III) phosphate phase. The bismuth in the sludge solids is best described as BiFeO3. The behavior of phosphorus during caustic leaching of the bismuth phosphate process sludge solids is also discussed.

Advective Removal of Intraparticle Uranium from Contaminated Vadose Zone Sediments, Hanford, U.S.

A column study on U(VI)-contaminated vadose zone sediments from the Hanford Site, WA, was performed to investigate U(VI) release kinetics with water advection and variable geochemical conditions. The sediments were collected from an area adjacent to and below tank BX-102 that was contaminated as a result of a radioactive tank waste overfill event. The primary reservoir for U(VI) in the sediments are micrometer-size precipitates composed of nanocrystallite aggregates of a Na-U-Silicate phase, most likely Na-boltwoodite, that nucleated and grew within microfractures of the plagioclase component of sand-sized granitic clasts. Two sediment samples, with different U(VI) concentrations and intraparticle mass transfer properties, were leached with advective flows of three different solutions. The influent solutions were all calcite-saturated and in equilibrium with atmospheric CO2. One solution was prepared from DI water, the second was a synthetic groundwater (SGW) with elevated Na that mimicked groundwater at the Hanford site, and the third was the same SGW but with both elevated Na and Si. The latter two solutions were employed, in part, to test the effect of saturation state on U(VI) release. For both sediments, and all three electrolytes, there was an initial rapid release of U(VI) to the advecting solution followed by slower near steady-state release. U(VI)aq concentrations increased during subsequent stop-flow events. The electrolytes with elevated Na and Si depressed U(VL)aq concentrations in effluent solutions. Effluent U(VI)aq concentrations for both sediments and all three electrolytes were simulated reasonably well by a three domain model (the advecting fluid, fractures, and matrix) that coupled U(VI) dissolution, intraparticle U(VI)aq diffusion, and interparticle advection, where diffusion and dissolution properties were parameterized in a previous batch study.

Flowsheet Modeling and Testing of Pseudohydroxide Extraction from Aqueous Sodium Hydroxide Solutions with 3,5‐di‐tert‐Butylphenol in Isopar® L Modified with Exxal® 8

A conceptual counter‐current process flowsheet was developed for sodium hydroxide recovery from alkaline solutions via pseudohydroxide extraction (PHE). PHE relies on a simple sodium ion/proton exchange mechanism at elevated pH using a weak organic acid extractant. The contact of the sodium‐loaded organic phase with water results in the reconstitution of the extractant in the organic phase and sodium hydroxide in the aqueous phase. In this work, the 3,5‐di‐tert‐butylphenol (35‐DTBP) cation exchanger was used in the Isopar® L diluent modified with isooctyl alcohol Exxal® 8. Equilibrium isotherms determined for PHE from pure sodium hydroxide solutions and simulated radioactive waste leachate were used to develop a semi‐empirical model that could be used for designing PHE process flowsheets. Using this model, a conceptual PHE flowsheet was developed for recovering NaOH from solutions generated by caustic leaching of radioactive tank sludges. The flowsheet consists of the extraction, scrub, and strip processes, each employing four equilibrium stages. The modeling of this flowsheet indicates 97% recovery of the sodium hydroxide from the waste leachate feed solution. An experimental demonstration, performed with a simulated radioactive waste leachate using batch contacts in a co‐current analog of the counter‐current flowsheet, confirmed the potential for practical application of PHE technology.

Cancrinite and sodalite formation in the presence of cesium, potassium, magnesium, calcium and strontium in Hanford tank waste simulants

High-level radioactive tank waste solutions that have leaked into the subsurface at the US Department of Energy Hanford Site, Washington, are chemically complex. Here, the effect of five cations, Cs +, K +, Sr 2+, Ca 2+ and Mg 2+, on mineral formation and transformation pathways under conditions mimicking Hanford tank leaks is investigated. Sodium silicate was used to represent the dissolved silicate from sediments. The silicate was added into a series of simulants that contained 0.5 M aluminate, 1 M or 16 M NaOH, and the NO 3 - salts of the cations. The precipitates were monitored by X-ray diffraction, scanning electron microscopy, and X-ray energy dispersive spectroscopy. In the 1 M NaOH simulants, low concentration of Cs + (<100 mM) did not affect the formation of lepispheric cancrinite and sodalite, whereas only highly crystalline cancrinite formed when Cs + concentration was ⩾250 mM. An unidentified feldspathoid or zeolite intermediate phase was observed in the presence of high concentrations of Cs + (500 mM). The presence of K + did not alter, but slowed, the formation of cancrinite and sodalite. The presence of divalent cations led to the formation of metastable or stable silicates, aluminates, hydroxides, or aluminosilicates. The formation of these intermediate phases slowed the formation of cancrinite and sodalite by consuming OH −, silicate, or aluminate. Compared with the concentrations used in this study, the concentrations of radioactive Cs + and Sr 2+ in the tank solutions are much lower and divalent cations (Ca 2+ and Mg 2+) released from sediments likely precipitate out as hydroxides, silicates or aluminates; therefore, the authors do not expect that the presence of these monovalent and divalent cations significantly affect the formation of cancrinite and sodalite in the sediments underneath the leaking waste tanks.

