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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 199 records · Page 11
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Investigation of Methane Generation Rates from Simulated and Radioactive Waste at Evaporator Conditions

Savannah River National Laboratory researchers have performed 9 experiments with simulated waste and 13 experiments with radioactive waste to determine methane generation rates (MGR) applicable to Savannah River Site waste at temperatures greater than or equal to 100 °C. Data from these experiments was used to generate temperature-dependent expressions to conservatively account for methane generation at elevated temperatures. Measured MGRs and the derived expressions are reported in units of standard cubic feet per hour per gallon of solution, where standard conditions are 25 °C and 1 atm.

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

Design and Materials of Reference Electrodes for Radioactive Waste Tank Service – A Literature Review

The Hanford site stores approximately 55 million gallons of radioactive and chemically hazardous wastes from the production of weapons materials. The wastes are stored in 177 underground, carbon steel storage tanks, 149 of these are single shell tanks (SSTs) and 28 of these are double shell tanks (DSTs). The DSTs provide critical retrieval and interim storage before the waste is vitrified in the Waste Treatment and Isolation Plant (WTP). The DSTs have been in service for 38 to 56 years and current plans indicate that WTP operations will be completed in 2075. Thus, the tanks will need to remain in service far beyond the initial 40-year life expectancy. For life extension of the tanks, effective corrosion control practices must remain in force. This effort includes direct measurements of the extent of corrosion (e.g., ultrasonic measurements and corrosion coupons) and electrochemical processes (e.g., linear polarization measurements).

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

Processes in Salt Repositories for Radioactive Waste Disposal

This document summarizes the key processes (thermal, hydrological, mechanical, and chemical; THMC) impacting the features of a deep geological repository for radioactive waste in salt. Some processes are natural and on-going whether the repository is there or not, and other processes are driven by the perturbation associated with the repository. The features considered here include both engineered and natural components of the repository system. The engineered barrier system (EBS) in a salt repository is quite different from those implemented for a repository in clay or crystalline rocks, because it is comprised mostly of granular salt and salt-compatible cements, rather than bentonite. When compared to other rocks (i.e., silicates), salt has unique properties that make it an excellent potential host rock. Openings and fractures in salt creep closed readily. Salt has high thermal conductivity, which can reduce peak temperatures. Additionally, far away from the excavations the porosity of salt is unconnected, which leads to essentially zero advective or diffusive transport. The small amount of hypersaline brine occurring in salt minimizes microbial activity, reduces colloid-assisted transport, and eliminates in-package criticality (i.e., chloride is a neutron poison). At the end of the report, we present a brief outline for a potential salt repository, including considerations avoided in previous repository disposal concepts. We propose considering higher-temperature processes in future disposal concepts, rather than trying to minimize the thermal perturbation of the repository. Since hot salt is drier, a dry repository would limit corrosion, gas generation, and solute transport. Openings and fractures creep shut faster in hot salt. Therefore, higher temperatures could be seen as beneficial, rather than something to minimize, through increased spacing between waste packages (increasing repository costs).

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

Cliff Retreat Rates Associated with a Low-Level Radioactive Waste Disposal Facility in Los Alamos, New Mexico, USA

We present an analysis and interpretation of potential cliff stability at a low-level waste disposal facility at Los Alamos National Laboratory, New Mexico, using cliff morphologic and fracture characteristics coupled with carbon-14 surface exposure dating. Our study is important as it directly bears on the licensing criteria for low-level radioactive waste sites. We find that future characteristic cliff failures will likely not breach disposal pits and shafts over the 1000-year minimum regulatory period. Further, we find, using a multivariate regression model, that slope angle and cliff face aspect are sub-equal in importance to predict regions of high risk of failure when combined with surface exposure ages and assuming that old exposure ages are most indicative of stability (instability) and therefore can aid decision making in final design implementation.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

NEVADA NATIONAL SECURITY SITE 2020 DATA REPORT- GROUNDWATER MONITORING PROGRAM AREA 5 RADIOACTIVE WASTE MANAGEMENT SITE

This report presents groundwater and leachate sample results from the Area 5 Radioactive Waste Management Site (RWMS) at the Nevada National Security Site in Nye County, Nevada. Since 1993, groundwater samples have been collected and static groundwater depths have been measured from the aquifer immediately below the Area 5 RWMS. The data are evaluated for evidence of effects on the aquifer related to the Area 5 RWMS. Leachate from the Cell 18 lined mixed waste cell has been sampled since 2011, and leachate from the Cell 25 lined mixed waste cell was first sampled in 2019 after it began receiving waste in August 2018. Leachate data are analyzed for hazardous contaminants to determine appropriate leachate handling and disposal. This report includes five years of data from 2016 through 2020.

