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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 OSTI2018

Improved ZnS nanoparticle properties through sequential NanoFermentation

Sequential NanoFermentation (SNF) is a novel process which entails sparging microbially produced gas containing H 2 S from a primary reactor through a concentrated metal-acetate solution contained in a secondary reactor, thereby precipitating metallic sulfide nanoparticles (e.g., ZnS, CuS, or SnS). SNF holds an advantage over single reactor nanoparticle synthesis strategies, because it avoids exposing the microorganisms to high concentrations of toxic metal and sulfide ions. Also, by segregating the nanoparticle products from biological materials, SNF avoids coating nanoparticles with bioproducts that alter their desired properties. Herein, we report the properties of ZnS nanoparticles formed from SNF as compared with ones produced directly in a primary reactor (i.e., conventional NanoFermentation, or “CNF”), commercially available ZnS, and ZnS chemically synthesized by bubbling H 2 S gas through a Zn-acetate solution. The ZnS nanoparticles produced by SNF provided improved optical properties due to their smaller crystallite size, smaller overall particle sizes, reduced biotic surface coatings, and reduced structural defects. Furthermore, SNF still maintained the advantages of NanoFermentation technology over chemical synthesis including scalability, reproducibility, and lower hazardous waste burden.

59 BASIC BIOLOGICAL SCIENCES↗
DOE OSTI2026

Toward the performance assessment of advanced nuclear waste forms: temperature dependence of lanthanide borosilicate glass dissolution

Lanthanide borosilicate (LaBS) glasses are among the most promising waste forms for the immobilization of high-level radioactive waste generated from advanced nuclear fuel cycles. However, the temperature dependence of their dissolution kinetics remains poorly understood and constrained, limiting the integration of these materials into established performance assessment models. Here, we investigate the dissolution behavior of the legacy AmCm2-19 LaBS glass and the benchmark alkali aluminoborosilicate ISG-1 in deionized water between 50 °C and 250 °C using ASTM C1285 (Product Consistency Test-B) protocols. For AmCm2-19 LaBS glass, normalized elemental release rates for boron and silicon increase with temperature before plateauing near 150 °C, consistent with solubility-limited behavior. From data obtained at 50 °C and 100 °C, Arrhenius analysis yields activation energies of E a (B) = 24.8 ± 0.3 kJ mol⁻¹ and E a (Si) = 14.4 ± 0.2 kJ mol⁻¹, similar or slightly lower than those previously reported for two other compositions of LaBS glasses. No secondary phases or alteration layers were detected by SEM-EDX or pXRD. These results establish one of the first temperature-dependent kinetic datasets for LaBS glass dissolution, providing quantitative parameters to inform mechanistic corrosion models and predictive simulations of glass degradation in geological disposal environments.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Technetium, Iodine, and Chromium Adsorption/Desorption Kd Values for Vadose Zone Pore Water, ILAW Glass, and Cast Stone Leachates Contacting an IDF Sand Sequence

Performance and risk assessments of immobilized low-activity waste (ILAW) at the Integrated Disposal Facility (IDF) have shown that risks to groundwater are quite sensitive to adsorption-desorption interactions occurring in the near- and far-field environment. These interactions between the underlying sediments and the contaminants present in the leachates that descend from the buried glass, secondary waste grouts, and potentially Cast Stone low-activity waste packages have been represented in these assessments using the contaminant distribution coefficient (Kd) construct. Some contaminants (99Tc, 129I, and Cr) present in significant quantities in these wastes have low Kd values and tend to drive risk to public health and the environment. Relatively small changes in the Kd value can cause relatively large changes in the retardation factor. Thus, even relatively small uncertainty in the Kd value can result in a relatively large uncertainty in the risk determined through performance assessment modeling. The purpose of this study is to further reduce the uncertainty in Kd values for 99Tc, iodine (iodide and iodate), and Cr (chromate; CrO42-) by conducting systematic adsorption-desorption experiments using actual sand-dominated Hanford formation sediments from beneath the IDF and solutions that closely mimic Hanford vadose zone pore water and leachates from Cast Stone and ILAW glass waste forms. Twenty-four batch and 21 flow-through column experiments were conducted, yielding 261 Kd measurements for these key contaminants, and contributing to our understanding for predicting transport from wastes disposed to the IDF. While the batch Kd methodology is not well-suited for measuring Kd values for non-sorbing species (as noted by the U.S. Environmental Protection Agency), the batch Kd results presented here are not wholly inconsistent with the column Kd results, and could be used for sensitivity purposes. Results from the column experiments are consistent with the best estimate and lower range of Kd values reported by Krupka et al. and Cantrell et al.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Chemical Characterization and Removal of Carbon-14 from Irradiated Graphite II - 13023