Clay mineral weathering and contaminant dynamics in a caustic aqueous system: II. Mineral transformation and microscale partitioning

Microscopic and spectroscopic studies were conducted to assess mineral transformation processes in aqueous suspensions of illite (Il), vermiculite (Vm) and montmorillonite (Mt) that were subjected to weathering in a simulated high-level radioactive tank waste leachate (0.05 m Al T, 2 m Na +, 1 m NO 3 −, pH ∼14, Cs + and Sr 2+ present as co-contaminants). Time series (0 to 369 d) experiments were conducted at 298 K, with initial [Cs] 0 and [Sr] 0 concentrations from 10 −5 to 10 − mol kg −. Incongruent clay dissolution resulted in an accumulation of secondary aluminosilicate precipitates identified as nitrate-sodalite, nitrate-cancrinite and zeolite X, by molecular spectroscopy and electron microscopy (XRD, IR, NMR, SEM-EDS and TEM-EDS). Contaminant fate was dependent on competing uptake to parent clays and weathering products. TEM-EDS results indicated that high Il affinity for Cs was due to adsorption at frayed edge sites. The Il system also comprised Sr-rich aluminous precipitates after 369 d reaction time. In Mt systems, Cs and Sr were co-precipitated into increasingly recalcitrant spheroidal precipitates over the course of the experiment, whereas contaminant association with montmorillonite platelets was less prevalent. In contrast, Cs and Sr were found in association with weathered Vm particles despite the formation of spheroidal aluminosilicate precipitates that were comparable to those formed from Mt dissolution.

Vitrification and testing of a Hanford high-level waste sample. Part 1: Glass fabrication, and chemical and radiochemical analysis

The Hanford radioactive tank waste will be separated into low-activity waste and high-level waste that will both be vitrified into borosilicate glasses. To demonstrate the feasibility of vitrification and the durability of the high-level waste glass, a high-level waste sample from Tank AZ-101 was processed to glass, analyzed with respect to chemical composition, radionuclide content, waste loading, and the presence of crystalline phases and then tested for leachability. The glass was analyzed with inductively coupled plasma-atomic emission spectroscopy, inductively coupled plasma-mass spectrometry, γ-energy spectrometry, α-spectrometry, and liquid scintillation counting. The WISE Uranium Project calculator was used to calculate the main sources of radioactivity to the year 3115. The observed crystallinity and the results of leachability testing of the glass will be reported in Part 2 of this paper.

Removal of Cesium from Acidic Radioactive Tank Waste by Using Ionsiv IE‐911

IONSIV IE‐911, or the engineered form of crystalline silicotitanate (CST), manufactured by UOP Molecular Sieves, has been evaluated for the removal of cesium from Idaho National Engineering and Environmental Laboratory (INEEL) acidic radioactive tank waste. A series of batch contacts and column tests were performed by using three separate batches of CST. Batch contacts were performed to evaluate the concentration effects of nitric acid, sodium, and potassium ions on cesium sorption. Additional batch tests were performed to determine if americium, mercury, and plutonium would sorb onto IONSIV IE‐911. An equilibrium isotherm was generated by using a concentrated tank waste simulant. Column tests using a 1.5 cm3 column and flow rates of 3, 5, 10, 20, and 30 bed volumes (BV)/hr were performed to elucidate dynamic cesium sorption capacities and sorption kinetics. Additional experiments investigated the effect of CST batch and pretreatment on cesium sorption. The thermal stability of IONSIV IE‐911 was evaluated by performing thermal gravimetric analysis/differential thermal analysis. Overall, IONSIV IE‐911 was shown to be effective for cesium sorption from complex, highly acidic solutions; however, sorbent stability in these solutions may have a deleterious effect on cesium sorption.