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All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Summary of MCU Flush Sample Results

Starting in May 2021, samples from the Modular Caustic-Side Solvent Extraction Unit (MCU) were sent to the Savannah River National Laboratory (SRNL) for analysis. Samples resulting from flushes of the Strip Effluent Hold Tank (SEHT) and the Decontaminated Salt Solution Hold Tank (DSSHT) were sent on multiple dates. Over the range of samples, all had less than detectable concentrations of Isopar-L™, and there was an overall decline in 137 Cs activity.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Rheology of simulated radioactive waste slurry and cold cap during vitrification

During the vitrification of radioactive waste in a Joule-heated melter, aqueous melter feed slurry forms a cold cap, a reacting and melting material, which floats on the surface of the molten glass. The rheological behavior of the feed affects cold cap formation and shape, and is vital for modeling the feed-to-melt conversion process. Here, we used slurry feed simulant and fast-dried slurry solids representing the cold cap to investigate the rheological behavior of the feed as it transforms into glass. Both low-temperature and high-temperature rheometry were performed and a new scheme was applied to estimate the feed viscosity. This study shows that the conversion advances in four sequential stages that form distinct regions in the cold cap: (i) a fast-spreading boiling slurry from which water evaporates, (ii) a porous solid region (viscosity > 10 8 Pa s) containing reacting solids and molten salts, (iii) a plastic region in which glass-forming melt connects the refractory solids (~10 8 to ~10 6 Pa s), and (iv) a viscous foam layer in which the viscosity drops from ~10 5 to ~10 1 Pa s. The implications for the mathematical modeling of the cold cap are discussed.

glass melting↗
DOE OSTI2025

Application of Raman Spectroscopy to Determine Uranium Content in ADUN Solution

The work presented in this report is part of the ongoing efforts to address the nuclear material control and accounting needs for advanced reactor fuel fabrication facilities. This work was supported by the Materials Protection, Accounting, and Control Technologies (MPACT) program under the US Department of Energy Office of Nuclear Energy‘s Nuclear Fuel Cycle and Supply Chain program. The activities and engagements under the MPACT program are designed to support a robust US civilian nuclear energy enterprise. In the work described in this report, we supported MPACT objectives by developing measurement techniques that could be used for material accounting and process monitoring and by working with industry partners to identify existing gaps and areas for improvement. Oak Ridge National Laboratory has been working with commercial tristructural isotropic (TRISO) fuel fabricators such as Standard Nuclear to develop technology for rapid and cost-effective uranium content assessment. This work has focused on demonstrating advanced measurement techniques (e.g., Raman spectroscopy) that can be used for rapid, reliable, and cost-effective routine measurements of uranium content in feed solutions and liquid waste streams as well as for monitoring in-line process measurements and product streams. Specifically, this report explores techniques for accurately determining uranium content in acid-deficient uranyl nitrate (ADUN) solutions and detecting low uranium concentrations in ammonia solutions. Developing such measurement techniques will benefit TRISO fuel fabrication facilities, facilities involved in other parts of the fuel cycle that require online monitoring of aqueous solutions, and potentially molten salt fuel reactors. This work supports developing Raman spectroscopy procedures to determine uranium concentrations in ADUN solutions, which are used as feedstock in the sol–gel process for creating TRISO fuel. Some additional benefits of using Raman spectroscopy for uranium quantification in fabrication facilities include enabling online monitoring of the chemical process, which would provide near real-time feedback; eliminating the need for sample transfers, preparation, or dilution; providing nondestructive measurements; and user friendliness. In this fiscal year, FY25, we determined the identity of the unknown Raman band at approximately 853 cm−1 that was discovered in ADUN Raman spectra in FY24, created calibration curves and determined uranium concentrations of two ADUN solutions, and compared the Raman results to results obtained from inductively coupled plasma mass spectrometry and Davies–Gray titration. Furthermore, we have identified focus areas for experimentation in future fiscal years. A key result is that the accuracy of using Raman can provide accuracy comparable to destructive analysist techniques, With a well-developed calibration curve, using standards and a large number of samples (more than five samples), uncertainty on the order of 1%–3% is achievable. Given that the uncertainties achieved by Raman spectroscopy were on the order of the uncertainties achieved using ICP-MS, we conclude that with a well-developed procedure Raman spectroscopy can be used to determine uranium concentrations in ADUN solutions for NMC&A applications. The benefits of such an approach are that the time and effort will be less than that of comparable destructive analysis techniques, with approximately the same level of technical expertise. This will be attractive to operators of fuel fabrication facilities as it will lower costs and improve efficiencies.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Heat Transfer Model of a Small-Scale Waste Glass Melter with Cold Cap Layer

