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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 379 records · Page 21
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Inorganic Ba-Sn Nanocomposite Materials for Sulfate Sequestration from Complex Aqueous Solutions

Selective sequestration of sulfate (SO42-) in the form of barite (BaSO4) from alkaline solutions of high ionic strength containing carbonate at high concentrations resembling legacy radioactive waste generated during plutonium cold war era production is problematic due to the preferential formation of BaCO3. Incorporation of sulfate into the insoluble and thermally stable BaSO4 phase can potentially benefit radioactive waste processing in reducing operational challenges and suppressing volatilization of other waste components such as technetium-99. To enhance selectivity of SO42- sequestration, a series of Ba(II)-Sn(IV) materials was prepared using simple hydrothermal synthesis from different Sn(II) and Sn(IV) precursors. Structural characterization indicated that all obtained products predominantly contained BaSn(OH)6 and Ba2SnO2(OH)4(H2O)10 crystalline phases which were disrupted upon exposure to SO42- due to formation of BaSO4. Performance of the Ba(II)-Sn(IV) materials was tested using complex alkaline solutions simulating radioactive waste containing 0.094 M SO42- and 0.5 M CO32- among other constituents. About 54 – 66% of SO42- was converted to BaSO4 when a quantity of Ba(II)-Sn(IV) material containing approximately a stoichiometric amount of Ba2+ relative to SO42- was used. In comparison, previous studies indicate negligible BaSO4 formation under similar conditions when a simple Ba2+ salt is used. This improvement is attributed to the selective replacement of the Ba(II) stannate by SO42-. Thermal stability of the sulfate-loaded product material up to 1100 °C was demonstrated. Obtained materials offer a convenient and economical option for the selective sequestration or removal of SO42- from carbonate containing solutions.

Johnson, Isaac↗
DOE OSTI2025

Conceptual Model Testing Related to SDU 6 Drainwell Observations

From its inception in the early 1950s through the end of the Cold War in the early 1990s, the Savannah River Site (SRS) produced nuclear materials for national defense in five reactors. Additionally, irradiated reactor fuel and target tubes were dissolved in nitric acid to recover plutonium and uranium using the PUREX (Plutonium Uranium Reduction EXtraction) process. Liquid waste from these chemical separations processes was then stored onsite in 51 underground tanks. Eight waste storage tanks have been operationally closed (i.e. cleaned and grouted) and the remaining tanks hold a mixture of liquids, insoluble solids, and precipitated salts (SRMC-LWP-2022-00001), the latter generated by evaporating water from the liquid waste. Waste is currently being retrieved from tanks and separated into 1) high-radioactivity, low-volume, and 2) low-radioactivity, high-volume components, principally through the Salt Waste Processing Facility (SWPF) (SRMC-LWP-2023-00001). The former waste stream is vitrified in the Defense Waste Processing Facility (DWPF), stored onsite, and destined for offsite disposal in a deep geologic repository. The latter stream is mixed with dry cementitious materials in the Saltstone Production Facility (SPF) and the wet slurry placed in onsite Saltstone Disposal Units (SDUs) within the Saltstone Disposal Facility (SDF), where it hardens into a cement waste form termed saltstone. A low-infiltration surface cover system will be placed over the SDF at closure, where SDUs will then be in the subsurface post-closure (SRR-CWDA-2019-00001).

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DOE OSTI2022

Multi-process tooling for discontinuous carbon and hybrid glass fiber thermoplastics

Expensive tooling often constraints the use of composites in the design and development of automotive parts. While there is significant confidence and knowledge in sheet and bulk metals, composite processes are less understood in mass production environment. The processes used to produce composites and resulting properties are influenced by fiber length attrition, resin to fiber ratio, process waste etc. Tool designs are determined very early in the engineering process. It is cost prohibitive to build additional tools, in the event it becomes obvious a better processing method and material would be beneficial, the original decision is not easily changed. In the present work we recognize the bottleneck of tooling costs and provide an approach of multi-process tooling. The innovation of this work is the design and demonstration of a single tool for different processes namely injection, injection-compression and extrusion-compression. The materials used in this study were long and short fiber thermoplastics (LFTs and SFTs). The resulting structure-property relationships have been reported for the materials and processing methods with a battery tray (BT) tool.

