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At least 1,207 records · Page 67
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Estimation of the Thermal-Fluids and Thermal-Structural Performance of Helium-Cooled Modular Finger-Type Divertors

Over the past decade, our group has investigated the thermal-fluid performance of the helium-cooled modular divertor with multiple jets (HEMJ) and a simplified “flat” design of the HEMJ for long-pulse magnetic fusion energy (MFE) reactors. Experimental studies were performed in a helium (He) loop at the prototypical pressure of 10 MPa, nearly prototypical He temperatures and incident heat fluxes using test sections made from stainless steel and tungsten alloys. Correlations for average Nusselt numbers and pressure loss coefficients were developed from the data and are used to validate computational fluid dynamics (CFD) models. Here, this work presents updated thermal-fluids performance curves based on these correlations that estimate the maximum heat flux that can be accommodated by the plasma-facing surface and coolant pumping power requirements at prototypical operating conditions. Thermal-structural performance curves developed from ITER structural design criteria are introduced, which include protection against ductile and non-ductile failure, ratcheting fatigue, and creep fatigue. The performance design curves for these finger-type divertors demonstrate that the “flat” design, with a significantly less complicated geometry than the HEMJ, has thermal-fluid and thermal-structural performance comparable to the original HEMJ concept, and may be superior in terms of non-ductile failure criteria.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Integrated simulation of U-10Mo monolithic fuel swelling behavior

Here, a separate computational branch has been implemented within the DART (Dispersion Analysis Research Tool) computational code to simulate the swelling behavior of U-10Mo monolithic fuel under the operating conditions of high-power research and test reactors (RTRs). The monolithic branch of the DART code implements a mechanistic rate-theory-based fission-gas-behavior model for the calculation of fission gas swelling, as well as a suite of thermal, physical, and mechanical models to take into account various processes occurring in RTR fuels during irradiation. In order to accurately simulate and eventually predict U-10Mo monolithic fuel irradiation behavior, the code uses materials properties calculated with lower length-scale computational methods, such as gas atom diffusivity and U-Mo surface energy from atomic simulations and grain-morphology-specific recrystallization kinetics (recrystallized fuel volume fractions vs. fission density) predicted using the phase-field method. The remainder of fission gas behavior parameters used in the model were calibrated with measured intergranular bubble size distributions. With this integrated simulation approach, the swelling behavior of U-10Mo monolithic fuel was simulated for various initial grain sizes at different operating conditions and compared with measured data. Furthermore, because limited experimental data exist for parameter calibration detailed sensitivity studies for the important parameters used in the fission gas behavior model were performed in order to examine their impact on both intergranular gas bubble morphology at low fission density, and on total porosity at high fission density.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Development of Graphite Thermal and Mechanical Modeling Capabilities in Grizzly

Nuclear-grade graphites are used extensively in the core designs of multiple types of advanced nuclear reactors. In the reactor environment, graphite is exposed over long durations to extreme conditions, including high temperatures, radiation and potentially molten salt and oxygen. Exposure to these conditions can cause several degradation mechanisms in graphite, including nonuniform volumetric strains induced by irradiation and thermal expansion, which lead to stresses that can compromise the performance of graphite components. Evaluating component integrity, predicting component performance over the reactor lifetime, and developing design standards all require robust tools for predicting fracture initiation and propagation in graphite structural components in nuclear reactors. This report documents progress in an ongoing effort to develop modeling and simulation tools in the Grizzly code for predicting the performance of graphite exposed to reactor conditions. Recent developments include a set of thermal and mechanical models that now include the IG-110, NBG-18, and H-451 graphite grades. Improvements have also been made to a nonlinear damaged plasticity model applicable to predicting damage under tension and compression to quasibrittle materials, including graphite. In addition, enhancements have been made to the extended finite element method implementation targeted at simulating graphite fracture. These include new capabilities for crack nucleation in the interior of a solid body, improved treatment of crack nucleation on free surfaces, and more robust modeling approaches for crack growth approaching free surfaces or other cracks.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Effects of Interactions Between Produced Formation Fluid and Rock Matrix on Pore Structure of Caney Shale, Southern Oklahoma