At the Hanford site in the state of Washington, more than 56 million gallons of radioactive waste is stored in underground tanks. The cleanup plan for this waste is vitrification at the Waste Treatment Plant (WTP), currently under construction. At the WTP, the waste will be blended with glass-forming materials and heated to 1423K, then poured into stainless steel canisters to cool and solidify. A fundamental understanding of the glass batch melting process is needed to optimize the process to reduce cost and decrease the life cycle of the cleanup effort. The cold cap layer that floats on the surface of the glass melt is the primary reaction zone for the feed-to-glass conversion. The conversion reactions include water release, melting of salts, evolution of batch gases, dissolution of quartz and the formation of molten glass. Obtaining efficient heat transfer to this region is crucial to achieving high rates of glass conversion. Computational fluid dynamics (CFD) modeling is being used to understand the heat transfer dynamics of the system and provide insight to optimize the process. A CFD model was developed to simulate the DM1200, a pilot-scale melter that has been extensively tested by the Vitreous State Laboratory (VSL). Electrodes are built into the melter to provide Joule heating to the molten glass. To promote heat transfer from the molten glass into the reactive cold cap layer, bubbling of the molten glass is used to stimulate forced convection within the melt pool. A three-phase volume of fluid approach is utilized to model the system, wherein the molten glass and cold cap regions are modeled as separate liquid phases, and the bubbling gas and plenum regions are modeled as one lumped gas phase. The modeling of the entire system with a volume of fluid model allows for the prescription of physical properties on a per-phase basis. The molten glass phase and the gas phase physical properties are obtained from previous experimental work. Finding representative properties for the cold cap region is more difficult, as this region is not a true liquid, but rather a multilayer region consisting of a porous and a foamy layer. Physical properties affecting heat transfer, namely the thermal conductivity and heat capacity, have been fit to closely match data and observations from laboratory experiments. Data from xray tomography and quenching of laboratory-scale cold caps provide insight into the topology of bubble distribution within the cold cap at various temperatures. Heat transfer within the melter was validated by comparison with VSL data for the pilot-scale melter.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
NASA NTRS2013

NASA Tech Briefs, April 2013

Topics covered include: Fully Integrated, Miniature, High-Frequency Flow Probe Utilizing MEMS Leadless SOI Technology; Nanoscale Surface Plasmonics Sensor With Nanofluidic Control; Advanced Dispersed Fringe Sensing Algorithm for Coarse Phasing Segmented Mirror Telescopes; Neural Network Back-Propagation Algorithm for Sensing Hypergols; Bulk Moisture and Salinity Sensor; Change-Based Satellite Monitoring Using Broad Coverage and Targetable Sensing; Circularly Polarized Microwave Antenna Element with Very Low Off-Axis Cross-Polarization; Ultra-Low Heat-Leak, High-Temperature Superconducting Current Leads for Space Applications; Flash Cracking Reactor for Waste Plastic Processing; An Automated Safe-to-Mate (ASTM) Tester; Wireless Chalcogenide Nanoionic-Based Radio-Frequency Switch; Compute Element and Interface Box for the Hazard Detection System; DOT Transmit Module; Composite Aerogel Multifoil Protective Shielding; Li-Ion Electrolytes with Improved Safety and Tolerance to High-Voltage Systems; Polymer-Reinforced, Non-Brittle, Lightweight Cryogenic Insulation; Controlled, Site-Specific Functionalization of Carbon Nanotubes with Diazonium Salts; Regenerable Sorbent for CO2 Removal; Sprayable Aerogel Bead Compositions With High Shear Flow Resistance and High Thermal Insulation Value; Lexan Linear Shaped Charge Holder with Magnets and Backing Plate; Robotic Ankle for Omnidirectional Rock Anchors; Wind, Wave, and Tidal Energy Without Power Conditioning; An Active Heater Control Concept to Meet IXO Type Mirror Module Thermal-Structural Distortion Requirement; Waterless Clothes-Cleaning Machine; Integrated Electrical Wire Insulation Repair System; LVGEMS Time-of-Flight Mass Spectrometry on Satellites; Surface Inspection Tool for Optical Detection of Surface Defects; Per-Pixel, Dual-Counter Scheme for Optical Communications; Certification-Based Process Analysis; Surface Navigation Using Optimized Waypoints and Particle Swarm Optimization; Smart-Divert Powered Descent Guidance to Avoid the Backshell Landing Dispersion Ellipse; Estimating Foreign-Object-Debris Density from Photogrammetry Data; Adaptive Sampling of Spatiotemporal Phenomena with Optimization Criteria; Building a 2.5D Digital Elevation Model From 2D Imagery; Eyes on the Earth 3D; Target Trailing With Safe Navigation for Maritime Autonomous Surface Vehicles; Adams-Based Rover Terramechanics and Mobility Simulator - ARTEMIS; ISTP CDF Skeleton Editor; Uplink Summary Generator (ULSGEN) Version 1.0; Robotics On-Board Trainer (ROBoT); Software Engineering Tools for Scientific Models; Automatic Data Filter Customization Using a Genetic Algorithm; Tracker Toolkit; Towards Efficient Scientific Data Management Using Cloud Storage; On a Formal Tool for Reasoning About Flight Software Cost Analysis; A Nanostructured Composites Thermal Switch Controls Internal and External Short Circuit in Lithium Ion Batteries; Spacecraft Crew Cabin Condensation Control; and Functional Near-Infrared Spectroscopy Signals Measure Neuronal Activity in the Cortex.

