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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 595 records · Page 33
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

High Waste Loading Glass Formulation Development for High-Sr HLW

This report describes work performed to develop and test new glass and feed formulations in order to increase waste loading and glass melting rates for Hanford HLW with high concentrations of strontium. Testing was designed to identify glass and melter feed formulations that optimize waste loading and waste processing rate while meeting all processing and product quality requirements.

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

Critical Need for Plutonium and Uranium Isotopic Standards with Lower Uncertainties

Certified reference materials (CRMs) traceable to national and international safeguards database are a critical prerequisite for ensuring that nuclear measurement systems are free of systematic biases. CRMs are used to validate measurement processes associated with nuclear analytical laboratories. Diverse areas related to nuclear safeguards are impacted by the quality of the CRM standards available to analytical laboratories. These include: nuclear forensics, radio-chronometry, national and international safeguards, stockpile stewardship, nuclear weapons infrastructure and nonproliferation, fuel fabrication, waste processing, radiation protection, and environmental monitoring. For the past three decades the nuclear community is confronted with the strange situation that improvements in measurement data quality resulting from the improved accuracy and precision achievable with modern multi-collector mass spectrometers could not be fully exploited due to large uncertainties associated with CRMs available from New Brunswick Laboratory (NBL) that are used for instrument calibration and measurement control. Similar conditions prevail for both plutonium and uranium isotopic standards and for impurity element standards in uranium matrices. Herein, the current status of U and Pu isotopic standards available from NBL is reviewed. Critical areas requiring improvement in the quality of the nuclear standards to enable the U. S. and international safeguards community to utilize the full potential of modern multi-collector mass spectrometer instruments are highlighted.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2017

Utilizing the Shift Technical Engineer Pipeline to Develop the Next Generation of Engineers - 17354

As the liquid waste contractor for the U.S. Department of Energy's (DOE) Savannah River Site (SRS), Savannah River Remediation (SRR) is responsible for safely treating and dispositioning approximately 35 million gallons of liquid waste stored at SRS. To achieve this mission, SRR manages and operates several complex nuclear facilities. These nuclear facilities include the H and F Tank Farms, the Defense Waste Processing Facility (DWPF), and the Saltstone Processing and Disposal Facility. To support the operations of these complex nuclear facilities, a Shift Technical Engineer (STE) is stationed for every shift. The STE is the sole Engineering Organization representative on-shift and serves as a direct technical resource for operations, including safety basis interpretation, process monitoring and prompt troubleshooting support. The training and qualification for an STE is more intensive than for typical technical staff personnel at SRR, involving additional classes, under-instruction watches, demonstration of practical factors, and culminating in an intense oral board examination. Like many other contractors in the DOE complex, SRR faces the challenge of preparing a significant number of new generation engineers to take the place of an aging workforce. Beyond the traditional training to establish minimum technical qualifications, SRR has established a number of initiatives aimed at accelerating the process of preparing early career engineers to replace experienced senior engineers. These initiatives include an internal rotational program, technical exchange activities with companies at other DOE sites, and an STE pipeline. The STE role provides a unique mechanism for select, early career engineers to gain in-depth knowledge and significant responsibility normally not achieved until much later in their careers. Once qualification is complete, the engineers perform the STE duties, providing technical support for nuclear facility operations. In addition, creating an STE pipeline, where promising young engineers are continually being qualified to replace the on-shift STEs, provides the company with a source of experienced engineers with in-depth technical knowledge as well as valuable experience interfacing with operations and other facility support groups. This paper will discuss the STE position, the qualification process, and how SRR is using the STE qualification pipeline to benefit both the company and the individual engineer. (authors)

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

Sludge Batch 10 Flowsheet Testing with Non-radioactive Simulants

Fourteen Chemical Processing Cell (CPC) simulations were performed with nonradioactive sludge simulants at the Aiken County Technology Laboratory in Aiken, SC. Four of these experiments were performed with Tank 51 sludge simulant. The remaining ten were performed with Tank 40 sludge simulant. The purpose of these experiments was to elucidate the chemistry and characteristics of Sludge Batch (SB) 10 as anticipated in the Defense Waste Processing Facility (DWPF). Experiments were performed at acid stoichiometries between 76% and 138% of the Koopman Minimum Acid requirement (85% - 144% of the Hsu acid requirement) and at REDuction/OXidation (REDOX) targets between 0.1 and 0.3. Testing examined the impact of coupled operations and sludge-only operations during Sludge Receipt and Adjustment Tank (SRAT) and Slurry Mix Evaporator (SME) processing at both design basis and nominal boilup rates. This report shares conclusions made as a result of this testing.