Vaidya, Uday↗
DOE OSTI2016

Results For The Third Quarter Calendar Year 2016 Tank 50H Salt Solution Sample

In this memorandum, the chemical and radionuclide contaminant results from the Third Quarter Calendar Year 2016 (CY16) sample of Tank 50H salt solution are presented in tabulated form. The Third Quarter CY16 Tank 50H samples (a 200 mL sample obtained 6” below the surface (HTF-5-16-63) and a 1 L sample obtained 66” from the tank bottom (HTF-50-16-64)) were obtained on July 14, 2016 and received at Savannah River National Laboratory (SRNL) on the same day. Prior to obtaining the samples from Tank 50H, a single pump was run at least 4.4 hours, and the samples were pulled immediately after pump shut down. The information from this characterization will be used by Defense Waste Processing Facility (DWPF) & Saltstone Facility Engineering for the transfer of aqueous waste from Tank 50H to the Saltstone Production Facility, where the waste will be treated and disposed of in the Saltstone Disposal Facility. This memorandum compares results, where applicable, to Saltstone Waste Acceptance Criteria (WAC) limits and targets. Data pertaining to the regulatory limits for Resource Conservation and Recovery Act (RCRA) metals will be documented at a later time per the Task Technical and Quality Assurance Plan (TTQAP) for the Tank 50H saltstone task. The chemical and radionuclide contaminant results from the characterization of the Third Quarter CY16 sampling of Tank 50H were requested by Savannah River Remediation (SRR) personnel and details of the testing are presented in the SRNL TTQAP.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Solids Erosion Patterns Developed by Pulse Jet Mixers

Millions of gallons of radioactive waste are stored in underground storage tanks at the Hanford Site in Washington State. This waste will be vitrified at the Waste Treatment and Immobilization Plant that is under construction. Vessels in the pretreatment portion of the plant are being configured for processing waste slurries with challenging physical and rheological properties that range from Newtonian slurries to non-Newtonian sludge. Pulse jet mixing technology has been selected for mobilizing and mixing this waste. In the pulse jet mixing process, slurry is expelled from pulse tube nozzles directed towards the vessel floor. The expelled fluid forms a radial jet that erodes the settled layer of solids. The pulse tubes are configured in a ring or multiple rings and operate concurrently. The expelled fluid and mobilized solids traverse toward the center of the tank. At the tank center the jets from pulse tubes in the ring collide and lift solids upward in a central plume. At the end of the pulse, when the desired fluid volume is expelled from the pulse tube, the applied pressure switches to suction and the pulse tube is refilled. This cycle is used to mobilize and mix the tank contents. An initial step of the process is the erosion of solids from the vessel floor by the radial jets that form on the vessel flow beneath each pulse tube. Experiments have been conducted using simulants to evaluate the ability of the pulse jet mixing system radial jets to combine to develop the central upwell and lift solids into the vessel. These experiments have been conducted at three scales using a range of granular simulants over a range of concentrations. The vessels have elliptical, spherical, or flanged and dished bottoms. Process parameters evaluated include the velocity of fluid expelled from the pulse tube, the duration of the pulse and the duty cycle, the ratio of pulse duration to cycle time. Videos taken from beneath the vessel show the growth of the cleared area from each pulse tube as a function of time. All solids are lifted from the vessel bottom when the system is operating at the critical suspension velocity. The focus of this paper is to compare and contrast erosion patterns developed from different simulants and pulse tube configurations. The cases are evaluated to determine how changes in process parameters affects the PJM ability to mobilize solids from the vessel floor.

erosion↗
DOE OSTI2019

Analytical Results from Salt Batch 10 Routine DSSHT and SEHT Monthly Samples

Strip Effluent (SE) Hold Tank (SEHT) and Decontaminated Salt Solution (DSS) Hold Tank (DSSHT) samples from several “microbatches” of the Integrated Salt Disposition Project (ISDP) “macrobatch” Salt Batch 10 have been analyzed for 238 Pu, 90 Sr, 137 Cs, cations (via Inductively Coupled Plasma Emission Spectroscopy - ICPES), and in some cases anions (via Ion Chromatography Anions - IC-A). The analytical results from the current microbatch samples are similar to those from previous macrobatches. In the Modular Caustic-Side Solvent Extraction Unit (MCU), the Cs removal continues to exceed process requirements. The bulk chemistry of the DSSHT and SEHT samples do not show any signs of unusual behavior except for slightly elevated amounts of sodium in the SE samples. This summary represents the final Salt Batch processed through MCU prior to commencement of process and transfer line tie-ins to support the startup of the Salt Waste Processing Facility (SWPF).