Source record↗
NASA NTRS2015

Hexavalent Chromium IV-Free Primer Development

Primer materials provide corrosion protection for metal parts as well as an increased adhesion between metallic substrates and thermal protection systems (TPSs). Current primers for use in cryogenic applications contain hexavalent chromium. This hexavalent chromium provides excellent corrosion protection even in a cryogenic environment, but it is a carcinogen that requires special equipment and waste control procedures to use. The hazardous nature of hexavalent chromium makes it an obsolescence risk in the future. This study included two phases of evaluation. Thirteen primers were initially identified as candidates and twelve of those primers were tested in phase 1. Four of the best performing candidates from phase 1 continued into phase 2 testing. Phase 1 testing consisted mostly of liquid constituent and physical property testing. Cryoflex and salt fog testing were included in phase 1 because of their importance to the overall success of a candidate material. Phase 2 consisted of physical, thermal, and mechanical properties for nominally processed and fabricated specimens.

Alldredge, Michael J.↗
DOE OSTI2019

A UO 2 -based salt target for rapid isolation of fission products

Here, we report here an initial isolation study based upon the use of small uranium oxide particles dispersed in a soluble salt matrix to evaluate the relative recovery of fission products into acidic media. We further show that the macrostructures of the uranium microparticles are largely preserved, such that the bulk target material could be retained for additional irradiations or characterizations. Through this approach, fission products can be separated from the actinide-based target using low molarities of acid without the need to dissolve the actinide itself, reducing the amount of acidic waste. Extraction yields using two molarities of HCl and HNO 3 are compared.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2021

Transformation Kinetics of Phosphorus and Nitrogen in Iron-Rich Sewage Sludges during Hydrothermal Treatment and Recovery of Nutrients from Process Water

Hydrothermal treatment (HT) is an emerging technique for sustainable sewage sludge management and resource recovery. Many sludges are rich in iron (Fe) due to the common addition of Fe salts in water resource recovery facilities. To develop guidance for reaction conditions targeting nutrient recovery, this study systematically investigated the influence of HT temperature, treatment time, and sludge source on the dynamic speciation evolution of phosphorus (P) and nitrogen (N) during HT of Fe-rich sewage sludge. Complementary chemical extraction and X-ray spectroscopy analyses were conducted to characterize the treatment products. For the sludge mixture (a blend of primary and waste activated sludges), P speciation did not change significantly within 4.5 h at 125 °C HT, while soluble and labile P was converted into insoluble P over time at 175 and 225 °C HT. Strengite (FePO 4 · 2 H 2 O) preferentially formed in the hydrochars with increasing treatment temperature and/or time, whereas 125 °C HT within 1.5 h favored the formation of vivianite (Fe 3 (PO 4 ) 2 ·8H 2 O). Organic P was completely decomposed into orthophosphate when the HT temperature reached up to 175 °C. Pyrrole-N was enriched in the hydrochars. Similar reaction pathways were observed during HT of anaerobically digested sludge, though some minor differences in Fe-associated P and organic P were observed. Meanwhile, HT of the two sludges released orthophosphate and ammonia into the process waters at 175 and 225 °C, which can be recovered by a sequential process involving struvite (MgNH 4 PO 4 ·6H 2 O) precipitation and air stripping. This study provides new insights into the transformation of P and N during HT of Fe-rich sludges as well as a modular design for maximum P and N recovery from the treatment products.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2025