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

Milestones for Selection, Characterization, and Analysis of the Performance of a Repository for Spent Nuclear Fuel and High-Level Radioactive Waste at Yucca Mountain.

This report presents a concise history in tabular form of events leading up to site identification in 1978, site selection in 1987, subsequent characterization, and ongoing analysis through 2009 of the performance of a repository for spent nuclear fuel and high - level radioactive waste at Yucca Mountain in southern Nevada. The tabulated events generally occurred in five periods: (1) commitment to mined geologic disposal and identification of sites; (2) site selection and analysis, based on regional geologic characterization through literature and analogous data; (3) feasibility analysis demonstrating calculation procedures and importance of system components, based on rough measures of performance using surface exploration, waste process knowledge, and general laboratory experiments; (4) suitability analysis demonstrating viability of disposal system, based on environment - specific laboratory experiments, in - situ experiments, and underground disposal system characterization; and (5) compliance analysis, based on completed site - specific characterization . The current sixth period beyond 2010 represents a new effort to set waste management policy in the United States. Because the relationship is important to understanding the evolution of the Yucca Mountain Project , the tabulation also shows the interaction between the policy realm and technical realm using four broad categories of events : (a) Regulatory requirements and related federal policy in laws and court decisions, (c) Presidential and agency directives, (c) technical milestones of implementing institutions, and (d) critiques of the Yucca Mountain Project and pertinent national and world events related to nuclear energy and radioactive waste. Preface The historical progression of technical milestones for the Yucca Mountain Project was originally developed for 10 journal articles in a special issue of Reliability Engineering System Safety on the performance assessment for the Yucca Mountain license application [1-10]. The listing of mile stones, a distinct feature of those articles, has been collected and tabulated here. A lthough a brief description is presented here (based on the summaries in the 10 journal articles), the emphasis remains on the tabulation because of its usefulness in providing a comprehensive but concise history of the Yucca Mountain Project. The tabulation presented here is more elaborate than originally presented in that many of the interactions that occurred between the technical realm and policy realm can be depicted in separate columns. The usefulness of the milestones table is due in part to L.A. Connolly, for editorial and reference support, and S.K. Best, Raytheon, and L. Mays, Sandia National Laboratories (SNL), for illustration support. Reviewers P.N. Swift, SNL, and K. Gupta, University of Oklahoma, helped improve the discussion. The historical perspective presented is that of the author and is not necessarily held by reviewers, Sandia National Laboratories , and the US Department of Energy. As a historic perspective, the author is reporting on the work of others; however, any interpretative error s of the documentation are those of the author alone. The characterization and modeling of the Yucca Mountain disposal system required numerous participants with expertise in many areas of science and technology, as evident from the extensive reference list. Their diligent efforts are generally acknowledged here and through the many references to their impressive work, but the 10 journal articles acknowledge by name many of the numerous participants that contributed to the Yucca Mountain Project .

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Characterization of DWPF Recycle Condensate Tank Materials (Revision 1)

A Defense Waste Processing Facility (DWPF) Recycle Condensate Tank (RCT) sample was delivered to the Savannah River National Laboratory (SRNL) for characterization with particular interest in the concentration of I-129, U-233, U-235, total U, and total Pu. Since a portion of Salt Batch 8 will contain DWPF recycle materials, the concentration of I-129 is important to understand for salt batch planning purposes. The chemical and physical characterizations are also needed as input to the interpretation of future work aimed at determining the propensity of the RCT material to foam, and methods to remediate any foaming potential. According to DWPF the Tank Farm 2H evaporator has experienced foaming while processing DWPF recycle materials. The characterization work on the RCT samples has been completed and is reported here.

