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Results for “high- level waste”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 343 records · Page 19
All sourcesNASA NTRSDOE OSTI
NASA NTRS1982

Feasibility of an earth-to-space rail launcher system

The feasibility of earth-to-space electromagnetic (railgun) launchers (ESRL) is considered, in order to determine their technical practicality and economic viability. The potential applications of the launcher include nuclear waste disposal into space, deep space probe launches, and atmospheric research. Examples of performance requirements of the ESRL system are a maximum acceleration of 10,000 g's for nuclear waste disposal in space (NWDS) missions and 2,500 g's for earth orbital missions, a 20 km/sec launch velocity for NWDS missions, and a launch azimuth of 90 degrees E. A brief configuration description is given, and test results indicate that for the 2020-2050 time period, as much as 3.0 MT per day of bulk material could be launched, and about 0.5 MT per day of high-level nuclear waste could be launched. For earth orbital missions, a significant projectile mass was approximately 6.5 MT, and an integral distributed energy store launch system demonstrated a good potential performance. ESRL prove to be economically and environmentally feasible, but an operational ESRL of the proposed size is not considered achievable before the year 2020.

Rice, E. E.↗
DOE OSTI2024

Insights into the thermo-hydraulic properties of compacted MX80 bentonite during hydration under elevated temperature

In high-level radioactive waste geological repositories, compacted bentonite undergoes coupled thermo-hydraulic processes due to heat released from a central waste canister and groundwater imbibition from the surrounding host rock. An understanding of these processes is essential for long-term simulations of radionuclide migration and canister corrosion, which requires an understanding of temperature effects on the coupled thermo-hydraulic properties governing these processes. In this study, a tank-scale radial infiltration test was used to investigate water imbibition processes in compacted bentonite under a central heater temperature of 200 °C that simulates high thermal gradients in a repository. Interpretation of this test focuses on evaluation of the liquid water wetting front during hydration and the interpretation of the temperature-dependent transient soil water retention curve (SWRC), thermal conductivity function (TCF), and hydraulic conductivity function (HCF). The SWRC during imbibition follows a temperature-dependent wetting path. Temperature effects on the hydraulic conductivity of bentonite in saturated conditions had the greatest effect on the shape of the HCF, with minimal temperature effects at higher suctions. The transient thermal conductivity data matched well with a new TCF linked with the SWRC shape. Although the bentonite layer was restrained, local deformations during hydration may have affected the shapes of the TCF and HCF.

Engineering↗
DOE OSTI2024

Modeling the Behavior of Noble Metals During HLW Vitrification in the DM1200 Melter

The noble metals ruthenium (Ru), rhodium (Rh), and palladium (Pd) are present to various extents in defense and commercial high-level nuclear waste streams. Usually, their concentrations are higher in commercial waste. Ru, Rh, and Pd are sparingly soluble in silicate glass melts and they therefore form separate metal, oxide, or other phases. These phases are generally denser than the glass melt and tend to sediment to the bottom of the melter, where they form a "sludge" layer. Since the sludge layer has a higher electrical conductivity than the molten glass, accumulation of sludge can ultimately lead to electrical disruption of the operation of the melter. In addition, the sludge reduces the melt volume, which also changes the operational characteristics of the melter. The noble metals sludge is a highly viscous mixture of melt and noble metal phases. The concentration of noble metals in the sludge may be 10 to 100 times higher than that in the drained glass.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Impacts of Lanthanide Ion Complexation on the Radiation Robustness of Diglycolamide Extractants

Nuclear power reactors represent a clean and reliable source of baseload energy. However, nuclear reactors produce used nuclear fuel (UNF), which, if not recycled/reprocessed, must be disposed of as high-level radioactive waste, necessitating long-term storage options. UNF contains approximately a third of the periodic table, including the radioactive minor actinides (MA) americium and curium. Removal of these MA from UNF would greatly reduce the radiotoxic burden, volume, and cost of a storage facility or geological repository. Thus, a significant global effort has been devoted to the design of reprocessing strategies for the clean separation of the MA from UNF. As MA separations are difficult, due to the presence of lanthanide elements in UNF and their remarkably similar chemical properties, many extractants have been studied. The diglycolamide (DGA) class of extractants are promising for this separation as they exhibit high distribution coefficients for the MAs, high metal loading capacities, and are robust in highly acidic radiation environments.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2020