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

NEVADA NATIONAL SECURITY SITE 2021 WASTE MANAGEMENT MONITORING REPORT AREA 3 AND AREA 5 RADIOACTIVE WASTE MANAGEMENT SITES

Environmental monitoring data are collected at and around the Area 3 and Area 5 Radioactive Waste Management Sites (RWMSs) within the Nevada National Security Site (NNSS). This report summarizes the 2021 environmental data to provide an overall evaluation of RWMS performance and to support environmental compliance and performance assessment (PA) activities.

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

Coupled Thermo-Hydrological-Mechanical-Chemical Behavior of Anisotropic Granite for Geologic Disposal of High-Level Radioactive Waste: A Core-Scale Laboratory Investigation

The coupled thermo-hydrological-mechanical-chemical (THMC) behavior of rock within an Excavation Damaged Zone (EDZ) is critical for the safety and long-term performance of a geological repository for high-level radioactive wastes. While many laboratory experiments have been conducted to investigate EDZ rocks, the flow and deformation characteristics resulting from anisotropic rock textures and microcrack distribution under triaxial loading and elevated temperatures remain poorly understood. Particularly, cracks at various scales serve as fast paths for fluid flow and solute transport and present as focal points of mechanical weakness, which complicate the coupled THMC processes in anisotropic EDZ rocks and challenge modeling predictions. Here, in this study, a series of core-scale experiments was conducted on three granite samples under repository-relevant conditions. These rock samples were obtained from the Grimsel Underground Research Laboratory (URL), featured by anisotropic minerals (represented by bedding layers) and microcrack distributions and coarse cm-scale grain sizes. During the experiments, samples were subjected to an elevated temperature at 90 °C and different triaxial loading conditions either by radial (normal to bedding layers) or axial (parallel to bedding layers) compaction. Water was injected into the samples, and the rock permeability evolutions and effluent water chemistry were monitored closely. For intact samples, thermal expansion of minerals at 90 °C resulted in a large, 75% irreversible permeability reduction and rock strengthening under radial compaction, while thermal impact was limited to a 15% permeability reduction under axial compaction. In contrast, for a sample containing open cracks, the growth of fractures during the experiment resulted in an abrupt permeability increase and fast failure at 90 °C. The effluent water chemistry indicates much more considerable mineral dissolution from large shear sliding than that in rocks dominated by mechanical compaction. These results helped better understand the coupled THMC processes in anisotropic rocks containing cracks, evaluate the behaviors of EDZ rocks, and predict the long-term evolution of EDZ for the performance of the repositories.

Coupled THMC processes↗
DOE OSTI2022

Carrier gas attenuation in gamma assay of radioactive xenon samples

Gamma attenuation attributable to xenon and carbon dioxide carrier gas in radioactive xenon assay samples has been calculated using Monte Carlo n-Particle (MCNP) transport code for the 0.65 mL flame sealed ampoule (bean), and 15 mL and 50 mL Schlenk tube geometries. The attenuation was correlated as a % loss per standard mL of carrier gas. In this work, high Purity Ge gamma spectroscopy experiments were performed to measure against the MCNP results. The experimental results from the 133 Xe in xenon carrier gas in a bean geometry showed 0.97 ± 0.4% loss per standard mL xenon, which was in agreement with the MCNP predictions.

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

Mercury sequestration in alkaline salt low-level radioactive waste

Liquid low-level radioactive waste at the Savannah River Site contains several species of mercury, including inorganic, elemental and methylmercury. This waste is solidified and stabilized in a cementitious waste form referred to as saltstone. Soluble mercury is stabilized as β-cinnabar, HgS as the result of reaction between the mercury and sulfur present in blast furnace slag, one of the cementitious regents. In this investigation, Mersorb®, a commercial granular activated carbon impregnated with sulfur, was evaluated as a pretreatment to remove mercury from the solution prior to cementation. Mersorb® was found to remove more than 96 mass percent of the methylmercury in simulated tank waste solution when the mass ratio of Mersorb® to mercury was above 2.5. Slag sequestered relatively more inorganic mercury than organic mercury in simulated tank waste after 24 hours of contact. This is likely due to the mercury-carbon bond being more covalent than the mercury-oxygen bond and therefore more difficult to break and slower to form HgS.