ABSTRACT: Rock-fluid interactions change properties of shales during exploitation. To investigate effects of rock-fluid interactions on pore structure of shales matrix after hydraulic fracturing, powder samples from two late Mississippian Caney Shale cores in the Ardmore Basin, southern Oklahoma, were used to react with formation produced fluid from the field in the batch reactor analysis. X-ray diffraction for mineralogy and Low-pressure nitrogen adsorption isotherms for pore structure were measured for original, after-7days, and after-30days samples. Results show that the samples consist mainly of quartz, followed by clay minerals, carbonates, and feldspar. The pore sizes of micropore (<2 nm) and mesopore (2-50 nm) increase 14%-233% due to dissolution of pyrite, feldspar, and carbonates after 7 days. Due to the transformation from smectite to illite and the increase of pore size, the specific surface area (SSA) decreases after 7-days interactions. After 30-days interactions, the micropore volume slightly increases and the mesopore and macropore volume decreases. Due to the decrease of pore size, the SSA of 30-days reacted samples increases correspondingly and is lower (for the clay-rich sample) or higher (for the calcareous sample) than that of the unreacted samples. Findings improve our understanding of dynamic alteration of shale properties during production. 1. INTRODUCTION Energy demand will continuously grow owing to the increasing global population as well as energy consumption (EIA, 2023). On the other hand, shale gas and oil reshaped the energy market in the United States, enabling the United States to become a net-export of natural gas country in 2017 (EIA, 2023). However, shale reservoirs are challenging tight formations that are still poorly understood in the extraction and production of hydrocarbons (Ross and Bustin, 2009; Curtis et al., 2012; Xiong et al., 2015, 2021a; Li Y. et al., 2016; Gong et al., 2019a; Benge et al., 2021; Awejori et al., 2022; Huang et al., 2022). One of the most challenging topics is the rock-fluid interactions post hydraulic fracturing and its subsequent impacts on the pore structures of fractured formation matrix.

Xiong, Fengyang↗
NASA NTRSDate not supplied

Lattice Confinement Fusion & Fast Fission for Space

Lattice Confinement Fusion (LCF) 1 • Fusion of hydrogen isotopes into helium isotopes • No tokamak magnets or laser power supplies • Nuclear fuel is confined and triggered within lattice • Nuclear reactions enhanced by electron screening • Extended Electrodynamics (EED) role • NASA published results in Physical Review C 2,3 • Locally hot but globally cold • Commercialized in 2025 to produce medical radioisotopes 4 Lattice Confinement Fusion Fast-Fission Hybrid Reactor • Fusion neutrons fission uranium, spent fuel rods or thorium • No enriched uranium, cleaner fission! • Demonstrated with US Navy and GEC • Supported by NASA STMD (NIAC) and NSF Application • LEO, Deep Space Power and High I sp Propulsion • (Terrestrial)

lattice confinement fusion↗
DOE OSTI2020

An Efficient 1-D Thermal Stratification Model for Pool-Type Sodium-Cooled Fast Reactors

Investigating thermal stratification in the upper plenum of a sodium fast reactor (SFR) is presently a technology gap in SFR safety analysis. Understanding thermal stratification will promote safe operation of the SFR before its commercial deployment. Stratified layers of liquid sodium with a large vertical temperature gradient could be established in the upper plenum of an SFR during a down-power or a loss-of-flow transient. These stratified layers are unstable and could result in uncertainties for the core safety of an SFR. In order to predict the occurrence of the thermal stratification efficiently, we developed a one-dimensional (1-D) transport model to estimate the temperature profile of the ambient fluid in the upper plenum. This model demands much less computational effort than computational fluid dynamics (CFD) codes and provides calculations with higher fidelity than historical system-level codes. Two flow conditions were considered separately in the current study depending on if in-vessel components are presented in the upper plenum. For the condition where in-vessel components, specifically the upper internal structure, are presented, we assumed that the impinging sodium was evenly dispersed in the ambient fluid within the distance between the bottom of the in-vessel component and the jet inlet surface. For the condition where no in-vessel components are presented, we assumed that the impinging sodium was evenly dispersed in the ambient fluid within the jet length, which was determined through data-driven trainings. The newly developed 1-D model showed similar performance with the CFD model in both cases. However, due to the assumption of flat profiles of the impinging jet axial dispersion rate, nonnegligible discrepancies between the 1-D prediction and the measured data were observed.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2026