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

Preliminary Analysis of Species Partitioning in the DWPF Melter

The work described in this report is preliminary in nature since its goal was to demonstrate the feasibility of estimating the off-gas entrainment rates from the Defense Waste Processing Facility (DWPF) melter based on a simple mass balance using measured feed and glass pour stream compositions and time-averaged melter operating data over the duration of one canister-filling cycle. The only case considered in this study involved the SB6 pour stream sample taken while Canister #3472 was being filled over a 20-hour period on 12/20/2010, approximately three months after the bubblers were installed. The analytical results for that pour stream sample provided the necessary glass composition data for the mass balance calculations. To estimate the “matching” feed composition, which is not necessarily the same as that of the Melter Feed Tank (MFT) batch being fed at the time of pour stream sampling, a mixing model was developed involving three preceding MFT batches as well as the one being fed at that time based on the assumption of perfect mixing in the glass pool but with an induction period to account for the process delays involved in the calcination/fusion step in the cold cap and the melter turnover.

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

Effect of Na Concentration on Cs Distribution with Crystalline Silicotitanate in Tank Waste Simulants

Crystalline silicotitanate (CST) ion exchange media is currently utilized in the tank side cesium removal (TSCR) system on the Hanford site to remove Cs-137 from the tank waste supernate. As the main dose contributor to the liquid supernate in Hanford tank waste, it is necessary to remove the Cs-137 to expedite low activity waste processing and immobilization. Ongoing batch contact experiments with CST have been done to tease out the impact of group II metals, anions, potassium, and temperature on Cs removal to better understand bounding conditions for operations as well as aid in development of a wholesome isotherm model to predict Cs behavior in the tank waste. This paper discusses the impact of Na concentration on Cs exchange and provides insight into the challenges associated with activity coefficient estimations for modeling Cs distribution in tank waste matrices.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2019

Extraction of Mercury Utilizing 3D Printed Media - Mercury Extraction Coalescer (MEC) (Revision 1)

The presence of mercury in environmental or industrial settings presents a number of issues from bio-accumulation to production of toxic by-products. The primary objective of this project was the development of a 3D printed component designed to act as a coalescing filter for the removal or segregation of elemental mercury from aqueous streams. The use of 3D printing allows for the design and integration of surface microstructures and internal geometries that could be difficult if not impossible to achieve through traditional manufacturing techniques. The initial phase of testing focused on the integration of 3D printed components into both environmental remediation testing (university subcontracted testing) and simulations of high-level nuclear waste processing (SRNL). Structures ranged from simple torturous paths that provided regions where the elemental mercury could settle out of the recirculation process to micro-structured frits that have increased surface area for aid in droplet coalescence and surface adsorption.

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

Literature Review on the Impact of Thermolytic Hydrogen Generation Rate from the Next Generation Solvent

A literature review was undertaken to assess the thermolytic hydrogen generation potential of the Next Generation Solvent (NGS) system, which is planned to replace the Original Caustic-Side Solvent Extraction (CSSX) Solvent at the Salt Waste Processing Facility. NGS uses a new extractant and suppressor and the same modifier and organic carrier solvent as the Original CSSX Solvent at different concentrations.

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

Retained Hydrogen Literature Study for DWPF

During Defense Waste Processing Facility (DWPF) operation in the Slurry Receipt and Adjustment Tank (SRAT) and the Slurry Mix Evaporator (SME), hydrogen is primarily produced by radiolysis and catalytic generation. The hydrogen is released from the sludge by agitation and removed from the vapor space by purging with air. Upon loss of agitation, the hydrogen begins accumulating in the sludge, and releases rapidly upon the resumption of agitation. If enough hydrogen accumulated in the sludge, and released rapidly upon the start of agitation, the hydrogen concentration in the vapor space could exceed the lower flammability limit (4 vol %). Savannah River Remediation (SRR) currently manages the risk for retained hydrogen in the DWPF vessels by implementing a Retained Hydrogen Program. The program primarily relies on the timely operation of vessel agitators to ensure that the retained hydrogen eruptions that could lead to flammable vapor spaces do not occur. The current program relies on conservative assumptions concerning gas release and retention (e.g., instantaneous release, conservative hydrogen bubble fraction, no release during quiescent periods) to develop allowable vessel quiescent times (Q-times). SRR would like to relax the conservative assumptions to the Retained Hydrogen Program and allow the Retained Hydrogen Program to extend past Sludge Batch 9. Relaxing the assumptions could potentially allow longer Q-times during agitator stoppages. SRR requested Savannah River National Laboratory (SRNL) personnel to perform a literature study to extract industry data pertaining to retained hydrogen and apply it to the DWPF process. The authors conducted a review of the technical literature, including work from the Savannah River Site (SRS), SRNL, the Hanford site, Pacific Northwest National Laboratory (PNNL), and the Sellafield site to identify literature describing the retention and release of gases with non-Newtonian slurries. The study was undertaken with the knowledge that SRS wastes are unique which makes using non-SRS data challenging. The literature search focused on the following topics: Mixing or blend time in Newtonian and non-Newtonian fluids, Quiescent release of gas from an unagitated non-Newtonian fluid, Release of gas from an agitated non-Newtonian fluid, and Changes in fluid rheology as a function of settling time.