Approximately 250,000 tonnes of irradiated graphite waste exists worldwide and that quantity is expected to increase with decommissioning of Generation II reactors and deployment of Generation IV gas-cooled, graphite moderated reactors. This situation indicates the need for a graphite waste management strategy. Of greatest concern for long-term disposal of irradiated graphite is carbon-14 (C-14), with a half-life of 5730 years. Study of irradiated graphite from some nuclear reactors indicates C-14 is concentrated on the outer 5 mm of the graphite structure. The aim of the research presented last year and updated here is to identify the chemical form of C-14 in irradiated graphite and develop a practical method by which C-14 can be removed. A nuclear-grade graphite, NBG-18, and a high-surface-area graphite foam, POCOFoam{sup R}, were exposed to liquid nitrogen (to increase the quantity of C-14 precursor) and neutron-irradiated (10{sup 13} neutrons/cm{sup 2}/s). Finer grained NBG-25 was not exposed to liquid nitrogen prior to irradiation at a neutron flux on the order of 10{sup 14} /cm{sup 2}/s. Characterization of pre- and post-irradiation graphite was conducted to determine the chemical environment and quantity of C-14 and its precursors via the use of surface sensitive characterization techniques. Scanning Electron Microscopy (SEM) was used to evaluate the morphological features of graphite samples. The concentration, chemical composition, and bonding characteristics of C-14 and its precursors were determined through X-ray Photoelectron Spectroscopy (XPS), Time-of-Flight Secondary Ion Mass Spectrometry (SIMS), and Energy Dispersive X-ray Analysis Spectroscopy (EDX). Results of post-irradiation characterization of these materials indicate a variety of surface functional groups containing carbon, oxygen, nitrogen and hydrogen. During thermal treatment, irradiated graphite samples are heated in the presence of an inert carrier gas (with or without oxidant gas), which carries off gaseous products released during treatment. Graphite gasification occurs via interaction with adsorbed oxygen complexes. Experiments in argon were performed at 900 deg. C and 1400 deg. C to evaluate the selective removal of C-14. Thermal treatment also was performed with the addition of 3 and 5 volume % oxygen at temperatures 700 deg. C and 1400 deg. C. Thermal treatment experiments were evaluated for the effective selective removal of C-14. Lower temperatures and oxygen levels correlated to more efficient C-14 removal. (authors)

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2012

Simulation of flow and mercury transport in Upper East Fork Poplar Creek, Oak Ridge, Tennessee