Cementitious Material↗
DOE OSTI2025

Eddy covariance towers as sentinels of abnormal radioactive material releases

Ensuring accurate detection and attribution of abnormal releases of radioactive material is critical for protecting human health and safety. Most commonly, such detection is accomplished via active monitoring approaches involving the collection of physical samples. Further, this is labor intensive and limits the temporal and spatial resolution of any detected events to a relatively coarse level. As an alternative first step towards passive monitoring, we developed an approach using eddy flux tower data records to identify signals from a known abnormal release and quantify the extent to which that signal also occurs at other times in the data record. Through two case studies, one of which targeted the Fukushima nuclear disaster and the other targeting an abnormal release event at a radioisotope production facility in Fleurus, Belgium, we tested our approach and identified several potential heretofore unidentified abnormal events that were consistent with atmospheric circulation patterns and/or wind direction from known release sites. Because our approach is relatively simple and is resistant to systematic errors in the observational record, it has broad applicability beyond specific constituents and ecosystem types to identify a wide variety of limited-duration anomalies in flux tower data to ensure human health and industrial safety.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2022

Experimental validation of the mechanistic scale-up methodology of gas–solid spouted beds using radioactive particle tracking (RPT)

The very high-temperature reactors (VHTRs) are highly ranked among candidates of Generation IV of nuclear reactors due to their high efficiency, safety, the resistance to proliferation, and reliability. The VHTRs are preferentially fueled by Tristructural-isometric (TRISO) coated fuel particles which has fuel kernels of fissile material coated by four coating layers: a porous buffer pyrolysis carbon layer (buffer PyC), an inner dense pyrolysis carbon layer (IPyC), a silicon carbide layer (SiC) and an outer dense pyrolysis carbon layer (OPyC). The heart of the operation and safety of the VHTRs significantly depends on the reliability of the coating layers of TRISO particles to retain metallic and gaseous fission products within the particles. The technique used for coating TRISO particles are gas-solids spouted beds via chemical vapor deposition (CVD). Fabrication of high-quality low-defect TRISO fuel particles fuel at larger scale spouted beds is required to support the commercialization of the VHTRs. In this work, our new developed mechanistic scale-up methodology of gas-solids spouted beds based on matching the radial profile of gas-holdup has been demonstrated and validated using radioactive particle tracking (RPT). Two spouted beds of small and larger scales were used in the study. Three sets of conditions were carried out which include the conditions of the reference case in the large scale, conditions that provide similar gas holdup radial profile to that of the reference case and conditions that provided dissimilar gas holdup radial in the small-scale spouted beds. The results confirm the validation of the scale-up methodology in terms of the dimensionless values of the spout diameter, cumulative probability distribution of the solids particles penetration into the spout, fraction of cycle time in each region of the bed, the radial profiles of the dimensionless values of the root-mean-square particle velocities and solids eddy diffusivity. Finally, the results further advance the knowledge and understanding of the gas-solids spouted beds provide deeper insight into their solids dynamics and presenting important benchmarking data for validating computational fluid dynamics codes and models. At last, procedures are established for the implementation of the new scale-up methodology.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Chemical degradation of fly ash blended concrete with the seasonal variation of rainwater in a radioactive waste repository: A thermodynamic modeling approach

Highlights: • The thermodynamic model for the cementitious system was successfully developed. • The chemical degradation of concrete was greatly affected by rainwater types. • Inorganic carbon species in rainwater reduced the chemical resistance of concrete. • The durability of concrete to rainwater decreased with increasing fly ash content. This study presents a long-term degradation behavior of fly ash blended concrete in a vault type low- and intermediate-level radioactive waste (LILW) repository by thermodynamic equilibrium calculations using PHREEQC combined with CEMDATA18 database. Since rainwater plays a predominant role as leachate for the concrete waste container in the repository, the effect of chemical properties of seasonal rainwater on the chemical degradation of concrete was analyzed. In addition, the impact of the blending with fly ash on the chemical degradation was evaluated through a gradual replacement of ordinary Portland cement (OPC) by fly ash. Regardless of the replacement level, autumn rainwater showed the greatest influence on the concrete degradation owing to the presence of inorganic carbon species, which induce the carbonation. The resistance of concrete to chemical degradation by rainwater was persistently reduced due to the decreasing formation of portlandite and calcium silicate hydrate (C-S-H) with increasing replacement level.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Enhancement of Thermal Conductivity of Bentonite Buffer Materials with Copper Wires/Meshes for High-Level Radioactive Waste Disposal