08 HYDROGEN↗
DOE OSTI2024

Effect of 60 Co Irradiation on Boehmite Dissolution in Caustic Solutions

Here, in this work, we examine how radiation impacts the dissolution behavior of boehmite by subjecting dry nanoparticles of different sizes to 60 Co γ radiation and subsequently analyzing their dissolution behavior in caustic solutions as a function of temperature. The measured kinetics show that irradiation with an amount 228.24 Mrad significantly slows the dissolution rate, particularly for smaller sizes at lower temperatures. Specifically, the temperature-dependent dissolution rates of irradiated 20 nm boehmite versus pristine material in 3 M NaOH solutions were several times lower (e.g., rate constant of 0.026 vs 0.075 h –1 at 60 °C), with an apparent activation energy 40 kJ mol –1 higher. Although various imaging techniques and X-ray diffraction measurements consistently revealed no obvious differences between pristine and irradiated samples, after irradiation significant binding energy shifts were detected in the X-ray photoelectron Spectroscopy peaks of Al 2p and O 1s, and a change in their relative intensities indicated a lower O/Al ratio. This suggests that γ-irradiation may stabilize boehmite particle surfaces by driving their chemistry and structure toward more stable aluminum oxide forms. This finding may help explain slower dissolution rates of boehmite in nuclear waste and may be useful for the development of more robust predictive models and effective strategies for waste processing.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2023

Technical Evaluation of Accelerated Basin De-Inventory Material Addition to Sludge Batch 11 (Rev. 1)