As a result of nuclear processing activities started back in the 1950s, the environment in the vicinity of the Y-12 National Security Complex (Y-12 NSC) in Oak Ridge, Tennessee, and surrounding watersheds has been contaminated by nearly 1,000 tons of elementary mercury. To comply with the state and federal surface water quality standards, a significant reduction in mercury concentration to parts-per-trillion levels has been proposed. In order to analyze the mercury cycle in the environment and provide forecasting capabilities for the flow and transport of mercury within the Upper East Fork Poplar Creek (UEFPC) watershed, an integrated surface and subsurface flow and transport model has been developed using the hydrodynamic and transport numerical package, MIKE, developed by the Danish Hydraulic Institute. The model has been constructed and calibrated using an extensive collection of historical records (i.e., hydrological data, and mercury concentration measurements in groundwater, soil, and sediment) obtained from the Oak Ridge Environmental Information System database. Daily fluctuations in stream flow, as a result of scattered rainfall, flooding, and flow augmentation, resuspend the contaminated streambed sediments and/or erode the polluted streambank soil and provide a secondary source of mercury to the creek. In order to investigate the significance of sediment-mercury interactions on the fate and transport of mercury within the UEFPC study domain, simulations were performed for two different cases (i.e., with and without consideration of sediment-mercury interactions). Computed total suspended solids and mercury concentrations at the integration point of the creek are compared with the corresponding historical records in both cases. As confirmed by the numerical simulations, a substantial portion of the mercury detected in the river is likely in the form of sediment particle–bound mercury (i.e., mercury particulates).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Twenty years of dry interim storage of spent nuclear fuel (SNF) in Germany- History and developments - 15575

According to article 6 of the German Atomic Energy Act, storage of nuclear fuel requires a license granted by the competent authority in this field, which is the Federal Office of Radiation Protection. In 1983 a license to store spent nuclear fuel (SNF) and later expanded for vitrified high level waste (HAW) for the interim storage facility in Gorleben was granted. Dry Interim storage at the NPP sites in its actual form started in 2002 in the interim storage facility next to the NPP Lingen. Since this time each NPP erected its own storage facility and three central storage facilities have been built. All of these facilities use dry storage in metallic casks. The actual storage licenses for SNF and the HAW (which is stored in Gorleben and Rubenow) are limited to 40 years starting with the first cask emplacement. As no opening of the cask is intended during the storage period the condition of the inventory and the cask before loading has to be known in detail. This requires for instance detailed knowledge of the history of the SNF loaded as well as supervision of the manufacturing process of the cask. Within the last twenty years several mishaps influenced the regulatory point of view. During a loading campaign in a boiling water reactor traces of pool water remained in the gasket slot and during the drying procedure crystalline boron salts prevented the required tightness. After that event a long-term experiment was started and is still going on which studies the effects of water and minor damages to the gasket. As well changes in the loading procedure have been introduced. The pressure switch used to monitor the pressure in the space between primary and secondary lid has a self surveillance function. This indicated a malfunction 23 times in the last twenty years. Examinations of all events indicated no systematic failure, but a weak point seems to be the plug-in connection. The manufacturer introduced some modifications to improve the system. Introduction of FEM methods in safety analysis to amend the existing methods of testing allowed a new view on and in the cask safety analysis. Progress in this field is closely connected to the available computing capacity and stipulates experts to ongoing improvement. After expiring of the storage license which will be the case for the first cask loaded in 1992 and stored in Ahaus 2032 an extension or a new license can only be granted if the at this time relevant requirements are met. This will require detailed knowledge about the happenings and accidents during the previous storage period. All experience of the periodic safety assessment (PSA) which started in 2013 and accompanying experiments conducted till now and during the licensed storage time will be needed. (authors)

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

Additive Manufacturing: The Rise of a Technology and the Need for Quality Control and Inspection Techniques

Additive manufacturing (AM) technology, also known as three-dimensional (3D) printing, rapid prototyping (RP), or solid free forms (SFF), was developed back in the early 1980s. The technology was quickly commercialized in the late 1980s, and since then, the technology has been widely used in different applications at several industrial sectors. AM offers distinct advantages from traditional, subtractive manufacturing techniques. Traditional subtractive manufacturing techniques, including milling, drilling, and turning, are techniques that start with a large block of stock material, which is then subtracted from until the final desired product is reached, leaving much of the initial block as wasted scrap. In the AM process, the part is constructed by depositing material layer by layer in the Z-direction until the final product is produced, leaving little to no waste. The process starts with a computer-aided design (CAD) model that gets sliced into individual layers using software that generates instructions, known as G-code, which are sent to the AM machine. Here, this process allows complex geometry construction that traditional methods would have diffi-culty accomplishing or fail completely. In some cases, a secondary finishing process is needed such as polishing, sintering, curing, sanding, powder removal, or painting.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Full Scale HEPA Filter Encapsulation in Ultra-High-Performance Grout - Proof of Concept