Evaluation of corrosion coupons exposed to molten-chloride electrorefiner salts long term

Corrosion mitigation has long been considered a challenge in long-term service of both pyrochemical reprocessing and molten salt reactor applications. Corrosion coupons consisting of 2.25Cr 1Mo ASME SA387 Grade 22 Class 2 steel, were installed in a uranium electrorefiner containing molten chloride salt at 500°C. This electrorefiner went into service at Idaho National Laboratory in 1994. Seven of these coupons were removed in 2019 (i.e., submerged in molten salt for 25 years) for analysis and characterization including neutron imaging, optical microscopy, scanning electron microscopy, and wave/energy dispersive x-ray spectroscopy. The results showed that all seven coupons experienced between 30 to 175 µm of surface degradation without evidence of brittle fracture. Meaning there was no evidence of accelerated corrosion or fractures, and the average steel surface loss was of less than 5 µm per year. All indications were that the environmental conditions inside the electrorefiner were favorable for the steel alloy.

36 - MATERIALS SCIENCE↗
DOE OSTI2020

Examination of Legacy Metallic Fuel Pins (U-10Zr) Tested in FFTF

The MFF series of metallic fuel (U-10wt%Zr) tests performed in the Fast Fuel Test Facility (FFTF) were the beginning tests to qualify the fuel as a driver fuel for FFTF. They all performed very well, to relatively high burnup and with no pin breaches. Tests MFF-3 and MFF-5 were chosen to be destructively examined because they were run at high peak cladding temperatures, 643 and 649°C respectively, and to modest to high burnup (138 and 101 MWd/kgM respectively). They were the only sodium fast reactor (SFR) metallic tests, with long (91.4 cm) fuel columns and clad in HT9, to be operated at these high temperatures. Extensive operating condition reconstruction was performed to accompany the examination, providing detailed operating conditions axially along the pins. In addition, detailed ORIGEN calculations provided fuel burnup along the pin length. These calculations and detailed cross-section metallography allowed a PhD dissertation to be performed where the fuel/cladding chemical interaction (FCCI) was modeled using Fickian and Soret Effect driven diffusion of rare earth fission products to the cladding surface and interact with the cladding. The examination of the MFF-3 and MFF-5 pins included neutron radiography, axial profilometry including pin bow and length measurements, precision gamma scanning, plenum gas puncturing to measure fission gas release, detailed chemical/isotopic analyses of fuel samples to confirm burnup calculations, and metallography of pin cross-sections, including micro-hardness measurements of fuel and cladding.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Extended Modeling of DOE Sealed Canisters with Updated Chemistry Models

