Low-activity Waste Research Articles

Iodine valence and local environments in borosilicate waste glasses using X-ray absorption spectroscopy

The radioisotope 129I, a fission product in spent nuclear fuel, has a long half-life, and can be highly mobile in the environment. Iodine K-edge X-ray absorption spectra were collected to characterize the iodine valence and coordination environment in simulated Hanford low activity waste glasses. Both iodine XANES and EXAFS data for eleven borosilicate glasses indicate iodide-like environments within the glass structure, where I− has Na or Li nearest-neighbors, and where the nearest-neighbor cation-type correlates to the most common network-modifying cation in the glass. This is further supported by the systematic increase of iodine incorporation with the combined Na2O+Li2O content in the glass. EXAFS analyses determined I–Na distances near 3.04Å with coordination numbers near 4.0 and I–Li distances near 2.80Å with coordination numbers near 3.0. I–Na environments determined for the glasses are similar to the tetrahedral INa4 coordination found in NaI-sodalite. These weakly bound iodine-alkali configurations may be the only pathways for iodine to be retained in the glass. These environments may be precursors to NaI-sodalite crystallization in Na-rich glass. Iodine also shows distinct differences from chlorine in terms of the preferred sites in the glass structure.

Toward Understanding the Effect of Low‐Activity Waste Glass Composition on Sulfur Solubility

The concentration of sulfur in Hanford low‐activity waste (LAW) glass melter feed will be maintained below the point where the salt accumulates on the melt surface. The allowable concentrations may range from near zero to over 2.05 wt% (of SO3 on a calcined oxide basis) depending on the composition of the melter feed and processing conditions. If the amount of sulfur exceeds the melt tolerance level, a molten salt will accumulate which may upset melter operations and potentially shorten the useful life of the melter. At the Hanford site, relatively conservative limits have traditionally been placed on sulfur loading in melter feed, which in turn significantly increases the amount of LAW glass that will be produced. Crucible‐scale sulfur solubility data and scaled melter sulfur tolerance data have been collected on simulated Hanford waste glasses over the last 15 years. These data were compiled and analyzed. An empirical model was developed to predict the solubility of SO3 in glass based on 253 simulated Hanford LAW glass compositions. This model represents the data well, accounting for over 85% of the variation in data, and was well validated. The model was also found to accurately predict the maximum amount of sulfur in melter feed that did not form a salt layer in 13 scaled melter tests of simulated LAW glasses. The model can be used to help estimate glass volumes and make informed decisions on process options (e.g., scale of supplemental LAW treatment facility, and pretreatment facility performance requirements). The model also gives quantitative estimates of component concentration effects on sulfur solubility. The components that increase sulfur solubility most are Li2O > V2O5 > CaO ≈ P2O5 > Na2O ≈ B2O3 > K2O. The components that decrease sulfur solubility most are Cl > Cr2O3 > Al2O3 > ZrO2 ≈ SnO2 > Others (i.e., the sum of minor components) ≈SiO2. The order of component effects is similar to previous literature data, in most cases.

Iodine solubility in a low-activity waste borosilicate glass at 1000 °C

The purpose of this study was to determine the solubility of iodine in a low-activity waste borosilicate glass when heated inside an evacuated and sealed fused quartz ampoule. The iodine was added to glass frit as KI in quantities of 99.4–24,005ppm iodine (by mass). Each mixture was added to an ampoule, heated at 1000°C for 2h, and then air quenched. In samples with ⩾11,999ppm iodine, low viscosity salt phases were observed on the surface of the melts that solidified into a white coating upon cooling. These salts were identified by X-ray diffraction as mixtures of KI, NaI, and Na2SO4. Iodine concentrations in glass specimens were analyzed with inductively-coupled plasma mass spectrometry, and the overall iodine solubility was determined to be 10,000ppm. Several crystalline inclusions of iodine sodalite, Na8(AlSiO4)6I2, were observed in the 24,005ppm specimen.