The Hanford Tank Waste Treatment and Immobilization Plant (WTP) currently being constructed to treat radioactive waste, includes vitrification facilities for both the high-level waste (HLW) and low activity waste (LAW) fractions. Operation of the WTP will produce contaminated high-efficiency particulate air filters (HEPA), as part of the solid secondary waste (SSW) stream. The HEPA filters receive off-gas from the vessel vent header and primary off-gas treatment system in LAW Facility and remove particulate contaminants including 99 Tc and 129 I salts. The current disposal method for HEPA filters is encapsulation in metal containers using cementitious material (CM) and disposal in the Integrated Disposal Facility (IDF). Results from the 2017 IDF Performance Assessment (PA) WRPS (2018) demonstrated that while compliance is maintained for the 1000 year compliance period mandated by DOE O 435.1 and its accompanying manual, release of constituents from the HEPA filters result in exceedance of the performance objective imposed as the groundwater regulatory limit at later times. For example, at about 1500 years post-closure, solid secondary waste (SSW), including HEPA filters is predicted to become a dominant contributor to 99 Tc release and over the 10,000-year sensitivity analysis period, SSW is the dominant contributor of 129 I release to the groundwater. The estimated release could potentially be reduced if the HEPA filters are not compacted and waste containers could be distributed throughout a large space, thereby diluting the contaminant release. Additionally, a better cementitious material could be used to encapsulate the HEPA filters. One alternative method for disposal of contaminated HEPA filters is encapsulation of the filters in ultra high-performance grout (UHPG). UHPG is a variation of ultra-high-performance concrete (UHPC) is commonly used in the prestressed concrete industry for large structural members. Recent studies of UHPG show it has excellent properties for containing radionuclides such as 99 Tc and 129 I Nichols and Kaplan (2021). This report presents the results of the first attempt to encapsulate a clean, full-size HEPA filter in UHPG and evaluate the effectiveness of the immobilization process and final waste form. A full-scale proof-of-concept simulated waste form was prepared by encapsulating a HEPA filter in a 110-gallon stainless steel (SS) drum using UHPG. A change from Type I/II PC to Type 1L PLC was made after American Rock Products informed the team that they would no longer be using Type I/II by the end of 2024 and the northwest was phasing out Type I/II PC overall. Type I/II PC used in previous studies of UHPG for encapsulation (Nichols and Kaplan 2021). After the UHPG was cured both the scaled mockup and the full-scale simulated waste forms were sectioned for visual examination. UHPG completely encapsulated the filters and bonded to the external surfaces of materials comprising the filters. No cracks were observed in the sectioned waste forms.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Tailored Perovskite Waste Forms for Plutonium Trapping

Perovskite ceramics have been extensively studied as host matrixes for radionuclide entrapment for nuclear waste disposal. As an expansion of these investigations, cerium, neodymium, and plutonium were incorporated into a perovskite phase, ACu 3 FeTi 3 O 12 (A = Nd, Ce, Pu), using sol–gel methods under oxidizing and reducing atmospheres. The targeted materials contained varying levels of Ce 3+ and Nd 3+ on the A site, yielding potential compositions of Nd 1–x Ce x Cu 3 FeTi 3 O 12 (x = 0, 0.1, 0.2, 0.3, 0.4, 0.8). However, interrogation of these materials shows that the maximum Ce 3+ loading is achieved near x ≈ 0.2. A single composition with plutonium was targeted, Nd 0.9 Pu 0.1 Cu 3 FeTi 3 O 12 , in order to properly model more realistic loading levels for a repository-destined material. These compounds were characterized using powder X-ray diffraction with Rietveld refinements of the structures and by a variety of spectroscopic techniques. Finally, the data suggest that, in order to achieve Pu 3+ substitution onto the A sites in the Nd 0.9 Pu 0.1 Cu 3 FeTi 3 O 12 , a reducing atmosphere must be employed. Otherwise, the redox activity of plutonium results in substitution onto multiple sites in the material as well as the formation of secondary phases such as TiO 2 .