Road-ready and final disposition packaging configurations for the advanced test reactor (ATR) fuel currently specifies storage within helium backfilled DOE sealed standard canisters. The aluminum cladding of the ATR fuel contains an oxyhydroxide layer of boehmite/bayerite that generates hydrogen when subjected to irradiation. Understanding the effect of this hydrogen buildup over time to important for long term storage considerations. Previous modeling efforts have built a coupled CFD-chemical model to simulate the temperature gas phase concentrations within the DOE sealed standard canisters. These models have coupled the temperature conditions to both the gas phase radiolysis chemistry and the surface chemistry associated with the oxyhydroxide layer. The previous iteration of the model utilized constant G-values for the hydrogen generation a 50-year period. This new iteration of the model utilizes new experimental data to update the hydrogen generation rate, as well as increase the simulated time to a 200-year period. Given the half-life assumed for the primary Cs-137 isotope responsible for gamma radiation in the ATR spent fuel, the 200-year period decreases the decay heat and dose rate of the spent fuel by a factor of 100, which combined with the updated chemistry substantially decreases new hydrogen generation. Continued experimental work has identified trends for aluminum surrogate samples with oxyhydroxide layers for tests done at higher dose rates. At low initial doses a fast generation rate of hydrogen occurs which starts to roll over to a lower generation as the total dose applied increase. A small-scale chemical model was built to replicate as mini-canister surrogate system at SRNL as well as for the smaller capsule tests performed at INL. A variety of chemical models to capture this effect were tested, and a back reaction of H radical absorbing onto the surface, or inhibition of the reaction by significant H 2 cover gas were both able to fit both the mini-canister and the small capsule test data. Both kinetic fits were used to generate data from the new 200-year simulation. As additional experiments continue, the kinetic fits may be adjusted to adapt to new data. However, updated data shows that hydrogen atmosphere has little effect on actual generation data, so the model was reverted to use a star-stepped G-value for low-dose and high-dose regions. For the undried fuel case, the three models differ significantly with an end concentration of 1% for back reaction, 4.2% for inhibition, and 13.1% for constant G-value for the nominal scenario. For the dried fuel case, the nominal cases showed end concentrations of 0.23% for back reaction, 1.9% for inhibition, and 4.2% for const G-value for the nominal scenario. For the constant G-value case that is less conservative than the other two, the pressure for undried fuel increases to 1.94 atm over 200 years for the nominal case and 2.11 atm for the high decay heat case. The total hydrogen concentration after 200 years is 4.92% for the low decay heat case and 27.4% for the high decay heat case for undried fuel and is 1.4% and 9.6% for low and high decay heat fuel for the dried fuel case. If small amount of residual air is present, the potential for nitric acid formation of 595, 1568, and 2816 ppm for the lower, nominal, and upper fuel decay heat can occur. At long-term timeframes no significant shift in major species present occurs, so no appreciable amount of oxygen is present in the system. The continued rate of hydrogen generation in the canister occurs at an increase of 0.04% by mole over the final 10 years. Increasing the model range out to 1000 years continues to drop the hydrogen generation rate through decreasing dose rate.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Analysis of attenuation data from the decommissioned ZIon unit 1 reactor pressure vessel beltline weld

In order to examine the attenuation of radiation damage through the thickness of an irradiated reactor pressure vessel (RPV), four segments were acquired from the Zion Unit 1 power plant RPV after the plant was decommissioned. The Zion Unit 1 RPV Beltline Weld Segment 1 was cut into seven blocks, consisting of five base metal and two beltline welds from the high fluence region of the segment. Through-wall test specimens were machined and tested. Specimens included those used for Charpy impact, Master Curve fracture toughness testing, and chemical analysis. The observed through-thickness ductile-to-brittle transition temperatures in the beltline weld deviated significantly from the expected behavior based on the attenuation of fast fluence as a function of depth into the RPV. Beginning at the inside surface, the 41-J Charpy transition temperature was either flat or slightly increasing until the ¾ -T location. The results of a simple, model-based analysis of the Zion beltline weld material that included the irradiation conditions and material chemistry were generally consistent with industry trend curves and the standard attenuation model, rather than the observed data. Although there was no archive material from the RPV available to permit measurement of the unirradiated properties, fracture toughness specimens fabricated from archive surveillance weld were used to obtain an estimate of the initial through-thickness values of the Charpy transition temperature. The Charpy shifts obtained using this approach were similarly in disagreement with the predictions of the US NRC Regulatory Guide 1.99, Rev. 2. However, testing of irradiated Charpy specimens taken from the RPV following post-irradiation annealing (10 hr. at 500 °C) provided a quite different estimate of the unirradiated properties which improved the agreement between the inferred through-thickness Charpy shifts and exponential attenuation model included in Regulatory Guide 1.99/2. In conclusion, the analysis of the Zion data and data obtained in previous post-mortem examinations of decommissioned RPVs indicates that more work is needed to understand the through-thickness properties of RPV materials in order to properly assess through-wall damage attenuation.