An intrinsically safe facility for forefront research and training on nuclear technologies — Burnup and transmutation

The currently dominant open fuel cycles have resulted in the gradual accumulation of (relatively) large quantities of highly radioactive or fertile materials in the form of depleted uranium, plutonium, minor actinides (MA) and long-lived fission products (LLFP). For low-activity wastes a heavily shielded surface repository is required. Spent fuel can be instead directly buried in deep geological repositories or reprocessed in order to separate U and Pu and eventually also MA and LLFP from other materials. These elements can be further burnt by modern reactors but not yet in sufficient quantities to slow down the steady accumulation of these materials in storage. Using ADS, the residual long-lifetime isotopes can be transmuted by nuclear reactions into shorter-lifetime isotopes again storable in surface repositories. However, in order to perform transmutations at a practical level, high-power reactors (and consequently high-power accelerators) are required; particularly, a significant transmutation can be reached not only by increasing the beam current to something of the order of a few tens of mA, but also by increasing the beam energy above 500MeV in order to reach the spallation regime. Such high-power infrastructures require intermediate test facilities with lower power and higher safety level for the investigation of their dynamics and transmutation capabilities: the ADS proposed in this study could accomplish many of these constraints.

Redox-dependent solubility of technetium in low activity waste glass

The solubility of technetium was measured in a Hanford low activity waste (LAW) glass simulant, to investigate the extent that technetium solubility controls the incorporation of technetium into LAW glass. A series of LAW glass samples, spiked with 500–6000ppm of Tc as potassium pertechnetate, were melted at 1000°C in sealed fused quartz ampoules. Technetium solubility was determined in the quenched bulk glass to be 2000–2800ppm, with slightly reducing conditions due to choice of milling media resulting in reductant contamination and higher solubility. The chemical form of technetium obtained by X-ray absorption near edge spectroscopy is mainly isolated, octahedrally-coordinated Tc(IV), with a minority of Tc(VII) in some glasses and TcO2 in two glasses. The concentration and speciation of technetium depends on glass redox and amount of technetium added. Salts formed at the top of higher technetium loaded glasses during the melt. The results of this study show that technetium solubility should not be a factor in technetium retention during melting of Hanford LAW glass.

Experience gained at the Novovoronezh nuclear power plant with development and introduction of a unified automated system for accounting and control of radioactive substances and radioactive wastes

A unified automated system for accounting and control of radioactive substances and radioactive wastes was developed at the Novovoronezh nuclear power plant. The system consists of several interconnected subsystems, which can be developed and put in operation in a stage-wise manner. As of the first quarter of 2013, the subsystems for accounting liquid and solid radioactive wastes, very low-active wastes, and tracers have been developed. The experience gained from the development and putting in use of these systems showed that this is an efficient, timely, and much needed software product.

Performance of the Fluidized Bed Steam Reforming product under hydraulically unsaturated conditions

Several candidates for supplemental low-activity waste (LAW) immobilization at the Hanford site in Washington State, USA are being considered. One waste sequestering technology considered is Fluidized Bed Steam Reforming (FBSR). The granular product resulting from the FBSR process is composed primarily of an insoluble sodium aluminosilicate matrix with the dominant phases being feldspathoid minerals with a 1:1:1 molar ratio of Na, Al and Si. To demonstrate the durability of the product, which can be disposed of at the unsaturated Integrated Disposal Facility (IDF) at Hanford, a series of tests has been performed using the Pressurized Unsaturated Flow (PUF) system, which allows for the accelerated weathering of the solid materials. The system maintains hydraulically unsaturated conditions, thus mimicking the open-flow and transport properties that will be present at the IDF. Two materials were tested using the system: 1) the FBSR granular product and 2) the FBSR granular product encapsulated in a geopolymer to form a monolith. Results of the experiments show a trend of relatively constant effluent concentration of Na, Si, Al, and Cs as a function of time from both materials. The elements I and Re show a steady release throughout the yearlong test from the granular material but their concentrations seem to be increasing at one year from the monolith material. This result suggests that these two elements may be present in the sodalite cage structure rather than in the predominant nepheline phase because their release occurs at a different rate compared to nepheline phase. Also, these elements to not seem to reprecipitate when released from the starting material. Calculated one-year release rates for Si are on the order of 10−6 g/(m2 d) for the granular material and 10−5 g/(m2 d) for the monolith material while Re release is seen to be two orders of magnitude higher than Si release rates. SEM imaging and XRD analysis show how the alteration of the two materials is dependent on their depth in the column. This phenomenom is a result of depth-dependent solution concentrations giving rise chemical environments that may be supersaturated with respect to a number of mineral phases.