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2015

Enhanced chemical CRUD decontamination combined with electrokinetic process - 15242

This study has tested about the deposition and dissolution of Chalk River Unidentified Deposit (CRUD), and investigated the remediation of CRUD. CRUD is an accumulated corrosion product on external fuel rod cladding surfaces and composed of either dissolved ions or solid particles such as Ni and Co. As nuclear power plants are getting older, decontaminating process has been concerned due to increased contamination. Corrosion product can be activated through nuclear reaction to form radionuclide such as Co-60. Because the CRUD is formed at cladding surface and radionuclides can be contained in CRUD, it can affect to human health and environment. Therefore CRUD needs to be removed safely during NPP's decontaminating process by efficient. In order to decontaminate the CRUD, first thing it has to be dissolved and removed. Deposition behavior of corrosion products has been well studied, but the study about behavior of CRUD dissolution is required in various environments. In this study, the CRUD synthesis, the dissolution method of CRUD, and the CRUD decontamination method was investigated. There are three major approaches the remediation of CRUD. First one is well synthesized the CRUD, second one is finding the optimum condition for CRUD dissolution, and last one is applying the Electro-Kinetic (EK) process to CRUD remediation. Synthesis of simulated CRUD for the decontamination experiments was successful. The combination of EK process with chemical decontamination of CRUD proved to be a more economical method for removing CRUD; it can be directly applied to the NPP decommissioning and decontamination process, and it can decrease the volume of secondary radioactive waste as well. (authors)

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

Catalytic Consequences of Hierarchical Pore Architectures within MFI and FAU Zeolites for Polyethylene Conversion

The benefits of hierarchical zeolites for the conversion of bulky molecules like polymeric waste have been reported in the literature; however, the impact of mesopore sizes and connectivities on rates, product selectivities, and catalyst deactivation in the context of plastic upcycling has not been systematically probed. Here, in this study, we synthesized a suite of hierarchical MFI and FAU zeolites via desilication under varying conditions for metal-free polyethylene conversion reactions under batch and flow conditions (473–523 K). Polyethylene (solid) conversion rates (normalized by Bro̷nsted acid site density) were higher on hierarchical than parent microporous MFI regardless of mesopore connectivities, i.e., open or constricted, suggesting that the incorporation of mesopores facilitates diffusion of intermediate products to access medium-pore protons for successive scission events. Furthermore, higher branched:linear gaseous product ratios were produced on hierarchical than parent MFI, since mesopores allow for egress of bulkier molecules without undergoing further secondary events, e.g., isomerization back to linear alkanes/alkenes or beta scission. Solid conversion rates on hierarchical FAU synthesized via desilication with cetyltrimethylammonium bromide (CTABr), however, were not higher than parent FAU, likely because the presence of CTABr facilitates recrystallization of leached species to form composites (hierarchical FAU and ordered mesoporous materials) with more isolated mesopores. The stagnation in rates, despite increased mesopore volumes (>0.22 cm 3 g –1 ), highlights the importance of confinement effects provided by micropores for cleaving C–C bonds at modest reaction conditions. In situ 1 H MAS NMR performed on polyethylene with MFI zeolite show that PE isomerizes (and potentially deconstructs) at temperatures near 450 K, highlighting the role of Bro̷nsted acid sites in activating C–C bonds under mild reaction conditions. Catalyst recyclability studies showed that all catalysts undergo deactivation during plastic upcycling reactions, but to varying extents. Overall, hierarchical materials have better catalyst stability than parent materials, although the differences in stability between hierarchical and parent FAU are smaller than those for MFI. Taken together, these findings demonstrate how rates, selectivities, and catalyst deactivation from plastic upcycling reactions can be controlled via fine-tuning the identity and connectivity of mesopores.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2021