The Accelerated Basin De-inventory (ABD) program involves discarding spent nuclear fuel that is currently stored in L-Basin to the Defense Waste Processing Facility (DWPF) for vitrification. The first ABD discards will occur during the preparation of Sludge Batch (SB) 11. Savannah River Mission Completion has requested that the Savannah River National Laboratory assess the technical gaps related to the increased gadolinium poisoning requirement and the impacts of performing the Low Temperature Aluminum Dissolution (LTAD) process in Tank 51 with H-Canyon discards present. The following summarizes the evaluation of the impacts of increasing the quantity of gadolinium (and related topics) from what was previously evaluated in the SRNL studies of gadolinium-poisoned ABD material solubility, the overall ABD flowsheet review, and increasing the fissile mass loading in glass: 1) Based on literature surveys, there is no indication that organic interactions with gadolinium will be significant at the high pH (typically >13) conditions of the Concentration, Storage, and Transfer Facilities. Any interactions of gadolinium with organics in DWPF are not expected to adversely impact DWPF or downstream facilities. Thus, there is little-to-no residual risk from organic interactions with gadolinium [Gap closed]; 2) Adding depleted uranium to ABD material, targeting 235 U enrichment of 4.90% within each transfer window, will mitigate potential impacts from an increase in soluble 235 U enrichment during sludge washing and LTAD. The plan to take advantage of previous transfers and allow 235 U enrichment of >5% during the final transfer window carries a risk that Tank 51 supernate will have a 235 U enrichment of >5%, which should be evaluated for acceptance; 3) Increasing the gadolinium mass ratio to 3.0:1 Gd: 235 U(eq SLU ) should lead to the same or higher partitioning of gadolinium into the solid phase within the DWPF Chemical Process Cell, resulting in both liquid and solid phases with expected partitioning of Gd consistent with the prior solubility study [Gap closed for SB11]; 4) There are no expected impacts on DWPF melt temperature and melter operations due to the minimal ~0.2 weight percent (wt%) increase in Gd concentration relative to previous sludge batches [Gap closed for SB11]; 5) As observed previously, Gd is expected to enter the off-gas system via physical entrainment, but at a slightly higher concentration than what was observed for SB9 melter off-gas pluggage deposits (0.07 wt%) [Gap closed for SB11] ; 6) There are no expected impacts on DWPF recycle or the Recycle Collection Tank glycolate destruction process. [Gap closed for SB11]; 7) Gd is projected to be a trace component in the SB11 glass (<0.5 wt%) and can be ignored for process control. Trace components do not significantly impact glass durability, thus the conclusions of the previous Product Consistency Test evaluation at a fissile mass loading of 2,500 g fissile/m3 glass still applies to SB11. The ~0.1 wt% increase in Gd2O3 concentration relative to the previous study will not impact the predictability of SB11 glass with the DWPF Product Composition Control System (PCCS) models for durability or the acceptability of glass according to the Waste Acceptance Product Specifications (WAPS) criterion for product consistency [Gap closed for SB11]; 8) No additional Toxicity Characteristic Leaching Procedure testing is necessary for SB11 and the hazardous waste specification of the SB11 DWPF waste form is unchanged after the addition of the ABD stream [Gap closed for SB11]. The following summarizes the evaluation of the impacts of adding two-thirds of the ABD material to Tank 51 prior to LTAD: 1) The addition of two-thirds of the ABD increases overall aluminum mass from 1.39×10 4 kg to 1.64×10 4 kg (15.5% ABD Al). The form of the insoluble portion of the Al resulting from ABD addition should be the more readily dissolved Al(OH) 3 and amorphous forms. The portion of the ABD aluminum that is processed by LTAD is expected to be completely soluble, thus requiring that less of the boehmite in the sludge be dissolved to reach the same Al target in the SB. [Gap closed for SB11]; The expected LTAD impact on other components, as related primarily to the components in ABD, are discussed. Gd is expected to remain insoluble during LTAD and not impact the solubility of other components. [Gap closed for SB11]; The addition of two-thirds of the ABD increases overall projected SB11 uranium mass from 4,740 kg to 13,100 kg (63% ABD U) and the projected plutonium mass from 86.0 kg to 89.5 kg (3.9% ABD Pu). The addition of all of the ABD increases overall projected SB11 uranium mass from 4,740 kg to 16,100 kg (70% ABD U) and the projected plutonium mass from 86.0 kg to 90.4 kg (5.3% ABD Pu). The 235 U enrichment will be ≤5%. The fissile uranium will be adequately poisoned by Gd and the fissile Pu will be adequately poisoned by Fe from the sludge. [Gap closed for SB11]; There is a low risk that ABD addition will impact the rheology or pumpability of the slurry. There is a low but higher risk of ABD addition prior to LTAD impacting the settling rate; Based on the evaluation of adding two-thirds of the ABD material and all of the ABD material prior to the LTAD process, there is no volume or mass limit that would need to be imposed on ABD additions prior to LTAD. [Gap closed for SB11]. Revision 1 of this report addresses a variation on the ABD additions and LTAD strategy where sodium hydroxide additions for LTAD may be performed intermittently or concurrently with an ABD addition window. The proposed change does not alter the conclusions of this evaluation.

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

Sludge Washing And Demonstration Of The DWPF Flowsheet In The SRNL Shielded Cells For Sludge Batch 8 Qualification

The current Waste Solidification Engineering (WSE) practice is to prepare sludge batches in Tank 51 by transferring sludge from other tanks to Tank 51. Tank 51 sludge is washed and transferred to Tank 40, the current Defense Waste Processing Facility (DWPF) feed tank. Prior to transfer of Tank 51 to Tank 40, the Savannah River National Laboratory (SRNL) typically simulates the Tank Farm and DWPF processes using a Tank 51 sample (referred to as the qualification sample). WSE requested the SRNL to perform characterization on a Sludge Batch 8 (SB8) sample and demonstrate the DWPF flowsheet in the SRNL shielded cells for SB8 as the final qualification process required prior to SB8 transfer from Tank 51 to Tank 40. A 3-L sample from Tank 51 (the SB8 qualification sample; Tank Farm sample HTF-51-12-80) was received by SRNL on September 20, 2012. The as-received sample was characterized prior to being washed. The washed material was further characterized and used as the material for the DWPF process simulation including a Sludge Receipt and Adjustment Tank (SRAT) cycle, a Slurry Mix Evaporator (SME) cycle, and glass fabrication and chemical durability measurements.