The performance of Inconel 693 electrodes for processing an iron phosphate glass melt containing 26 wt.% of a simulated low activity waste

Iron phosphate glass is a candidate fixation medium for storing radioactive waste. The Department of Energy supported a program to assess the viability of using Fe-phosphate glass for vitrifying low activity waste in a Joule Heated Melter (JHM). In this study, Inconel 693 electrodes were tested in a research-scale joule-heated melter (RSM) at Pacific Northwest National Laboratory. After a 10-day test at 1030°C that yielded 124kg of glass, the electrodes exhibited a dimensional loss rate of ∼1.6mm/year, which is comparable to that of Inconel 690 electrodes used in a JHM for processing borosilicate melts. Microstructural changes occurred within the outermost 700μm of the electrodes and are consistent with an earlier study of Inconel coupons in Fe-phosphate melts. The results indicate that Inconel 693 should have an acceptable corrosion resistance as the electrode for JHM processing of iron phosphate melts.

Modeling the long-term durability of concrete barriers in the context of low-activity waste storage

The paper investigates the long-term durability of concrete barriers in contact with a cementitious wasteform designed to immobilize low-activity nuclear waste. The high-pH pore solution of the wasteform contains high concentration level of sulfate, nitrate, nitrite and alkalis. The multilayer concrete/wasteform system was modeled using a multiionic reactive transport model accounting for coupling between species, dissolution/ precipitation reactions, and feedback effect. One of the primary objectives was to investigate the risk associated with the presence of sulfate in the wasteform on the durability of concrete. Simulation results showed that formation of expansive phases, such as gypsum and ettringite, into the concrete barrier was not extensive. Based on those results, it was not possible to conclude that concrete would be severely damaged, even after 5,000 years. Lab work was performed to provide data to validate the modeling results. Paste samples were immersed in sulfate contact solutions and analyzed to measure the impact of the aggressive environment on the material. The results obtained so far tend to confirm the numerical simulations.

Separation and Recovery of Cs from High Active Waste Simulant using Resorcinol Formaldehyde Polycondensate Resin: Batch and Column Studies

Resorcinol formaldehyde polycondensate resin has been extensively used for the separation of 137Cs from low and medium active alkaline waste solutions. The present study examines the suitability of the resin for separation of Cs from high active solutions. Batch uptake of Cs by RFPR has been evaluated from test solutions containing varying Na (0.6–3.0 M) and Cs (1.00 × 10−4 to 0.10 M) concentrations. Batch results have been used to establish Cs exchange isotherms of the resin and were found to follow Freundlich as well as Dubinin-Radushkevich (D-R) isotherm equations. The loading behavior of Cs on RFPR column has been evaluated by conducting five column runs using different Cs bearing solutions. Based on these results, it is concluded that the D-R equation can be used to predict saturated loading of Cs in the column. The saturation loading of Cs on 1 L bed was estimated to be as high as 13 g and 34 g from feed solution containing Cs concentration 2.57 × 10−4 and 4.30 × 10−3 mol/L respectively. Sequential elution of Na and Cs resulted in a high Cs bearing concentrate containing very low concentration of sodium. This study is relevant in the preparation of gamma source for radiation technology applications after immobilization of the concentrate in glass.

Gamma Shooting of the Territory of Municipal Landfill of the Dushanbe City

The rearrangement of Municipal landfill of Dushanbe city has begun; in this connection the gamma and neutron shooting of territory of landfill has been made. The landfill territory has been divided into 128 routes, and it has been conducted 5691 measurements of gamma background. The gamma background in territory is defined by background from environing soils 0.18±0.02 µZv/h and background from low-act ive wastes 0.11±0.02 µZv/h. During conducting gamma shooting the ownerless gamma source Cesiu m-137, with equivalent dose rate 40 µZv/h, has been detected. For localization and source extraction specialists of Co mmittee on emergency situations under government RT and Agency on Radiation and Nuclear Safety at Academy of Sciences RT have been invited. Fluency of neutrons on surface does not exceed background values and that testifies about absence of neutron sources.

Structure of Rhenium‐Containing Sodium Borosilicate Glass

A series of sodium borosilicate glasses were synthesized with KReO4 or Re2O7, to 10,000 ppm (1 mass%) target Re, to assess effects of large concentrations of rhenium on glass structure and fto estimate solubility of 99Tc, a radioactive component in low active waste nuclear glasses. Rhenium was used as a surrogate for 99Tc for laboratory testing, due to similarities in chemistry, ionic size, and redox. Magic angle spinning nuclear magnetic resonance, Fourier transform infrared spectroscopy, and Raman spectroscopy were performed to characterize the glasses. Si was coordinated in Q2 and Q3 units, Al was four‐coordinated, and B was mostly three‐coordinated. The rhenium additions did not have significant effects on the glass structure up to approximately 3000 ppm Re by mass, the maximum concentration that remained dissolved in glass. Rhenium likely exists in isolated anions in the interstices of the glass network, as evidenced by polarized Raman spectrum of the Re glass in the absence of sulfate. Analogous to in similar glasses, is a network modifier and above solubility forms alkali salt phases on the surface and in the bulk. Comparisons of phase separation and crystallization in containing borosilicate glasses can also be made to containing glasses.