Acetonitrile Destruction and Fate of Organics in the Reverse Osmosis System at the ETF

The Hanford Site Effluent Treatment Facility (ETF) currently treats aqueous waste streams that include condensates from the 242-A evaporator, leachate from the Environmental Restoration Disposal Facility (ERDF), as well as laboratory wastes and, in the future, will treat liquid effluents from the Hanford Tank Waste Treatment and Immobilization Plant (WTP) and Integrated Disposal Facility (IDF) leachate. Liquid effluents from the WTP will have significant concentrations of acetonitrile. Acetonitrile is formed by reaction of nitrates and sugar in the WTP low activity waste (LAW) melters and is prevalent in the submerged bed scrubber (SBS) and wet electrostatic precipitator (WESP) liquid effluents from WTP off-gas treatment. When these liquids are concentrated in the WTP Effluent Management Facility (EMF) evaporator in the direct feed low activity waste (DFLAW) flow-sheet, testing has shown that the majority of the acetonitrile partitions to the evaporator condensate. Since the evaporator condensate is directed to the ETF, this creates a potential issue with the ETF waste acceptance criteria. Consequently, there is a need to validate flow-sheet assumptions on the fate of acetonitrile and other organics within the ETF. The present plan includes the addition of a steam stripper to the ETF to remove acetonitrile. There is, therefore, also a need to determine a suitable method to destroy acetonitrile in the overhead condensate stream from the new steam stripper. Washington River Protection Solutions, LLC (WRPS) previously contracted with Atkins and the Vitreous State Laboratory (VSL) of The Catholic University of America (CUA) to perform development and testing work to evaluate potential methods for destruction of acetonitrile in WTP secondary liquid effluents. Based on the results of that work, WRPS requested that follow-on testing be conducted to further evaluate acetonitrile destruction in the steam stripper condensate using ultraviolet oxidation (UV/OX) with persulfate. WRPS also requested testing to assess the rejection rate of organics in the reverse osmosis (RO) system installed in the ETF. This report presents the results from testing to address those needs.

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

In-situ X-ray and visual observation of foam morphology and behavior at the batch-melt interface during melting of simulated waste glass

We report to attain a basic understanding of the primary foam structure and behavior, which affects the heat and mass transfer and the efficiency of the glass melting process, we investigated the primary foam layer under the glass batch floating on molten glass. The primary foam affects mass transfer during batch melting, in turn affecting the melting process. The recently performed direct in-situ three-dimensional X-ray computed tomography of the batch melting in a laboratory-scale melter vessel allowed us to visualize the features of the reacting batch layer and the foam that develops at its bottom, though with an insufficient resolution of images. In this study, we obtained better temporal and spatial resolution using the two-dimensional X-ray radiography and visual observation of the structure and behavior of transient primary foam as it formed and decayed. As soon as the batch was charged onto the melt surface, foam bubbles began to evolve, grow, and coalesce, forming a primary foam layer, 5-10 mm thick, within tens of seconds. This foam layer was sustained by ongoing gas evolving reactions counterbalanced by bubble coalescence into cavities that moved sideways and escaped to the atmosphere. Eventually, the entire remaining batch turned into foam that gradually decayed at the melt surface. The decay rate agreed with literature observations of surface foam produced by secondary foaming.