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

Nitric-glycolic flowsheet testing for maximum hydrogen generation rate

The Defense Waste Processing Facility (DWPF) at the Savannah River Site is developing for implementation a flowsheet with a new reductant to replace formic acid. Glycolic acid has been tested over the past several years and found to effectively replace the function of formic acid in the DWPF chemical process. The nitric-glycolic flowsheet reduces mercury, significantly lowers the chemical generation of hydrogen and ammonia, allows purge reduction in the Sludge Receipt and Adjustment Tank (SRAT), stabilizes the pH and chemistry in the SRAT and the Slurry Mix Evaporator (SME), allows for effective adjustment of the SRAT/SME rheology, and is favorable with respect to melter flammability. The objective of this work was to perform DWPF Chemical Process Cell (CPC) testing at conditions that would bound the catalytic hydrogen production for the nitric-glycolic flowsheet.

08 HYDROGEN↗
NASA NTRS2009

Environmental Program

NASA's White Sands Test Facility has six core environmental compliance capabilities: remote hazardous testing of reactive, explosive and toxic materials and fluids; hypergolic fluids materials and systems testing; oxygen materials and system testing; hypervelocity impact testing; flight hardware processing; and, propulsion testing. The facility's permit status and challenges are reviewed. Historic operations and practices dating from the 1960s through the early 1980s resulted in contamination of the facility's groundwater. An environmental restoration effort has been employed to protect public health and the health of the workforce. The restoration seeks to properly handle hazardous materials and waste processes; determine the nature and extent of the contamination; stop the migration of contaminated groundwater; stabilize the plume front which has been assessed as the greatest risk to public health; and, clean-up the environment to restore it to preexisting conditions. The Plume Front Treatment System is operational and seeks to stop the westward movement of the plume to protect drinking water and irrigation well. Specifically, the treatment system will extract contaminated water from the aquifer, remove chemical using the best available technology, and return (inject) the treated water back to the aquifer. The Mid-Plume Interception Treatment System also seeks to stop the migration of containment, as well as to evaluate new technologies to accelerate cleanup, such as bioremediation.

Fischer, Holger↗
DOE OSTI2020

Effect of Cr(III) Adsorption on the Dissolution of Boehmite Nanoparticles in Caustic Solution

Incorporation of relatively minor impurity metals onto metal (oxy)hydroxides can strongly impact solubility. In complex highly alkaline multicomponent tank waste such as that at the Hanford Nuclear Reservation, tests indicate that the surface area-normalized dissolution rate of boehmite ($γ$-AlOOH) nanomaterials is at least an order of magnitude lower than that predicted for the pure phase. In this work we examine the dissolution kinetics of boehmite coated by adsorbed Cr(III), which adheres at saturation coverages as sparse chemisorbed monolayer clusters. Using 40 nm boehmite nanoplates as a model system, temperature-dependent dissolution rates of pure versus Cr(III)-adsorbed boehmite showed that the initial rate for the latter is consistently several times lower, with an apparent activation energy 16 kJ·mol -1 higher. Although the surface coverage is only around 50%, solution analytics coupled to multimethod solids characterization reveals a phyicochemical armoring effect by adsorbed Cr(III) that substantially reduces the number of dissolution-active sites on particle surfaces. Such findings could help improve kinetics models of boehmite and/or metal ion adsorbed boehmite nanomaterials, ultimately providing a stronger foundation for the development of more robust complex radioactive liquid waste processing strategies.

54 ENVIRONMENTAL SCIENCES↗
NASA NTRS1991

The Space Exploration Initiative: a challenge to advanced life support technologies: keynote presentation

President Bush has enunciated an unparalleled, open-ended commitment to human exploration of space called the Space Exploration Initiative (SEI). At the heart of the SEI is permanent human presence beyond Earth orbit, which implies a new emphasis on life science research and life support system technology. Proposed bioregenerative systems for planetary surface bases will require carefully designed waste processing elements whose development will lead to streamlined and efficient and efficient systems for applications on Earth.

Ecological Systems, Closed↗