Automated Radioanalytical System Incorporating Microwave-Assisted Sample Preparation, Chemical Separation, and Online Radiometric Detection for the Monitoring of Total 99Tc in Nuclear Waste Processing Streams

An automated fluidic instrument is described that rapidly determines the total (99)Tc content of aged nuclear waste samples, where the matrix is chemically and radiologically complex and the existing speciation of the (99)Tc is variable. The monitor links microwave-assisted sample preparation with an automated anion exchange column separation and detection using a flow-through solid scintillator detector. The sample preparation steps acidify the sample, decompose organics, and convert all Tc species to the pertechnetate anion. The column-based anion exchange procedure separates the pertechnetate from the complex sample matrix, so that radiometric detection can provide accurate measurement of (99)Tc. We developed a preprogrammed spike addition procedure to automatically determine matrix-matched calibration. The overall measurement efficiency that is determined simultaneously provides a self-diagnostic parameter for the radiochemical separation and overall instrument function. Continuous, automated operation was demonstrated over the course of 54 h, which resulted in the analysis of 215 samples plus 54 hly spike-addition samples, with consistent overall measurement efficiency for the operation of the monitor. A sample can be processed and measured automatically in just 12.5 min with a detection limit of 23.5 Bq/mL of (99)Tc in low activity waste (0.495 mL sample volume), with better than 10% RSD precision at concentrations above the quantification limit. This rapid automated analysis method was developed to support nuclear waste processing operations planned for the Hanford nuclear site.

Combined Experimental and Computational Approach to Predict the Glass-Water Reaction

The use of mineral and glass dissolution rates measured in laboratory experiments to predict the weathering of primary minerals and volcanic and nuclear waste glasses in field studies requires the construction of rate models that accurately describe the weathering process over geologic timescales. Additionally, the need to model the long-term behavior of nuclear waste glass for the purpose of estimating radionuclide release rates requires that rate models be validated with long-term experiments. Several long-term test methods have been developed to accelerate the glass-water reaction [drip test, vapor hydration test, product consistency test B, and pressurized unsaturated flow (PUF)], thereby reducing the duration required to evaluate long-term performance. Currently, the PUF test is the only method that mimics the unsaturated hydraulic properties expected in a subsurface disposal facility and simultaneously monitors the glass-water reaction. PUF tests are being conducted to accelerate the weathering of glass and validate the model parameters being used to predict long-term glass behavior. A one-dimensional reactive chemical transport simulation of glass dissolution and secondary-phase formation during a 1.5-year-long PUF experiment was conducted with the Subsurface Transport Over Reactive Multiphases (STORM) code. Results show that parameterization of the computer model by combining direct benchscale laboratory measurements and thermodynamic data provides an integrated approach to predicting glass behavior over the length of the experiment. Over the 1.5-year-long test duration, the rate decreased from 0.2 to 0.01 g/(m2·day) based on B release for low-activity waste glass LAWA44. The observed decrease is approximately two orders of magnitude higher than the decrease observed under static conditions with the SON68 glass (estimated to be a decrease by four orders of magnitude) and suggests that the gel-layer properties are less protective under these dynamic conditions.

Geochemical composition of natural waters near a storage site of low-activity radioactive wastes

Sampling data on subsoil and surface water near a storage site of low-activity liquid radioactive wastes, receiving neutralized uranium-containing pulps are given. Thermodynamic calculations are used to describe the behavior of major polluting and petrogenic components in the drainage system, allowing the occurrence forms of elements in solution and/or solid phase to be determined. The results suggest a positive effect of natural geochemical barriers hampering the process of uranium migration into natural waters. Over-all, the technogenic-natural system under consideration was found to effectively cope with the removal of uranium-containing compounds from industrial water.