Reducing waste of the hydride-dehydride process for U-6 wt% Nb spherical powders through lower impact and targeted milling

The breakdown of solid metal into powder during the hydride-dehydride process is commercially important for the formation of titanium and other metal powders. Typically, the milling of the brittle hydride powder occurs in a ball mill, where milling media impacts powder particles to break them down. The milling media can impact particles that are larger than desired, as desired, or smaller than desired, indiscriminately making all particles smaller. In this work, we investigate how to minimize waste powder production during milling using two different milling methods, planetary ball milling and milling in a sieve shaker (sieve-milling). Both processes yielded similar amounts of 20–75 μm diameter powder (the target size range); however, sieve-milling generated a significantly smaller amount of undersized waste powder. The powders were characterized by X-ray diffraction, SEM/STEM, and magnetic susceptibility. Several differences between ball milling and sieve-milling processes are discussed. We then conclude that the decreased yield of undersized powder in sieve-milling was due to a combination of lower impact energy in sieve-milling, unreacted metallic cores in the hydride flakes, and the ability to mill target particle sizes during sieve-milling. While these results are from the milling of brittle hydride powder, similar methods may be applicable to other brittle powders, including ceramics or salts.

Chemistry↗
DOE OSTI2019

Synthesis of Anhydrous Acetates for the Components of Nuclear Fuel Recycling in Dialkylimidazolium Acetate Ionic Liquids [plus supplemental information]

A series of anhydrous acetate salts with uranium {[C 2 C 1 im][UO 2 (OAc) 3 ] (1), [C 2 C 2 im][UO 2 (OAc) 3 ] (2), and [C 4 C 1 im][UO 2 (OAc) 3 ] (3)}, lanthanides {[C 2 C 2 im] 2 [La(OAc) 5 ] (4) and [C 2 C 1 im] 2 [Nd(OAc) 5 ] (5)}, and strontium {[C 2 C 1 im] n [Sr- (OAc) 3 ] n (6)} (where C 2 C 1 im = 1-ethyl-3-methylimidazolium, C 2 C 2 im = 1,3-diethylimidazolium, C 4 C 1 im = 1-butyl-3-methylimi-dazolium, and OAc = acetate) have been prepared and structurally characterized. Both lanthanides and strontium are common components of the nuclear fuel waste, and their separation from uranium is an important but still challenging task. A new synthetic method with dialkylimidazolium acetate ionic liquids (ILs) as the solvent has been developed for the direct synthesis of homoleptic acetates from the corresponding hydrates and, unexpectedly, hardly soluble f-element oxides. Although the group of characterized compounds shows perfect structural variability, all actinide and lanthanide metal ions form monomeric complex anions where the metal cation coordinates to five ligands including two oxygen atoms in the case of uranium, as is commonly observed for uranyl compounds. Crystallographic analyses revealed that the complex [UO 2 (OAc) 3 ] - anions possess rather standard D 3h symmetry featuring a hexagonal-bipyramidal coordination environment, while the lanthanide anions [Ln(OAc) 5 ] 2- are fully asymmetric and the Ln 3+ cations are 10-coordinated in the form of a distorted bicapped tetragonal antiprism. This is the first report of lanthanide ions coordinated in this fashion. For Sr 2+ , 9-fold coordination through oxygen atoms in the form of a strongly distorted tricapped trigonal prism is observed. Furthermore, the crystallization of anhydrous, homoleptic, anionic acetate complexes from such a large variety of different metal salts appears to be due to the properties of dialkylimidazolium acetate ILs themselves, including enhanced basicity from the high concentration of free anions and their greater affinity for hydrogen-bonding solutes relative to metal cations.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2018