99 TcO 4 - removal from legacy defense nuclear waste by an alkaline-stable 2D cationic metal organic framework

Removal of 99 TcO 4 - from legacy defense nuclear tank waste at Savannah River Site is highly desirable for the purpose of nuclear safety and environmental protection, but currently not achievable given the extreme conditions including high alkalinity, high ionic strength, and strong radiation field. Herein, we present a potential solution to this long-term issue by developing a two-dimensional cationic metal organic framework SCU-103, showing ultrahigh stability in alkaline aqueous media and great resistance to both β and γ radiation. More importantly, it is very effective for 99 TcO 4 - separation from aqueous media as demonstrated by fast exchange kinetics, high sorption capacity, and superior selectivity, leading to the successful removal of 99 TcO 4 - from actual Savannah River Site high level tank waste for the first time, to the best of our knowledge. In addition, the uptake mechanism is comprehensively elucidated by molecular dynamics simulation and density functional theory calculation, showing a unique chemical recognition of anions with low charge density.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2025

Mk-IV Salt Crystallization Hot Finger Apparatus for Partitioning Used Electrorefiner Salt

Electrorefining is a controlled redox process used to regulate the behavior of ionic species. Through this process, metals can be deposited onto a cathode from an electrolyte solution in a controlled manner. The Mk-IV electrorefiner (Mk-IV ER) at Idaho National Laboratory is an engineering-scale, molten salt-based electrorefining cell that has been used for decades to recover metallic uranium from spent fuel. As a result, highly stable fission product chlorides have accumulated in the electrolyte. This accumulation results in changes to the salt’s properties, such as melting temperature, thermal conductivity, and density, as well as elevated product impurity and fissile materials criticality margin. These factors prompt the need for a salt regeneration process, such as melt-crystallization and species drawdown. This work focuses on providing a conceptual design to regenerate ER salt from used Mk-IV-ER salt in-situ, while minimizing salt waste volumes by concentrating the fission products in a final processed salt heal. We propose using a hot-finger crystallization apparatus design to fractionally crystallize salt in the Mk-IV-ER head space (or baffle space), allowing the collection of solid and liquid fractions. By using a cup-drain design, the used salt will be allowed to slowly solidify on the walls of a stainless-steel cup. The apparatus drain plug will then open to allow the liquid salt phase to drain to a lower cup, effectively separating the liquid phase from the solid phase. Under the hypothesis that the liquid phase salt concentrates the fission products, which is under examination in the accompanying work package, this separation allows the recovered solid salt to be reused while minimizing the high-level salt waste volume of used ER salt.

36 - MATERIALS SCIENCE↗
DOE OSTI2020

Insights into the mechanisms controlling the residual corrosion rate of borosilicate glasses

Borosilicate glasses are widely used to confine high-level radioactive wastes. The lifetime of these materials could reach hundreds of thousands of years if leaching of the glass into groundwater enables the formation of a passivating gel layer. Even in this regime, the glass will never stop corroding as thermodynamic equilibrium between glass and solution cannot be achieved. Therefore, accurate predictions of glass durability including passivation, require a deep understanding of the mechanisms controlling the so-called residual rate. However, despite tremendous efforts, these mechanisms remain poorly understood. Here, focusing on the behavior of the soluble elements of the International Simple Glass (B, Na, and Ca), we show that the residual rate is controlled by the behavior of B, a glass former supposed to dissolve instantaneously when in contact with water and thus widely considered as an ideal tracer. We then demonstrate that B release is controlled by multiple processes highly dependent on the pH. At the beginning of the passivating layer formation, the hydrolysis of B-O-Si linkages is rate-limiting and has an activation energy of ~60 kJ mol –1 , a value slightly lower than that for breaking Si-O-Si linkages. Once the fraction of closed pores resulting from gel restructuring is high enough, then diffusion of both reactants (water molecules) and some products (mainly B aq , Ca aq ) through the growing gel layer becomes rate-limiting. Consequently, B and Ca accumulate in an inner layer referred to as the active zone, with potential feedback on the B-O-Si hydrolysis. A new paradigm, including B as a key element of the system, is proposed to develop a comprehensive model for the corrosion of borosilicate glass.