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

Evaluation of New Additions to OLI Software in Predicting Mercuric and Mercurous Species in Liquid Waste Operations

Speciation of mercury during the pretreatment steps of tank waste processing is critical to successful mercury removal prior to vitrification during Liquid Waste Operations (LWO) at SRS. OLI software has been used to predict mercury speciation and activity throughout LWO. The OLI software operates based on a thermodynamic framework called the Mixed Solvent Electrolyte (MSE) framework. The MSE framework allows prediction in theoretically infinitely dilute to concentrated mixtures (e.g., purely solute solutions). Before modification to the MSE framework databanks, certain critical mercury species were missing in the MSE databank, and some thermodynamic data needed to be updated for the OLI software to accurately predict mercury chemical species in SRS waste tanks. To better reflect streams across LWO, new mercury species were integrated into the MSE database. To evaluate the changes to the OLI MSE framework per the Technical Task Request (TTR) and the Task Technical and Quality Assurance Plan (TTQAP), waste stream compositions from Tanks 38, 43, and Tank 50 decontaminated salt solution (DSS) were used as model inputs. Models were developed and executed using both the old and new databases. Compositional analyses from caustic Tank 50 DSS and caustic Tanks 38 and 43 were used as the input streams. These streams represent the most comprehensive chemical data sets where both mercury and tank constituents were measured together. Results for Tank 50 DSS predict HgO as the predominant species in both databases. Both methyl and dimethyl Hg species are present when the new database is ‘on’ and are not predicted with the new database turned ‘off’. The new database predicts a greater amount of HgO and a greater fraction of it in the solid phase. Pourbaix diagrams (potential vs. pH) generated for each Tank 50 DSS were identical regardless of which database was used. Elemental Hg and HgO were predicted in the water stable region under basic conditions. Tanks 38 and 43 follow similar trends as the Tank 50 DSS models. Unlike Tanks 38 and 50 DSS, the Tank 43 Pourbaix plot shows a region of stability for an aqueous HgOHCO3 - species between approximately pH 7-11. In all streams, when MeHg+ is included in the inputs, the new database predicts aqueous MeHgOH as the dominant species. If elemental or dimethyl mercury is in the waste stream, the new database model predicts they are unchanged and remain in those states and quantities. Additionally, the total mercury values are reported for both the measured input data and the OLI output data for all considered tanks. The summary indicates that the percentage error between the measured and calculated values is less than 1% in all cases The reconciliations and generation of the Pourbaix diagrams for Tank 50 DSS took approximately ten times longer with the new database ‘on’. In addition, over the course of that time, models with the new database ‘on’ were more likely to crash or display an error. Some modest performance improvements were noted when modeling with an i7 processor versus an i5. An example error is found in Appendix A. Furthermore, Appendix B provides V&V for two chemical systems analyzed with the OLI software, results were satisfactory. It is recommended to utilize the new databases (i.e., HCO.ddb and SR-Hg.ddb) in future Savannah River Mission Completion applications of OLI to represent pseudo steady-state. Furthermore, the integration and utilization of the new databases (i.e., HCO.ddb and SR-Hg.ddb) in modeling applications (e.g., Aspen) is also recommended.

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

Minimizing Transuranic Waste Generated in Hot Cells at Oak Ridge National Laboratory Through Use of a Non-Destructive CO2 Pellet Cleaning System

Wastes generated in shielded hot cells containing materials contaminated with transuranic (TRU) waste present unique problems in waste disposal at Department of Energy (DOE) sites. Reduction of the TRU waste footprint is desirable from a disposal cost perspective due to constraints with RH waste processing and a lack of space availability at the WIPP. Segregation of waste through a pelletized CO2 system provides a viable method this goal.

Ford, Mark↗
NASA NTRS2005

Impact Crater Deposits in the Martian Highlands

The martian highlands of Noachis Terra (20-30 deg S, 20-50 deg E), Tyrrhena Terra (0-30 deg S, 50- 100 deg E) and Terra Cimmeria (0-60 deg S, 120-170 deg E) preserve long and complex histories of degradation, but the relative effects of such factors as fluvial, eolian, and mass wasting processes have not been well constrained. The effects of this degradation are best observed on large (D greater than 10 km) impact craters that characterize the ancient highlands. Some craters exhibit distinct interior deposits, but precise origins of these deposits are enigmatic; infilling may occur by sedimentary (e.g., fluvial, lacustrine, eolian), mass wasting and (or) volcanic processes.