In high-level radioactive waste disposal, a heat-generating waste canister is generally encased with a layer of bentonite-based buffer material acting as an engineered barrier to limit water percolation and radionuclide release. The low thermal conductivity of bentonite (~0.5 W/m∙K) combined with a high thermal loading waste package may result in a high surface temperature on the package that can potentially impact the structural integrity of the package itself as well as the surrounding buffer material. We show in this paper that the thermal conductivity of bentonite can be effectively enhanced by embedding copper wires/meshes across the buffer layer to form fully connected high heat conduction pathways. A simple calculation based on Rayleigh’s model shows that a required thermal conductivity of 5 W/m∙K for effective heat dissipation can be achieved simply by adding ~1 vol % of copper wires/meshes into bentonite. As a result, the peak surface temperature on a large waste package such as a dual-purpose canister can be reduced by up to 300°C, leading to a significant reduction in the surface storage time for waste cooling and therefore the overall cost for direct disposal of such waste packages. Because of the ensured full thermal percolation across the buffer layer, copper wires/meshes turn out to be much more effective than any other materials currently suggested (such as graphene or graphite) in enhancing the thermal conductivity of buffer material. Furthermore, the embedded copper wires/meshes can help reinforce the mechanical strength of the buffer material, thus preventing the material from a potential erosion by a possible intrusion of dilute groundwater.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Adsorption of Radioactive Iodine Using Nanocarbon on ETS-10 as Adsorbent

Here, laboratory-synthesized nanocarbon pelletized with titanosilicate (ETS-10) as a support matrix has been investigated for the capture of radioactive iodine present as methyl iodide (CH 3 I) in the off-gas streams produced during aqueous reprocessing of used nuclear fuel. The mass fraction of carbon in the sorbent matrix was 0.10. The effects of residence time and CH 3 I concentration were investigated using a continuous flow column setup to quantify the adsorption and desorption capacities of adsorbent under dynamic conditions from an air stream containing CH 3 I present at concentrations representative of those expected in the off-gas streams. Air with CH 3 I gas as a source in the column resulted in quantifiable CH 3 I adsorption with 0.98 mg/g of adsorption capacity. Laboratory-made nanocarbons had a larger adsorption capacity than those of the other carbons reported in the literature. Additionally, the adsorption capacity of nanocarbon on ETS-10 is compared to that of nanocarbon coated on cordierite in previous studies.

ETS-10↗
DOE OSTI2021

Performance-based simulants for Hanford radioactive waste treatment process testing

Radioactive wastes from nuclear fuel processing are stored in large underground storage tanks at the Hanford Nuclear Reservation. Treatment and remediation require that waste feed from the storage tanks be delivered to the Hanford Tank Waste Treatment and Immobilization Plant continually for the duration of the treatment mission. Furthermore, the complex physical and chemical processes required for this mission, the significant scale, and the hazardous nature of the waste necessitate the use of simulants in process testing. A new class of simulants, “performance-based simulants,” has been developed to match the process performance of the simulant to actual waste performance data.

Hanford tank waste treatment↗
DOE OSTI2024

Transfer functions for Q A /Q B international regulatory limits for the safe transport of radioactive materials

This paper presents a proposed revision of the International Atomic Energy Agency transport regulations, related to the A 1 and A 2 limit values used to determine the radioactive transport classification. Based on the 'Q system', a novel methodology was introduced to derive Q A and Q B values related to scenarios involving external exposure from a distant source. These values are key parameters that respectively represent the total effective dose and total equivalent dose to the skin, from all primary and secondary particles contributing to radiation exposure. The International Working Group (WG A 1 /A 2 ) is established and associated with the TRANSSC Technical Expert Group on Radiation Protection. A review of the A 1 and A 2 values is performed in response to identified limitations within the existing Q system. The followed approach is based on Monte Carlo simulations that enabled the development of transfer functions aimed at reducing computational time and increasing the flexibility of dose evaluations for any radionuclide with known particle emission spectra. This method allows updating the Q A and Q B values to account for future data evolutions (decay data, fluence-to-dose conversion coefficients) and standardizing the calculation of regulation limits across all referenced radionuclides and scenarios related to external exposure. The transfer functions are established using three Monte Carlo simulation codes—FLUKA, Geant4, and MCNP—and address the previous limitations of the 'Q system', reflecting the latest International Commission for Radiation Protection recommendations and improvements in calculation techniques. The results of the WG show consistent agreement across the codes, with minor discrepancies observed at low primary energies due to statistical uncertainties and different handling of stopping power for electrons/positrons in the codes. This revised approach aligns with current standards and recommendations, ensuring that the radiological consequences of transport accidents are acceptable for the new A 1 and A 2 limits from a radiological protection perspective.

61 RADIATION PROTECTION AND DOSIMETRY↗