36 MATERIALS SCIENCE↗
DOE OSTI2024

KOMPASS-II: Compaction of Crushed salt for Safe Containment – Phase 2

Long-term stable sealing elements are a basic component in the safety concept for a possible repository for heat-emitting radioactive waste in rock salt. The sealing elements will be part of the closure concept for drifts and shafts. They will be made from a welldefinied crushed salt in employ a specific manufacturing process. The use of crushed salt as geotechnical barrier as required by the German Site Selection Act from 2017 /STA 17/ represents a paradigm change in the safety function of crushed salt, since this material was formerly only considered as stabilizing backfill for the host rock. The demonstration of the long-term stability and impermeability of crushed salt is crucial for its use as a geotechnical barrier. The KOMPASS-II project, is a follow-up of the KOMPASS-I project and continues the work with focus on improving the understanding of the thermal-hydraulic-mechanical (THM) coupled processes in crushed salt compaction with the objective to enhance the scientific competence for using crushed salt for the long-term isolation of high-level nuclear waste within rock salt repositories. The project strives for an adequate characterization of the compaction process and the essential influencing parameters, as well as a robust and reliable long-term prognosis using validated constitutive models. For this purpose, experimental studies on long-term compaction tests are combined with microstructural investigations and numerical modeling. The long-term compaction tests in this project focused on the effect of mean stress, deviatoric stress and temperature on the compaction behavior of crushed salt. A laboratory benchmark was performed identifying a variability in compaction behavior. Microstructural investigations were executed with the objective to characterize the influence of pre-compaction procedure, humidity content and grain size/grain size distribution on the overall compaction process of crushed salt with respect to the deformation mechanisms. The created database was used for benchmark calculations aiming for improvement and optimization of a large number of constitutive models available for crushed salt. The models were calibrated, and the improvement process was made visible applying the virtual demonstrator.

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

Enhanced crevice corrosion of stainless steel 316 by degradation of Cr-containing hollandite crevice former

This study explores the corrosion interactions between stainless steel 316 (SS316), a candidate nuclear waste canister material, and Cr-containing hollandite (Ba 1.15 Cr 2.3 Ti 5.7 O 16 ), a model ceramic nuclear waste form. The two materials were corroded in close proximity in 0.6 M NaCl solution at 90 °C for 28 days. Severe crevice corrosion was observed on the stainless steel surface after the corrosion experiment, which is likely associated with the rapid accumulation of Cr 3+ cations originating from both SS316 and the Cr-containing hollandite. Furthermore, the enhanced corrosion of SS316 by the corrosion products of hollandite is quantitatively explained using a classical crevice corrosion model.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Stability of plutonium oxide nanoparticles in the presence of montmorillonite and implications for colloid facilitated transport