Mest, S. C.↗
DOE OSTI2022

Analysis of DWPF Sludge Batch 6 (Macrobatch 7): Pour Stream Glass Samples

The Defense Waste Processing Facility (DWPF) began processing Sludge Batch 6 (SB6), also referred to as Macrobatch 7 (MB7), in June 2010. SB6 is a blend of the heel of Tank 40 from Sludge Batch 5 (SB5), H-Canyon Np transfers and SB6 that was transferred to Tank 40 from Tank 51.1 SB6 was processed using Frit 418. Sludge is received into the DWPF Chemical Processing Cell (CPC) and is processed through the Sludge Receipt and Adjustment Tank (SRAT) and Slurry Mix Evaporator Tank (SME). The treated sludge slurry is then transferred to the Melter Feed Tank (MFT) and fed to the melter. During processing of each sludge batch, the DWPF is required to take at least one glass sample to meet the objectives of the Glass Product Control Program (GPCP) and to complete the necessary Production Records so that the final glass product may be disposed of at a Federal Repository. The DWPF requested various analyses of radioactive glass samples obtained from the melter pour stream during processing of SB6 as well as reduction/oxidation (REDOX) analysis of MFT samples to determine the impact of Argon bubbling. Sample analysis followed the Task Technical and Quality Assurance Plan (TTQAP) and an Analytical Study Plan (ASP). Four Pour Stream (PS) glass samples and two MFT slurry samples were delivered to the Savannah River National Laboratory (SRNL) from the DWPF. Table 1-1 lists the sample information for each pour stream glass sample. SB6 PS3 (S03472) was selected as the official pour stream sample for SB6 and full analysis was requested. This report details the visual observations of the as-received SB6 PS No.3 glass sample as well as results for the chemical composition, Product Consistency Test (PCT), radionuclide content, noble metals, and glass density. REDOX results will be provided for all four pour stream samples and vitrified samples of MFT-558 and MFT-568A. Where appropriate, data from other pour stream samples will be provided.

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

Impacts of the Addition of Sodium Reactor Experiment (SRE) and DR-3 Fuel from H-Canyon to Tank 40 on Acid Stoichiometry and Hydrogen Production

H-Canyon plans to discard Sodium Reactor Experiment (SRE) material that is currently stored in Tanks 16.3 and 16.4. Savannah River Remediation (SRR) determined that a portion of the SRE material can be added to Sludge Batch 9 (SB9) in Tank 40 and the remainder will be added to Sludge Batch 10 (SB10) that is currently being prepared in Tank 51. A processing flowsheet for both the nitric-formic acid flowsheet and the nitric-glycolic acid flowsheet were previously developed for SB9 sludge-only and coupled operation with the Actinide Removal Process – Modular Caustic Side Solvent Extraction Unit (ARP-MCU) and the Salt Waste Processing Facility (SWPF). Due to the compositional change in Tank 40 after the addition of SRE material, it is necessary to determine the influence on the acid stoichiometry operating window and hydrogen production for SB9 processing. An assessment was completed to address the influence of an addition of SRE and DR-3 Fuel from H-Canyon on the chemistry of Tank 40, the defined acid stoichiometry window (with and without SWPF streams) for the nitric-formic and nitric-glycolic acid flowsheets, and the resulting influence of the hydrogen production. Hydrogen production during the SRAT/SME cycles consists of thermolytic, radiolytic, and catalytic hydrogen. For the nitric-formic acid flowsheet, catalytic hydrogen dominates the hydrogen production during DWPF operation mode for the SRAT and SME vessels. This report was written to summarize that assessment. No SRNL testing is needed prior to adding up to 7,100 additional gallons of SRE and DR-3 Fuel from H-Canyon to Tank 40. Also, the addition of SRE will not exceed the Shielded Cells reported values for catalytic hydrogen for either the nitric-formic acid flowsheet (Technical Safety Requirement (TSR) limit of 0.15 lb/h) or the nitric-glycolic acid flowsheet (TSR limit 0.024 lb/h). The addition of the ARP stream from SWPF and the use of the nitric-glycolic acid flowsheet are both expected to decrease hydrogen generation, leading to a larger safety margin in the CPC.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
NASA NTRS2008