Effect of particle size distribution on slurry rheology: Nuclear waste simulant slurries

Controlling the rheological properties of slurries has been of great interest in various industries such as cosmetics, ceramic processing, and nuclear waste treatment. Many physicochemical parameters, such as particle size, pH, ionic strength, and mass/volume fraction of particles, can influence the rheological properties of slurry. Among these parameters, the particle size distribution of slurry is especially important for nuclear waste treatment because most nuclear waste slurries show a broad particle size distribution. We studied the rheological properties of several different low activity waste nuclear simulant slurries having different particle size distributions under high salt and high pH conditions. Using rheological and particle size analysis, it was found that the percentage of colloid-sized particles in slurry appears to be a key factor for rheological characteristics and the efficiency of rheological modifiers. This behavior was shown to be coupled with an existing electrostatic interaction between particles in a low salt concentration. Our study suggests that the particle size distribution is a critical factor in understanding and controlling the rheological properties in nuclear waste treatment plants, such as the U.S. Department of Energy's Hanford and Savannah River sites, because the particle size distributions vary significantly for different types of nuclear waste slurries.

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.

Determination of trace amounts of plutonium in low-active liquid wastes from spent nuclear-fuel reprocessing plants by flow injection-based solid-phase extraction/electrochemical detection system

A flow injection-based electrochemical detection system coupled to a solid-phase extraction column was developed for the determination of trace amounts of plutonium in low-active liquid wastes from spent nuclear-fuel reprocessing plants. The oxidation state of plutonium in a sample solution was adjusted to Pu(VI) by the addition of silver(II) oxide. A sample solution was made up in 3 mol L−1 HNO3 and loaded onto a column packed with UTEVA® with 3 mol L−1 HNO3 as the carrier. Plutonium(VI) was adsorbed onto the resin, and interfering elements were removed by rinsing the column with 3 mol L−1 HNO3. Subsequently, the adsorbed Pu(VI) was eluted with 0.01 mol L−1 HNO3, and then introduced directly into the flow-through electrolysis cell with boron-doped diamond electrode. The eluted Pu(VI) was detected by an electrochemical amperometric method at a working potential of 0.1 V (vs. Ag/AgCl). The current produced on reduction of Pu(VI) was continuously monitored and recorded. The plutonium concentration was calculated from the relationship between the peak area and concentration of plutonium. The relative standard deviation of ten analyses was 1.1% for a plutonium solution of 25 μg L−1 containing 50 ng of Pu. The detection limit calculated from three-times the standard deviation was 0.82 μg L−1 (1.6 ng of Pu).

Comparison of the Behavior of Technetium and Rhenium in Low Activity Waste Glass Formulations Subjected to the Vapor Hydration Test

Extended abstract of a paper presented at Microscopy and Microanalysis 2010 in Portland, Oregon, USA, August 1 – August 5, 2010.

Cementitious Wasteforms for Immobilization of Low-Activity Radioactive Wastes

AbstractSolidification of low-activity wastes with cementitious materials is a widely accepted technique that contains and isolates waste from the hydrologic environment. The radionuclides I-129, Se-75, Tc-99, and U-238 are identified as long-term dose contributors. The anionic nature of these radionuclides in aqueous solutions allows them to readily leach into the subsurface environment. Any failure of concrete encasement may result in water intrusion and consequent mobilization of radionuclides from the waste packages via mass flow and/or diffusion into the surrounding subsurface environment. Assessing the long-term performance of waste grouts for encasement of radionuclides requires understanding the: 1) speciation and interaction of the radionuclides within the concrete wasteform, 2) diffusion of radionuclide species when contacted with vadose zone porewater or groundwater under environmentally relevant conditions, and 3) long-term durability and weathering of concrete waste forms. An improved understanding of the interactions of long-lived radionuclides in cementitious matrices will improve predictions of the long-term fate of these sequestered contaminants. An integrated laboratory investigation has been conducted including a: 1) multifaceted spectroscopic investigation to interrogate the speciation and interaction of radionuclides within concrete wasteforms, 2) solubility tests to quantify the stability of solid phases identified as radionuclide-controlling phases, 3) quantify the diffusion of radionuclides from concrete wasteforms into surrounding subsurface sediment under realistic moisture contents (4%, 7%, and 15% by weight moisture content), 4) quantify the long-term durability of concrete waste forms as a function environmental parameters relevant to depository conditions, and 5) identify the formation of secondary phases or processes (microcracking) that influence radionuclide retention. Data obtained from this investigation provides valuable information for understanding the speciation, behavior, and fate of radionuclides immobilized within concrete wasteforms under vadose zone conditions and underscores the necessity for robust, multi-disciplinary performance assessments for concrete waste forms.