The physical and chemical stability of PuO 2 nanoparticles (intrinsic colloids) in groundwater will control their transport and may affect the performance of high-level radioactive waste repositories. In this study, we examined the chemical stability of two types of PuO 2 nanoparticles at both 25 and 80 °C. The “alkaline” PuO 2 nanoparticles were prepared by neutralizing an acidic Pu(IV) solution with dilute NaOH to pH ~9.5. The “acidic” PuO 2 nanoparticles were precipitated from 0.1 M nitric acid by heating a Pu(IV) solution at 60–80 °C for 30 min. The chemical stability of these PuO 2 nanoparticles was tested in the presence of montmorillonite, a common mineral in the environment and potentially relevant backfill material in some nuclear waste repository designs. The “alkaline” PuO 2 nanoparticles were found to be unstable over a timescale of months at both 25 and 80 °C, with elevated temperature enhancing their dissolution rates and sorption to montmorillonite. PuO 2 nanoparticle dissolution rates decreased with increasing Pu concentration, consistent with solution saturation. The “acidic” PuO 2 nanoparticles appeared to remain stable for much longer than the “alkaline” PuO2 nanoparticles. The “alkaline” PuO 2 nanoparticle dissolution rate constants were as high as 10 -11.9±0.4 mol m -2 s -1 and 10 -11.2±0.2 mol m -2 s -1 at 25 and 80 °C, respectively, while the “acidic” PuO 2 nanoparticle dissolution rate at 80 °C was 10 -13.5 mol m -2 s -1 . Based on transmission electron microscopy, the “alkaline” and “acidic” PuO 2 nanoparticles were of similar size (2.5–4.5 and 2–3 nm nanoparticles, respectively). However, the “acidic” PuO 2 nanoparticles formed more ordered nanoparticle aggregates. Our results suggest that the specific conditions experienced during PuO 2 nanoparticle formation could significantly affect the stability of PuO 2 in the presence of competing sorption processes and, in turn, the relative importance of intrinsic versus pseudocolloid transport in the environment.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2024

Refinement of Pitting Factor Basis to Evaluate Minimum Nitrite Requirement in Dilute Waste Chemistries

Considering that chloride concentrations are an order of magnitude lower at SRS than the maximum chloride previously tested to evaluate pitting susceptibility of carbon steel, it has been proposed that the Pitting Factor methodology be refined to determine if a reduction in the minimum nitrite requirement is warranted in more dilute solutions and those with chloride concentrations more indicative of waste tank chemistries at SRS. Through the use of cyclic potentiodynamic polarization, the susceptibility of A537 carbon steel to pitting corrosion was evaluated. The Pitting Factor was shown to accurately predict pitting susceptibility throughout the chemistry envelope evaluated. In addition, it was shown that nitrite is not required for inhibition of pitting corrosion in tank conditions with chloride concentrations observed at SRS and up to 4 M nitrate. Based upon these findings, new control limits were proposed that removed the minimum nitrite requirement within the chemistry envelope evaluated up to 50 °C.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Implementation and verification of a user-defined element (UEL) for coupled thermal-hydraulic-mechanical-chemical (THMC) processes in saturated geological media

Efficient and accurate modeling of the coupled thermal-hydraulic-mechanical-chemical (THMC) processes in various rock formations is indispensable for designing energy geo-structures such as underground repositories for high-level nuclear wastes. Here, this work focuses on developing and verifying an implicit finite element solver for generic coupled THMC problems in geological settings. Starting from the mass, momentum, and energy balance laws, a specialized set of governing equations and a thermoporoelastic constitutive model is derived. This system is then solved by an implicit finite element (FE) scheme. Specifically, the residuals and the Jacobians are scripted in a user-defined element (UEL) subroutine which is then combined with the general-purpose FE software Abaqus Standard to solve initial-boundary value problems. Considering the complexity of the system, the UEL development follows a stepwise manner by first solving the coupled hydraulic-mechanical (HM) and thermal-hydraulic-mechanical (THM) equations before moving on to the full THMC problem. Each implementation step consists of at least one verification test by comparing computed results with closed-form analytical solutions to ensure that the various coupling effects are correctly realized. To demonstrate the robustness of the algorithm and to validate the UEL, a three-dimensional case study is performed with reference to the in-situ heating test of ATLAS at Belgium in 1980s. A hypothetical radionuclide leakage event is then simulated by activating the chemical-concentration degree of freedom and prescribing a constant high concentration at the heater's surface. The model predicts a limited contaminated regime after six years considering both diffusion and advection effects on species transport.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Prediction of Long-Term Geochemical Change in Bentonite Based on the Interpretative THMC Model of the FEBEX In Situ Test