Mars Rocket Propulsion System

A report discusses the methane and carbon monoxide/LOX (McLOx) rocket for ascent from Mars as well as other critical space propulsion tasks. The system offers a specific impulse over 370 s roughly 50 s higher than existing space-storable bio-propellants. Current Mars in-situ propellant production (ISPP) technologies produce impure methane and carbon monoxide in various combinations. While separation and purification of methane fuel is possible, it adds complexity to the propellant production process and discards an otherwise useful fuel product. The McLOx makes such complex and wasteful processes unnecessary by burning the methane/CO mixtures produced by the Mars ISPP systems without the need for further refinement. Despite the decrease in rocket-specific impulse caused by the CO admixture, the improvement offered by concomitant increased propellant density can provide a net improvement in stage performance. One advantage is the increase of the total amount of propellant produced, but with a decrease in mass and complexity of the required ISPP plant. Methane/CO fuel mixtures also may be produced by reprocessing the organic wastes of a Moon base or a space station, making McLOx engines key for a human Lunar initiative or the International Space Station (ISS) program. Because McLOx propellant components store at a common temperature, very lightweight and compact common bulkhead tanks can be employed, improving overall stage performance further.

Zubrin, Robert↗
DOE OSTI2018

Reporting the Radionuclide Inventory of the DWPF Product

In the Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS), the radionuclides contained in the liquid high-level waste currently stored at SRS are being converted into a durable borosilicate glass. This glass in its sealed stainless-steel canister, the DWPF product, eventually must be sent to a federal repository for final disposal. In order to ensure that the DWPF product is suitable for acceptance by the U. S. Department of Energy’s (DOE) Civilian Radioactive Waste Management System (CRWMS)*, the DOE Office of Environmental Management (DOE-EM) established the Waste Acceptance Product Specifications for Vitrified High-Level Waste Forms (WAPS). DOE-EM is tentatively performing surrogate CRWMS waste acceptance functions, as the office was eliminated in 2009, for review and concurrence of technical baseline documents and waste form non conformances. The specifications are divided into five sections, dealing with the waste form (borosilicate glass) itself, the canister, the canistered waste form, quality assurance, and documentation and other administrative requirements. The DWPF is required to document its compliance with the specifications in the Waste Form Compliance Plan (WCP), the Waste Form Qualification Report (WQR), the Production Records, and the Storage and Shipping Records. The WCP provides general information about the DWPF process and product, and a general description of the methods and programs by which the DWPF will demonstrate compliance with each specification in the WAPS. The WCP was vital to DWPF startup with respect to operations planning and also assurance to DOE that the canistered waste forms would be acceptable for disposal at any future High-Level Waste (HLW) geologic repository. The WQR is a compilation of the results of the testing and analysis programs identified in the WCP. The WQR details the method of compliance with each specification, and the testing performed to demonstrate the effectiveness of each method. The WQR is published in separate volumes which each correspond to one or more of the specifications. The WQR also contains a description of the DWPF’s Glass Product Control Program. Each volume contains Part and Item designators which indicate the corresponding item(s) in the WCP. The Production Records, and the Storage and Shipping Records document that each DWPF product has been produced and stored in compliance with the WAPS. This document, Volume 4 of the WQR, responds to WAPS Specifications 1.2.2, 1.6, and 3.14 which require that the DWPF report the radionuclide inventory of each canistered waste form, and of each waste type. This volume describes the systems the DWPF uses to determine and report the radionuclide inventory of the glass product, and provides evidence that these systems allow the DWPF to satisfy the WAPS specification. This volume of the WQR has been prepared to establish that methods are available to determine and report the radionuclide inventory, and that the DWPF is ready to collect the facility-specific data necessary to provide definitive estimates of the errors associated with the use of these methods.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
NASA NTRS1998

Status of Regenerative Life Support Research and Technology Program at NASA Ames Research Center

Future long duration manned space missions will require life support systems that minimize resupply requirements and ultimately approach self-sufficiency in space. This presentation will provide an overview of the Advanced Life Support program unclassified fundamental research and technology development activities being conducted at NASA Ames Research Center. Top level program goals and technical objectives, and the role of NASA-Ames within the Advanced Life Support program, will be reviewed. The presentation will focus on FY97 and FY98 research tasks that were directed at physicochemical processes with emphasis on system closure and self-sufficiency. Research areas include solid waste processing and resource recovery, water recycling, air regeneration, and regenerative system dynamics. Proposed future work and potential applications of this research to both terrestrial and space closed ecology experimentation in space will be addressed.