Charpy impact↗
DOE OSTI2022

BISON validation of FeCrAl cladding mechanical failure during simulated reactivity-initiated accident conditions*

Here, a reactivity-initiated accident (RIA) is a postulated design basis accident in light water reactors (LWRs) in which a rapid reactivity insertion induces a fission rate increase and a fuel pellet temperature rise. During RIA, the fuel pellet thermally expands and may cause pellet-cladding mechanical interaction (PCMI). Separate effects PCMI tests were performed on C26M FeCrAl cladding tube samples, introducing biaxial stress via a well understood modified burst test (MBT) system. A high-speed camera in the MBT system captured the projections, covering a 360° view of the cladding deformation and enabling a digital image correlation (DIC) method to quantify the surface strains with high fidelity. Representative hot zero-power RIA mechanical loading conditions were applied to the sample, and the test duration ranged from 20 to 500 ms at an average temperature of 573 K. BISON finite element–based fuel performance code modeling was performed against the high-fidelity DIC data produced from the MBTs. Validation calculations were conducted with 2D models and systematically compared with test data of the burst time, burst pressure, burst hoop strains, and hoop strain rates. Based on the behaviors from the separate effects test, BISON calculations satisfactorily predicted the cladding deformation behaviors. Sensitivity analysis was conducted to identify highly influential mechanical properties responsible for the cladding failure behavior during the MBT experiments. The results highlight the significance of the cladding's mechanical strength in governing cladding strain, followed by the significance of the mechanical interactions between the pellet and the cladding.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Microstructural and compositional evolutions in $γ$-LiAlO 2 pellets during ion irradiation at an elevated temperature

This study reports on the microstructural, phasic and compositional evolutions of γ-LiAlO 2 during ion irradiation. Polycrystalline γ-LiAlO 2 pellets were irradiated sequentially with He + and D 2 + ions to the same combined fluences of up to 3 × 10 17 (He + +D + )/cm 2 at 773 K. The irradiated pellets were characterized using scanning transmission electron microscopy and atom probe tomography. Surface amorphization likely due to radiolysis and planar defects as a possible precursor for formation of precipitates are created at 5 × 10 16 (He + +D + )/cm 2 , followed by the formation of nano-sized precipitates and fractures at higher doses. Spinel-like precipitates of non-stoichiometric LiAl 5 O 8 and gas-filled cavities are observed to grow with increasing dose. Faceted precipitates and rounded cavities appear at 2 × 10 17 (He + +D + )/cm 2 . With further increasing ion fluence to 3 × 10 17 (He + +D + )/cm 2 , amorphization of the precipitates takes place and micron-sized fractures appear. Surface exfoliation could occur at an extremely high ion fluence. There are compositional changes in the γ-LiAlO 2 pellets during the microstructural evolution. In the precipitate and amorphized regions, Li concentrations decrease to ~7 and 3.7 at.% from 25 at.% in γ-LiAlO 2 , respectively. This study reveals a full-cycle microstructural evolution with corresponding compositional changes in γ-LiAlO 2 pellets during ion irradiation at 773 K. In conclusion, the data could help model, assess, and predict the material performance during neutron irradiation in reactors.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

Methane partial oxidation under periodic reaction conditions on Pt/Al 2 O 3

The increasing interest in utilizing methane, the primary component of natural gas, for chemical production has spurred research into methane partial oxidation (MPO) as an alternative to traditional steam methane reforming (SMR). MPO has lower energy requirements and potential for carbon capture, making it an attractive option for hydrogen production. Challenges remain, however, such as carbon deposition leading to degradation and achieving high hydrogen selectivity. Here, the impact of periodic reactor operation on MPO over a Pt/Al 2 O 3 catalyst was studied, primarily via varying reactor inlet compositions. Experiments were conducted using periodic operation strategies to assess the influence of changing reactant inlet concentrations on hydrogen formation during MPO. The results suggest that cycling between mixtures with low and high oxygen content can lead to transient hydrogen formation rates that surpass those achieved at steady state. Control experiments and density functional theory (DFT) calculations show that enhanced hydrogen formation can be attributed to the reaction between CO with hydroxyl groups at the metal and alumina support interface. Furthermore, this work underscores the critical role of surface coverages at the metal–support interface and suggests avenues for future exploration, including alternative support materials with higher OH mobility and changes in the cycling scheme to enhance catalyst performance under periodic conditions.