Since nuclear energy is crucial in the decarbonization of the energy supply, one hurdle to remove is the handling of high-level radioactive waste (HLW). Disposal of HLW in a deep geological repository has long been deemed a viable permanent option. In the design of a deep geological repository, compacted bentonite is the most commonly proposed buffer material. Predicting the long-term chemical evolution in bentonite, which is important for the safety assessment of a repository, has been challenging because of the complex coupled processes. Models for large-scale tests and predictions based on such models have been some of the best practices for such purposes. An 18-year-long in situ test with two dismantling events provided a unique set of chemical data that allowed for studying chemical changes in bentonite. In this paper, we first developed coupled thermal, hydrological, mechanical, and chemical (THMC) models to interpret the geochemical data collected in the in situ test and then extended the THMC model to 200 years to make long-term prediction of the geochemical evolution of bentonite. The interpretive coupled THMC model shows that the geochemical profiles were strongly affected by THM processes such as evaporation/condensation, porosity change caused by swelling, permeability change, and the shape of concentration profiles for major cations were largely controlled by transport processes, but concentration levels were regulated by chemical reactions, and the profiles of some species such as pH, bicarbonate, and sulfate were dominated by these reactions. The long-term THMC model showed that heating prolongs the time that bentonite becomes fully saturated in the area close to the heater/canister; however, once the bentonite becomes fully saturated, high concentrations of ions in bentonite near the heater, which was observed in the field test, will disappear; illitization continues for 50 years but will not proceed further.

Zheng, Liange↗
DOE OSTI2023

Kinetics of the Temperature-Dependent e aq - and ·OH Radical Reactions with Cr(III) Ions in Aqueous Solutions

The reactivity of chromium(III) species with the major oxidizing and reducing radiolysis products of water was investigated in aqueous solutions at temperatures up to 150 °C. The reaction between the hydrated electron (e aq - ) and Cr(III) species showed a positive temperature dependence over this temperature range. The reaction was also studied in pH 2.5 and 3.5 solutions for the first time. This work also studied the reaction between acidic Cr(III) species and the hydroxyl radical (·OH). It was found that Cr 3+ did not react significantly with the ·OH radical, but the first hydrolysis species, Cr(OH) 2+ , did with a rate coefficient of k= (7.2±0.3)×10 8 M -1 s -1 at 25 °C. The oxidation of Cr(OH) 2+ by the ·OH radical formed an absorbing product species that ultimately oxidized to give Cr(VI). In conclusion, these newly measured reaction rates allow for the development of improved models of aqueous chromium speciation for the effective remediation of liquid high-level nuclear waste via vitrification processes.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2020

Temperature response and brine availability to heated boreholes in bedded salt

Abstract There is a growing need for disposal of high‐level nuclear waste. To reduce uncertainty associated with brine availability to repository excavations in salt formations, a collaboration between Sandia, Los Alamos, and Lawrence Berkeley National Laboratories is performing a series of borehole‐scale coupled process tests. Here, we report on the first round of the Brine Availability Test in Salt (BATS) project, a “shakedown” experiment called Phase 1s. Experimental testing included placing a resistive heater, a 260‐W radiative heater, and a 750‐W radiative heater within previously drilled horizontal boreholes at the Waste Isolation Pilot Plant (WIPP) while monitoring temperature and water inflow. The experiments successfully achieved the targeted temperature of 120 °C when using the 750‐W radiative heater. Simulations using FEHM (Finite Element Heat and Mass transfer code) and TOUGH‐FLAC (Transport Of Unsaturated Groundwater and Heat–Fast Lagrangian Analysis of Continua) were able to accurately predict the coupled thermo–hydro–mechanical–chemical response of salt, matching the observed temperature and brine production. Due to the extremely low permeability of salt, these systems take many years to reach steady state when perturbed by mining activities. Long‐term numerical simulations are used to develop the initial pressure and saturation conditions. The inclusion of a damaged rock zone with higher permeability around the borehole also affects the saturation and pressure distributions and plays an important role in dissipating the potential for thermal pressurization. Knowledge gained from this round of experimentation and simulation will be used to conduct the next BATS project experiment in new boreholes at WIPP.