Evaluating the Chemical Resistance of Salt-stone Disposal Unit (SDU) Concrete and Polymeric Coatings - 18494

Salt-stone Disposal Units (SDUs) are multi-million liter concrete structures that are used at the Savannah River Site (SRS) for the permanent disposal of low-activity salt solution incorporated in a cementitious grout matrix, termed salt-stone. The major chemical constituents of the salt solution include sodium, aluminum, nitrate, nitrite, hydroxide, carbonate, and sulfate. The potential exists for the SDU interior to be chronically exposed to these corrosive constituents resulting in deleterious reactions and physical damage. Sulfate attack is an example of such an interaction whereby aqueous sulfates may penetrate the concrete, and react with hydration products in the cured concrete to form expansive phases that can cause excessive strains and cracking. Recently constructed SDUs have utilized sulfate-resistant Type V cement and polymeric coatings to minimize the potential impacts of SDU concrete exposure to the highly caustic (pH > 13) salt solution. Not only must the concrete and coatings exhibit high resistance to the caustic salt solution but they are also subjected to above ambient temperatures due to the exothermic nature of the grout hydration reaction. As such, both the SDU concrete and the coatings need to be tested with respect to stability in the aforementioned chemical and thermal environments. Two primary investigations were considered in this work. The first was to evaluate the resistance of two previously used epoxy coatings (Hempel Versiline TL-45 S Novolac and EC-66 elastomeric epoxies) to chemical attack from a simulated salt solution. In addition, a third coating was included in the study, a poly-urea coating (Sherwin Williams Envirolastic AR425) that may be considered as an alternative to the epoxy coatings on future SDUs. Each coating was applied to concrete cylinders (actual SDU 6 test cores) and subsequently exposed to salt-stone, or full and half strength caustic salt solution; 50% dilution was considered since salt-stone processing involves addition of flush water to the SDUs, which is expected to dilute any residual salt solution or bleed water in the SDU. Exposure testing was conducted at 68 deg. C for 1,000 hours followed by ambient exposure to provide a full exposure time of approximately 90 days. It is important to note that evaluation criteria for the coated materials were only visual in nature. Other than coating discoloration, which was limited to the surfaces, no other visual indications of reactivity between the three coatings and any of the exposure mediums were observed. Cross-sectioned samples indicated no through-thickness coating penetration or reaction, and showed no signs of delamination. The second investigation was also performed to determine if SDU concrete without a polymeric barrier would resist ingress and reaction with salt solution at the elevated temperature. Again, only slight discoloration of the concrete surfaces was observed after the 90-day exposure test irrespective of the salt solution concentration. A requirement of structural concrete in the SDUs is that it possesses a minimum compressive strength of 42 MPa (6,000 psi). The average compressive strength of samples exposed to the full-strength salt solution was recorded at 68 MPa compared to 66 MPa for the unexposed samples thus indicating that the exposure conditions considered in this study were not detrimental to the concrete strength. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