36 MATERIALS SCIENCE↗
NASA NTRS2015

“PowerCell”: The Interface Between Mars Resources and Human Exploration

The barriers to forming human settlements on Mars are high but surmountable within our lifetime. While the Apollo astronauts carried their life support with them, our success in exploring and forming settlements on Mars depends on our ability to use local Martian resources to generate the materials and conditions humans need to survive, so-called in situ resource utilization (ISRU). On Earth, biology provides us with food, shelter, oxygen, and other materials. Off-planet, synthetic biology will enable numerous parallel productions: optimized food production, water treatment, air treatment, environmental monitoring, regolith biomining, waste management, cell based biomaterial production, biocementation, and in situ synthesis based on received DNA sequences. How will the organisms responsible for these synthetic production systems obtain organic carbon and fixed nitrogen in the hostile Martian environment? We envision a synthetic-biology enabled Martian colony and introduce here the critical intermediate component a biological power source needed to transform the in situ resources found on Mars into biological feedstocks to enable growth of production organisms. Here, we present our first PowerCell, a photosynthetic and nitrogen-fixing filamentous cyanobacterium engineered to provide a carbon-rich fuel source for a biological life support system on Mars. We provide a vision of how the PowerCell system will operate in a Martian colony based on ground experiments and preparations for testing in space as a NASA secondary payload aboard the upcoming DLR Eu:CROPIS satellite mission experiments.

Rothschild, Lynn J.↗
DOE OSTI2022

Impact of aqueous solution pH on network structure of corrosion–induced surface layers of boroaluminosilicate glass

Understanding of the long-term stability of vitrified nuclear waste glass is critical to its safe storage, but knowledge gaps remain regarding the corrosion mechanism, especially with respect to surface alteration layers whose presence changes the observed corrosion rates. Here, we report the aqueous corrosion of a model boroaluminosilicate glass at ground water relevant pHs as investigated by vibrational spectroscopy to understand the network structure of the alteration layer that we hypothesize controls the corrosion rates. Both specular reflection infrared spectroscopy and infrared variable angle spectroscopic ellipsometry were used to elucidate details of the structure of the alteration layers formed at pH 7 and pH 9. After 2312 days, the thickness of the alteration layer is found to be ~2000 nm for corrosion at pH 7 and ~300 nm for corrosion at pH 9. Here, the interpretation of the vibrational spectra suggests that alteration layers formed at pH 7 and pH 9 both exhibit a porous network with "silica-like" structures, but with differences in the bond lengths and bond angles in the glass network. Moreover, the relative abundance of the hydrous species within the porous network appears to be dependent on pH. Additionally, the spectroscopic data suggest that the B and Ca trapped in the alteration layer during the secondary corrosion of the pH 9 pre-corroded sample in pH 3 solution are hydrated within the nano-porous layer and not bound to the silicate network. However, the precise chemical form remains to be determined. Based on these findings, we propose that the combination of the network structure of the alteration layer and the relative abundance of hydrous species within the alteration layer act to control the transport of ionic species dissolved from the reaction front to the aqueous solution outside the glass.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Tritiated Water Disposition Strategies

Tritium, a radioactive isotope of hydrogen, is typically confined within one or more confinement barriers. Tritium readily reacts with oxygen to form tritiated water or can exchange with protium or deuterium atoms in water to create tritiated water which is around 10,000 times more toxic than elemental tritium. Primary tritium confinement systems can create small volumes of highly tritiated water. Secondary confinement systems, like inert gloveboxes, will produce larger volumes of lower concentration tritiated water. Much larger volumes of tritiated water of very low tritium content are produced by tritium permeation/diffusion. Tritium from inside primary and secondary confinement barriers will diffuse from these systems into cooling water systems. Closed loop, recirculating cooling water systems will slowly accumulate tritium and create concentrations which exceed drinking water (or permitted discharge) concentrations. The cooling water systems designed as series of interconnected cooling loops that can slowly transfer tritium from one cooling loop to another cooling loop through heat exchangers. These large volumes of tritiated water may be above federal discharge limits and also lack convenient disposition pathways. This document will discuss tritiated water management strategies from the perspective of large volume, low tritium concentrations that exceed drinking water or permitted discharge limits. Applicable US regulatory requirements for a Molybdenum-99 (Mo-99) production facility concerning tritiated water are also cited. Tritiated water disposition options, based on typical waste minimization strategies will guide dispositions options. A Case Study example is presented to illustrate the pros and cons of several disposition options. The report does not recommend any particular disposition pathway but discusses strategies to be explored by a producer to guide selection of processes which best fits the needs of a particular waste stream.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Performance Assessment Results for the Hanford Integrated Disposal Facility - 19134