58 GEOSCIENCES↗
DOE OSTI2025

Evaluation of desiccation cracking characteristics of inorganic micro-fiber-reinforced engineered barrier material (IMEBM) for geological repository

Abstract Buffer material is crucial for the engineering barrier system to dispose of high-level radioactive waste in a geological repository. A reliable buffer material should be able to maintain good sealing characteristics and minimize desiccation cracking. In this study, the effectiveness of inorganic fiber-reinforced engineering barrier material in reducing desiccation cracks in bentonite was evaluated via desiccation tests, image analysis, and air permeability tests. The effects of fiber type (E-glass fiber and basalt fiber) and fiber content (0.0%, 0.5%, 1.0%, 1.5%, 3.0%, and 5.0% of dry weight of the bentonite) on the development of desiccation cracks in the fiber–bentonite mixtures with the same given initial moisture content were evaluated. The results indicated that the addition of fibers could significantly reduce the crack size and area in bentonite during the drying. Basalt fibers showed a slightly better reinforcement effect than E-glass fibers when the fiber content was lower than 3.0%. The addition of fibers prevented the development of penetrating cracks and significantly reduced the permeability of the bentonite after drying. The permeabilities of basalt fiber- and E-glass fiber-reinforced bentonite composites with 3% reinforcement were 5.81 × 10 –11 m 2 and 7.24 × 10 –11 m 2 , respectively, which were 64 and 51 times smaller than that of pure bentonite. X-ray–CT observation of the internal structure of the samples after drying showed that the addition of fibers significantly changed the crack morphology and potentially increased the tortuosity.

Feng, Yuan↗
DOE OSTI2022

Modeling gas breakthrough and flow phenomena through engineered barrier systems using a discrete fracture approach

Compacted clays are being considered to build engineered barrier systems (EBS) intended for the safe isolation of high-level nuclear waste (HLW) and spent nuclear fuel (SNF). The corrosion of the metallic canister containing the HLW/SNF will lead to the generation and buildup of the gas pressure in the more internal part of the clay buffer. This phenomenon would eventually trigger the formation and propagation of fractures in the clay barrier, jeopardizing its safety functions. Here, in this work, we propose to use the fragmentation technique (MFT) to model evolving fractures in clays triggered by gas pressurization. The MFT has been successfully used to model the formation of fractures in concrete, drying cracks in soil, hydraulic and thermo-fractures in rocks. In this work, we extend the MFT to deal with multiphase fluid flow in deformable porous media, and we upgraded a fully coupled computer finite element code using the extended technique. The proposed approach is first verified against analytical solutions and is then applied to model gas breakthrough experiments in clays. A very satisfactory performance of the method is observed in all the analyses, showing the potential of the MFT to tackle multiphase flow problems in deformable porous media with evolving discontinuities.

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

Numerical simulation of coupled THM behaviour of full-scale EBS in backfilled experimental gallery in the Horonobe URL

Bentonite-based engineered barrier system (EBS) is a key component of many repository designs for the geological disposal of high-level radioactive waste. Given the complexity and interaction of the phenomena affecting the barrier system, coupled thermo-hydro-mechanical (THM) numerical analyses are a potentially useful tool for a better understanding of their behaviour. In this context, a Task (the Horonobe EBS experiment) was undertaken to study, using numerical analyses, the thermo-hydro-mechanical (and thermo-hydro) interactions in bentonite based engineered barriers within the international cooperative project DECOVALEX 2023. One full-scale in-situ experiment and four laboratory experiments, largely complementary, were selected for modelling. The Horonobe EBS experiment is a temperature-controlled non-isothermal experiment combined with artificial groundwater injection. The Horonobe EBS experiment consists of the heating and cooling phases. Six research teams performed the THM or TH (depended on research team approach) numerical analyses using a variety of computer codes, formulations and constitutive laws. Finally, for each experiment, the basic features of the analyses are described and the comparison between calculations and laboratory experiments and field observations are presented and discussed.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