DECOVALEX-2023: Task F1 Final Report

DECOVALEX-2023 Task F is a comparison of models and methods for post-closure performance assessment (PA) of a deep geologic repository for radioactive waste. The general aims of Task F are to build confidence in the models, methods, and software used for PA and to stimulate additional research and development in PA methodologies. The task objectives are to motivate development of PA modelling skills and capabilities, to examine the influence of model choices on calculated repository performance, and to compare the uncertainties introduced by model choices to other sources of uncertainty. Task F involves no actual experiment or site. It is a PA modelling exercise that requires the conceptual development of hypothetical repository designs and geologic settings. Because three of the teams were interested in salt and the rest of the teams were interested in crystalline rock, Task F was split into two branches: Task F1 for crystalline rock and Task F2 for salt. This report is for Task F1, crystalline rock. Teams from seven countries (Canada, Czech Republic, Germany, Korea, Sweden, Taiwan, and United States) participated in Task F1. The teams worked together to define the features, events, and processes of the reference case repository and established a set of performance measures. In addition, they defined a set of benchmark problems designed to test and compare modelling capabilities for fracture flow and transport at different scales. The repository design and benchmark problems are documented in a Task Specification that evolved over time as the group honed the specifications. The benchmark problems verified that each team can aptly model flow and transport in fractured media in 1-, 2-, and 3-dimensions. Two general approaches were used for the 3-dimensional benchmarks: discrete fracture network (DFN) and equivalent continuous porous medium (ECPM). DFN modelling involves explicit meshing of each fracture while ECPM modelling aims to capture the effective porosity and directional permeability of each cell in a space-filling mesh as affected by intersecting fractures. In some models, a combination of the two is used, i.e., DFN for large known fractures and ECPM for the rest of the domain. Transport is solved by using either the advection-dispersion equation or particle tracking. Although some variation is observed among model breakthrough curves in the benchmark problems, there is strong agreement in breakthrough behaviour up to at least the 75 th percentile for all benchmarks. At the 90 th percentile, breakthrough results show larger differences, suggesting several models retain substantially higher fractions of tracer in regions of slower moving water. In addition to the flow and transport benchmarks, several teams completed the source term benchmark, verifying capabilities for modelling radionuclide decay and ingrowth, waste package breach, instant release fractions, fuel matrix degradation rates, and radionuclide solubility limitations. The reference case is conceptualized as a generic spent fuel repository at a depth of 450 m in fractured crystalline rock. The repository has 50 parallel backfilled drifts, each with 50 deposition holes 6 m apart. Each deposition hole contains a 4-PWR waste package and bentonite buffer. The rock domain is 5 km in length, 2 km in width, and 1 km in depth. It has 6 deterministic fractured deformation zones and a multitude of stochastic fractures. Teams generally used the ECPM approach for the entire rock or a hybrid approach in which the deterministic fracture zones are modelled with a DFN and the rest of the rock is modelled by ECPM. Of the reference case problems specified, only the results of the initial reference case problem are compared in this report. The initial problem focuses on transport from the deposition holes to the surface, i.e., it neglects waste package performance. Tracers are released at all waste package locations at time zero and tracked for their releases to the near field and ground surface. The water fluxes calculated at the ground surface entry and exit regions of the domain are similar for all models except for two that have considerably lower fluxes. For tracer transport, large differences are observed among models in the magnitude of tracer transported. Much of the difference appears to be due to how the repository is implemented and hence the different degrees of repository simplification. Models that exclude the drifts, buffer, and backfill from the domain tend to show greater release of tracers and radionuclides from the repository. The initial study presented here indicates that major differences in modelling important processes within the repository (e.g., diffusion through buffer and backfill) can produce broadly different release and transport results, especially when those processes are excluded. Even for the models that included all specified features, events, and processes, the results show significant differences and demonstrate the importance of examining multiple modelling approaches in performance assessment. The differences in results observed in this study are expected to motivate teams to either increase complexity in future versions of the reference case models or to improve methods to account for the effects of simplified features and processes. Either way, future improvements in these models are expected to produce results that more closely agree.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Corrosion Testing of Refractory inContact with Molten Glasses Designed for Waste Vitrification - VSL Touchpoint Matrix Glasses