In 2014, the US DOE Office of River Protection and its contractors began to develop a performance assessment for the near-surface disposal of low-level and mixed low-level waste at the Hanford Integrated Disposal Facility. The Hanford Integrated Disposal Facility is a doubly-lined landfill that was constructed between 2004 and 2006 to be the disposal facility for the vitrified low-activity waste containers that will be produced at the Hanford Waste Treatment and Immobilization Plant. The disposal facility is also expected to receive solid secondary waste produced at the Hanford Waste Treatment and Immobilization Plant and other solid wastes from site activities. The constructed facility is awaiting authorization to receive waste. Although no waste destined for the facility is currently being generated, operations cannot begin until the facility has received a Disposal Authorization Statement from US DOE and a permit modification from the State of Washington Department of Ecology. Both the Disposal Authorization Statement and permit modification require a demonstration that the system of engineered and natural features of the disposal facility will limit releases from the facility and be protective of human health and the environment for the next 1,000 to 10,000 years. Based on the analyses presented in the Integrated Disposal Facility Performance Assessment, US DOE issued a conditional Disposal Authorization Statement for the facility in June 2018. Demonstrating the long-term performance of the facility to be protective of human health and the environment was evaluated under the requirements of US DOE Order 435.1, Radioactive Waste Management. Computer simulations were performed to evaluate whether or not the disposal facility would comply with the performance objectives specified in the order. The evaluation was performed using numerical models to simulate source term releases from vitrified and cement-based waste forms. Source term releases were subsequently transported through an air pathway and a groundwater pathway to a point of compliance 100 meters from the disposed of waste. In addition, an analysis evaluating the consequences of an inadvertent intrusion into the waste was also performed. In accordance with US DoE's technical standard for performance assessments, deterministic and probabilistic analyses were performed to demonstrate that the Hanford Integrated Disposal Facility, together with the engineered features of the waste forms and natural features of the selected site, provides a reasonable expectation that the low-level waste and mixed low-level waste disposal will meet the radiological performance objectives established by US DOE. Post-closure doses to future members of the public were evaluated after the time of closure and included a 100-year period of institutional controls that limited exposure immediately following closure of the facility. Calculated doses to a future member of the public from the air pathway were below the 100 μSv/yr dose performance objective. Calculated doses to a future member of the public from all exposure pathways, including the air pathway and the groundwater pathway, were below the 250 μSv/yr dose performance objective. Doses following an inadvertent intrusion into the facility after the institutional control period were below the 5 mSv acute dose performance measure and the 1 mSv/yr chronic dose performance measure. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Possibilities of soil washing for decontamination at Belgoprocess

Contaminated soils form a large part of the nuclear waste arising from decommissioning activities. The storage and disposal of these large volumes of waste is costly. For this reason techniques which can decontaminate this waste stream to free release levels are economically very interesting. A feasibility study of the possibilities of soil washing to decontaminate such soils was ordered by NIRAS/ONDRAF and performed at Belgoprocess in collaboration with DEC. Initial contamination level and particle size distributions of contaminated soils from three different sources were determined. The main isotopes detected with gamma spectrometry contained in the waste were {sup 241}Am, {sup 137}Cs, {sup 226}Ra, {sup 60}Co and {sup 232}Th. The particle size distribution revealed that more than half of the mass of the quartz sand that makes up the soil has a particle size between 125 and 212 μm. This fraction is less contaminated than the fractions containing smaller particles. However, separation of the fines fraction (< 125 μm) was not enough to achieve the free release limit. Soil attrition was tested as an extra decontamination step for the sand fraction. The removal efficiencies for the different radionuclides contained in the soil were measured. The process conditions were optimized to achieve maximum removal and a treatment method for the secondary waste coming from this process step was determined. The soil washing process was not only performed with water but also with nitric acid to assess the possibilities of a combination of a mechanical and a chemical decontamination process. Reduction efficiencies of 60-80% for the most relevant radionuclides were recorded. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