Kliss, Mark↗
DOE OSTI2025

Phosphate-Based Approaches for Dechlorination and Treatment of Salt Waste from Electrochemical Processing of Used Nuclear Fuel: A Perspective on Recent Work

Phosphate-based reagents are being considered by the U.S. Department of Energy (DOE) Office of Nuclear Energy to process halide salt-based nuclear wastes for stabilization prior to disposal. As evidenced by the Experimental Breeder Reactor-II (EBR-II) project, electrochemical processing (pyroprocessing) can be employed to recover uranium and other actinides for reintegration into the nuclear fuel cycle from metallic fuels. The resultant salt-based wastes generated from electrochemical processing of EBR-II fuel contains fission products within a LiCl–KCl eutectic salt that necessitate appropriate disposal. This paper provides an overview of recent efforts to support halide-based salt waste treatment for disposition, as well as a basis for comparison with other related efforts in salt waste treatment through salt partitioning initiatives. The U.S. DOE has selected a phosphate waste form reference material for further investigation and longer-term studies.

electrorefiner↗
DOE OSTI2018

Hierarchical Materials as Tailored Nuclear Waste Forms: A Perspective

This perspective focuses on the synthesis, characterization and modeling of three classes of hierarchical materials with potential for sequestering radionuclides: nanoparticles, porous frameworks and crystalline salt inclusion phases. The scientific impact of hierarchical structures and the development of the underlying crystal chemistry is discussed as laying the groundwork for the design, local structure control, and synthesis of new forms of matter with tailored properties. This requires development of the necessary scientific understanding of such complex structures through integrated synthesis, characterization, and modeling studies that can allow their purposeful creation and properties. The ultimate practical aim is to provide the means to create novel structure types that can simultaneously sequester multiple radionuclides. The result will lead to the creation of safe and efficient, long lasting waste forms for fission products and transuranic elements that are the products of nuclear materials processing waste streams. The generation of the scientific base for working toward that goal is presented.

hierarchal material, review article, nuclear waste↗
DOE OSTI2023

Using machine learning with optical profilometry for GaN wafer screening

Abstract To improve the manufacturing process of GaN wafers, inexpensive wafer screening techniques are required to both provide feedback to the manufacturing process and prevent fabrication on low quality or defective wafers, thus reducing costs resulting from wasted processing effort. Many of the wafer scale characterization techniques—including optical profilometry—produce difficult to interpret results, while models using classical programming techniques require laborious translation of the human-generated data interpretation methodology. Alternatively, machine learning techniques are effective at producing such models if sufficient data is available. For this research project, we fabricated over 6000 vertical PiN GaN diodes across 10 wafers. Using low resolution wafer scale optical profilometry data taken before fabrication, we successfully trained four different machine learning models. All models predict device pass and fail with 70–75% accuracy, and the wafer yield can be predicted within 15% error on the majority of wafers.

36 MATERIALS SCIENCE↗
NASA NTRS2010

Impact of Water Recovery from Wastes on the Lunar Surface Mission Water Balance

Future extended lunar surface missions will require extensive recovery of resources to reduce mission costs and enable self-sufficiency. Water is of particular importance due to its potential use for human consumption and hygiene, general cleaning, clothes washing, radiation shielding, cooling for extravehicular activity suits, and oxygen and hydrogen production. Various water sources are inherently present or are generated in lunar surface missions, and subject to recovery. They include: initial water stores, water contained in food, human and other solid wastes, wastewaters and associated brines, ISRU water, and scavenging from residual propellant in landers. This paper presents the results of an analysis of the contribution of water recovery from life support wastes on the overall water balance for lunar surface missions. Water in human wastes, metabolic activity and survival needs are well characterized and dependable figures are available. A detailed life support waste model was developed that summarizes the composition of life support wastes and their water content. Waste processing technologies were reviewed for their potential to recover that water. The recoverable water in waste is a significant contribution to the overall water balance. The value of this contribution is discussed in the context of the other major sources and loses of water. Combined with other analyses these results provide guidance for research and technology development and down-selection.

Fisher, John W.↗