It is known that the predictive life of the refractory ceramic liner of nuclear waste glass melters is conservative, as demonstrated by performance of these materials such as in the Defense Waste Processing Facility (DWPF). The motivation for this task is to maximize the useful life of the melters that will be operated at the Waste Treatment and Immobilization Plant (WTP), which will in turn minimize procurement and disposal costs and melter outage times, as well as to identify maximum loadings in the waste glass of those species that corrode melter components. This task was initiated jointly with Pacific Northwest National Laboratory (PNNL) with the objective to develop a methodology and model to enable more accurate prediction of refractory service life under prototypic conditions from laboratory-scale material corrosion tests. Refractory corrosion is generally reported as physical material loss, measured in units of distance (e.g., inch) or as physical material loss rate, measured in units of distance per time (e.g., inch/day). In post-operational melters, the refractory corrosion is measured directly, sometimes reported as corrosion depth. Crucible tests are used in the laboratory to accelerate the refractory corrosion to facilitate a meaningful measurement in a commensurate amount of time. Crucible tests are particularly useful in understanding refractory corrosion across a large glass composition space, where operational testing would be prohibitive. Some of the critical parameters that are known to influence refractory corrosion by molten glass in a crucible test are temperature, system redox, molten salt phases, glass chemistry, and test duration. The majority of data collected for Monofrax® K-3 (hereafter referred to as K-3) corrosion is from crucible tests, but a small amount comes directly from scaled and production melters. Crucible test data has been collected under varying conditions, whereas data collected from operational melters is relatively fewer and represents conditions specific to the melter campaign. The result is that the published data can be grouped and analyzed in multiple ways, not all of which are readily comparable. The Standard Test Method for Isothermal Corrosion Resistance of Refractories to Molten Glass (ASTM C621) outlines the general guidelines used across industry. That method describes a sealed, static test in which the surface area of the refractory coupon and the volume of glass are fixed. A significant portion of the crucible data pertaining to nuclear waste glasses has been collected in a modified configuration; the most notable differences being the surface area of the refractory coupon to volume of the glass and use of a method for bubbling the melt. To our knowledge, the influence of those parameters on the refractory corrosion has not been quantified. In this work, it was determined that static tests and bubbled tests would be performed. Savannah River National Laboratory (SRNL) was tasked with setting up and performing static testing while PNNL was tasked with setting up and performing bubbled testing. Initial activities were performed to establish laboratory methods that reproduce data comparable to existing data sets of K-3 refractory corrosion by low activity waste (LAW) and high-level waste (HLW) glass compositions. Later activities were focused on refining the test parameters to establish a standard test practice to be used between Laboratories and collecting additional data to be used in the enhanced waste glass model development. This document serves primarily to convey the refractory loss measurement results from corrosion testing of K-3 refractory with waste glass compositions developed for use in the WTP melters. The data will be used in the enhanced property/composition models being developed for waste glass vitrification and melter operations.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Hydro-mechanical behavior of heated bentonite buffer for geologic disposal of high-level radioactive waste: A bench-scale X-ray computed tomography investigation

The behavior of heated bentonite buffer is critical for the security and long-term performance of a geological repository for high-level radioactive waste (HLW). While laboratory column experiments have been conducted to investigate compacted bentonite and coupled THMC (thermal-hydro-mechanical and chemical) processes for a moderate temperature range of up to 100 °C, data for a higher temperature range are limited. Understanding bentonite behavior and coupled THMC processes under higher temperatures (e.g., up to 200 °C) could allow for a more economic repository design and would expand the data and knowledge base for more reliable modeling. In this study, a bench-scale experiment was conducted in a compacted bentonite column experiencing both heating up to 200 °C in the center and hydration from a sand-clay boundary surrounding the column. During the experiment run for 1.5 years, frequent X-ray computed tomography (CT) scanning of bentonite provided insights into the spatiotemporal evolution of (1) hydration/dehydration, (2) clay swelling/shrinkage, (3) displacement, and (4) mineral precipitation. After the experiment, a comprehensive post-dismantling characterization of bentonite samples was conducted. Results showed that the bentonite hydration was axi-symmetrical despite the initial heterogeneity due to packing, confirming the ability of bentonite to seal fast flow/transport paths. Further, compared to a non-heated control experiment, the heated column showed greater CT density variations along the radial distance, indicating that homogenization of bentonite might be more difficult if a temperature gradient is maintained in the repository. Precipitation of an anhydrite layer occurred in the inner hot zone, pointing to potential concerns about salt precipitation causing canister corrosion. Ultimately, the experiments provided a high-resolution window into the strongly dynamic and coupled behavior of bentonite exposed to heating, hydration and swelling, which will be valuable for improving modeling of coupled processes, especially for the early state of a HLW repository.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Initial Results from the Third Round of Remediated Nitrate Salt Surrogate Formulation and Testing

High explosives science and technology (M-7) is currently working on the third round of formulation and testing of Remediated nitrate salt (RNS) surrogates. This report summarizes the calorimetry results from the 15% sWheat mixtures. All formulation and testing was carried out according to PLAN-TA9-2443 Rev B, "Remediated Nitrate Salt (RNS) surrogate formulation and testing standard procedure", released February 16, 2016. Results from the first and second rounds of formulation and testing were documented in memoranda M7-16-6042 and M7-16-6053.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Degradation of Polyester Polyurethane by Aspergillus sp. Strain S45 Isolated from Soil