03 NATURAL GAS↗
DOE OSTI2020

Generation of reactive species in water film dielectric barrier discharges sustained in argon, helium, air, oxygen and nitrogen

Activation of liquids with atmospheric pressure plasmas is being investigated for environmental and biomedical applications. When activating the liquid using gas plasma produced species (as opposed to plasmas sustained in the liquid), a rate limiting step is transport of these species into the liquid. To first order, the efficiency of activating the liquid is improved by increasing the ratio of the surface area of the water in contact with the plasma compared to its volume—often called the surface-to-volume ratio (SVR). Maximizing the SVR then motivates the plasma treatment of thin films of liquids. In this paper, results are discussed from a computational investigation using a global model of atmospheric pressure plasma treatment of thin water films by a dielectric barrier discharge (DBD) sustained in different gases (Ar, He, air, N 2 , O 2 ). The densities of reactive species in the plasma activated water (PAW) are evaluated. The residence time of the water in contact with the plasma is increased by recirculating the PAW in plasma reactor. Longer lived species such as H 2 O 2aq and NO 3 - aq accumulate over time (aq denotes an aqueous species). DBDs sustained in Ar and He are the most efficient at producing H 2 O 2aq , DBDs sustained in argon produces the largest density of NO 3 - aq with the lowest pH, and discharges sustained in O 2 and air produce the highest densities of O 3aq . Finally, comparisons to experiments by others show agreement in the trends in densities in PAW including O 3aq , OH aq , H 2 O 2aq and NO 3 - aq , and highlight the importance of controlling desolvation of species from the activated water.

42 ENGINEERING↗
DOE OSTI2024

Boron carbide films with reduced nodular defect density deposited by full-face erosion radio-frequency magnetron sputtering

Boron carbide coatings deposited by sputtering with conventional planar magnetrons exhibit nodular growth defects. Here, we demonstrate that the density of nodular defects in B 4 C films can be greatly reduced by using a full-face-erosion magnetron source that achieves a more uniform target erosion by sweeping the confining magnetic field over the target surface. Here, we systematically study properties of B 4 C films deposited on substrates with different tilt angles and demonstrate deposition of ultrathick films with compressive residual stress below 200MPa.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Understanding the Quadrupole Mass Filter and Testing a High-Resolution QMS RGA for ITER

A common type of residual gas analyzer is the quadrupole mass spectrometer. One of the main components within this instrument is a mass filter known as the quadrupole. It is responsible for the selective throughput of the ionized gas particles - by ascending mass number - prior to ion impacts on the analyzer (or detector) surface from which the ion current signal is generated for processing. However, the quadrupole is not fully described in relation to the electric field characteristics and the function as an ion mass separator. This paper describes the basic origins of the electrical design, the intricate assembly criteria, and performance of the quadrupole within the spectrometer. A specialized quadrupole mass spectrometer is part of a configuration for a diagnostic gas analyzer system planned for ITER, a fusion research machine. It has a verified capability, essential as a diagnostic criterion for this reactor project, to successfully deconvolute the mass signals of Helium-4 and deuterium (reactor fuel exhaust gases, separated by only 0.026 atomic mass units), down to a relative three-percent concentration of the former gas. The associated preliminary testing, performed at the Oak Ridge National Laboratory, is also addressed. Finally, one of the key parameters used to express gas concentration, the relative sensitivity factor, will be explained, including an evaluation of dependency on other variables.

Marcus, Chris [ORNL] (ORCID:0000000190139636)↗
NASA NTRS2010

Integration of In-Situ Resource Utilization Into Lunar/Mars Exploration Through Field Analogs