A polyurethane (PU)-degrading fungus designated as strain S45 was isolated from a solid waste-dumping site in Islamabad, Pakistan. Strain S45 was identified as Aspergillus sp. through microscopic, morphologic as well as 18S rRNA sequencing (99% similarity with Aspergillus fumigatus). The degradation ability of strain S45, was analyzed through measurement of weight loss and evolution of CO{sub 2} as a result of breakdown of PU film. The degradation was confirmed through scanning electron microscopy, fourier transform infrared (FTIR) spectroscopy and differential scanning calorimetry (DSC). Surface changes like pits, cracks and holes were observed in scanning electron micrographs upon degradation of PU film by strain S45. FTIR spectra were carefully examined, changes were observed in various peaks at different wavelength ranges that also correspond to degradation of PU film. Differential scanning calorimetric technique was used to measure the melting temperature of PU film pieces before and after incubation with strain S45. DSC curve indicated an increase in melting temperature of PU from 191 to 196 °C that might be due to increase in crystallinity of the exposed PU after degradation of amorphous soft segment. A gradual increase in specific activity of esterase was observed up to the maximum of 15th day of incubation with PU film in mineral salt medium. Enzyme activity was further confirmed by tween 20 agar plates, demonstrated by formation of calcium complex. It is concluded from the results that Aspergillus sp. strain S45 might be applied for the treatment of plastics-contaminated environments.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2017

Reuse of Ion Exchange Resins for Reduction of Spent Radioactive Resins - 17125

A process is proposed to reuse ion exchange resins. The process consists of ion exchange beds for the ion exchange and regeneration, regeneration solution (regenerant) make-up tanks, an evaporator for concentration of spent regenerant, a dryer for the drying of concentrate, a pulverizer for grinding dried salt and a mixer for polymer solidification of the salt. A spent resins storage tank and a spent regenerant storage tank are also involved in the process. The role of each component in the process for reuse of the resins is explained. In addition, the effect of the reuse of ion exchange resins by regeneration was estimated in view of the reduction of spent resins. As the number of regenerations increases, the volume of spent resins is obviously inversely proportional to the number of regenerations. Although a polymer waste form is generated instead of spent resins when the ion exchange resins are reused, its volume is only about one-fifth of that of spent resins. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Transfer and Flushing Evaluation for Potential Precipitated Solids in the 241-AP-106 to EMF Transfer Pipeline for DFLAW

Waste transfer pipeline flushing is necessary to maintain pipeline performance, but added flush fluid increases the nuclear waste inventory. During the Direct Feed Low-Activity Waste (DFLAW) mission, waste transfer lines from the Interim Low-Activity Waste Storage Tank (i.e., 241-AP-106) to the Hanford Waste Treatment and Immobilization Plant (WTP) Low-Activity Waste (LAW) Facility must have the capability of being flushed to prevent accumulation of solids and to mitigate corrosion concerns. Operational experience at Hanford has shown that the solids precipitation occurs during processing operations from dilute Hanford liquids. Pacific Northwest National Laboratory was requested to evaluate the potential precipitated solid particle transport and flushing operational capabilities of the AP-106 through the Effluent Management Facility (EMF) low-point drain LAW feed pipeline. The evaluation results do not provide operational requirements, but rather provide a scoping basis for understanding the potential operational significance of solids precipitation in the pipeline. The salt and aluminum phase solids most likely to precipitate from the LAW feeds during cooling, evaporation, or mixing have particle densities that are estimated to range from 1.62 to 2.78 g/mL and a spherical particle size range of 8 to 2100 μm, depending on the solid phase considered. Application of the method to calculate the critical deposition velocity required by TFC-ENG-STD-26 at conservatively bounding estimates for the solids concentrations demonstrates that potentially precipitated solids may deposit on the bottom of the transfer pipe invert at the lower LAW feed pipeline flow rate, but the upper LAW feed pipeline flow rate will likely prevent deposition. The system pressure limit for the AP-106 through the EMF low-point drain LAW feed pipeline likely exceeds any pressure loss at a calculated critical deposition velocity even for the most adverse potential precipitated solid. However, should solids settle as a result of no-flow conditions, the flow capabilities of the AP-106 through the EMF low-point drain LAW feed pipeline are shown, based on the available literature, to be potentially inadequate for effective flushing operations to remove solids, even at solids concentrations below the maximum specified for LAW feed. Thus, stepwise solids accumulations over multiple transfers may be an issue, and eventual line plugging may occur. Further inadequacies may potentially be realized should the cohesive nature of certain precipitates be accounted for.

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