The NASA project to develop In-Situ Resource Utilization (ISRU) technologies, in partnership with commercial and international collaborators, has achieved full system demonstrations of oxygen production using native regolith simulants. These demonstrations included robotic extraction of material from the terrain, sealed encapsulation of material in a pressurized reactor; chemical extraction of oxygen from the material in the form of water, and the electrolysis of water into oxygen and hydrogen for storage and reuse. These successes have provided growing confidence in the prospects of ISRU oxygen production as a credible source for critical mission consumables in preparation for and during crewed missions to the moon and other destinations. Other ISRU processes, especially relevant to early lunar exploration scenarios, have also been shown to be practical, including the extraction of subsurface volatiles, especially water, and the thermal processing of surface materials for civil engineering uses and for thermal energy storage. This paper describes these recent achievements and current NASA ISRU development and demonstration activity. The ability to extract and process resources at the site of exploration into useful products such as propellants, life support and power system consumables; and radiation and rocket exhaust plume debris shielding, known as In-Situ Resource Utilization or ISRU, has the potential to significantly reduce the launch mass, risk, and cost of robotic and human exploration of space. The incorporation of ISRU into missions can also significantly influence technology selection and system development in other areas such as power, life support, and propulsion. For example. the ability to extract or produce large amounts of oxygen and/or water in-situ could minimize the need to completely close life support air and water processing system cycles, change thermal and radiation protection of habitats, and influence propellant selection for ascent vehicles and surface propulsive hoppers. While concepts and even laboratory work on evaluating and developing ISRU techniques such as oxygen extraction from lunar regolith have been going on since before the Apollo 11 Moon landing, no ISRU system has ever flown in space, and only recently have ISRU technologies been developed at a scale and at a system level that is relevant to actual robotic and human mission applications. Because ISRU hardware and systems have never been demonstrated or utilized before on robotic or human missions, architecture and mission planners and surface system hardware developers are hesitant to rely on ISRU products and services that are critical to mission and system implementation success. To build confidence in ISRU systems for future missions and assess how ISRU systems can best influence and integrate with other surface system elements, NASA, with international partners, are performing analog field tests to understand how to take advantage of ISRU capabilities and benefits with the minimum of risk associated with introducing this game-changing approach to exploration. This paper will describe and review the results of four analog field tests (Moses Lake in 6/08, Mauna Kea in 11/08. Flagstaff in 9/09; and Mauna Kea in 1/10) that have begun the process of integrating ISRU into robotic and human exploration systems and missions, and propose future ISRU-related analog field test activities that can be performed in collaboration with international space agencies.

Sanders, Gerald B.↗
DOE OSTI2025

Inorganic Alterations in Unconventional Shale Reservoirs: Importance of Additive and Base Fluid Chemistry

The effective development of unconventional petroleum systems requires the use of significant water resources. In an effort to reduce the consumption of freshwater resources for hydraulic fracturing, highly saline produced waters are increasingly recycled for use as a base fluid. However, there are significant knowledge gaps regarding potential water–rock interactions resulting from the introduction of produced waters and associated additives into shale reservoirs such as formation and deposition of mineral scale, which can negatively affect hydrocarbon production through wellbore restriction and damage to hydraulically generated fractures. To assess the impacts of field stimulation practices in the subsurface, a series of laboratory experiments were completed using (a) three distinct sedimentary rock formations of the Midland Basin (Texas, USA) and (b) additives with two different base fluids: municipal fresh water and clean brine. The experimental approach used relevant injection sequences and mixing ratios in specialized reactors for 3 weeks. Static pressurized experiments and nonpressurized time-resolved experiments were undertaken. The resulting solids and liquids were analyzed by using a variety of laboratory- and synchrotron-based techniques. The use of an acid spearhead (15% HCl) resulted in texturing of both clay-rich and calcareous shales, which can temporarily enhance porosity but subsequently result in mineral scale deposition. The primary matrix scale was Fe(III)-bearing phases, which occurred in all experiments regardless of base fluid chemistry. Additionally, strontium sulfate (SrSO 4 ) precipitated on shale surfaces when clean brines were used. It was concluded that clean brine was the main source of Sr 2+ species, while persulfate breaker degradation and oxidation of pyrite were the sources of SO 4 2– . Sulfate scaling was more pronounced in clay-rich shales, suggesting that Sr sorption is important for promoting celestite formation. This work demonstrates that mineral scale deposition is a complex phenomenon, whereby the type and proportions of various mineral phases are determined from reservoir alteration processes and coprecipitation of constituents from injection fluids. In conclusion, the experimental results shown here should be considered when evaluating different base fluids and additives in order to mitigate mineral precipitation in unconventional shale reservoirs, which could result in reservoir degradation.

Jew, Adam D. [SLAC National Accelerator Laboratory↗