Search NASA⌕ Search

SEARCH · Search NASA

Results for “TREAT”

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.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Enhancing TREAT’s Pulsing Capabilities Using TREAT Upgrade Fuel

The Transient Reactor Test Facility (TREAT) at Idaho National Laboratory is an air-cooled, thermal, heterogeneous facility used to test reactor materials in simulated accident conditions by inducing fission heating with intense neutron pulses. TREAT operated from 1959 until 1994 with the primary goal of testing fast reactor fuels and was brought back online in 2017 to re-establish DOE's nuclear fuels transient testing capabilities. Beginning in the late 1970s, Idaho National Laboratory worked in conjunction with other organizations to increase TREAT's capability for in-pile testing. While these upgrades had not been implemented by the time the reactor was put into standby mode, new assemblies and graphite-urania fuel blocks with increased uranium concentrations had already been designed and fabricated. This project, as part of Idaho National Laboratory's Department of Reactor Physics, modeled the TREAT Upgrade (TU) fuel assemblies in MCNP and implemented them in an existing TREAT model. Calculations were performed using INL's high performance computer to find a critical combination of TU and standard fuel and this geometry's excess reactivity. Future work will include calculating a power coupling factor in the experiment, determining power peaking factors throughout the core, and assessing how much reactivity would be added by replacing the Inconel-625 cladding with silicon carbide.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2019

Enhancing TREAT's pulsing abilities using TREAT upgrade fuel. SULI Intern Report

The Transient Reactor Test Facility (TREAT) at Idaho National Laboratory is an air-cooled, thermal, heterogeneous facility used to test reactor materials in simulated accident conditions by inducing fission heating with intense neutron pulses. TREAT operated from 1959 until 1994 with the primary goal of testing fast reactor fuels and was brought back online in 2017 to re-establish DOE’s nuclear fuels transient testing capabilities. Beginning in the late 1970s, Idaho National Laboratory worked in conjunction with other organizations to increase TREAT’s capability for in-pile testing. While these upgrades had not been implemented by the time the reactor was put into standby mode, new assemblies and graphite-urania fuel blocks with increased uranium concentrations had already been designed and fabricated. This project, as part of Idaho National Laboratory’s Department of Reactor Physics, modeled the TREAT Upgrade (TU) fuel assemblies in MCNP and implemented them in an existing TREAT model. Calculations were performed using INL’s high performance computer to find a critical combination of TU and standard fuel and this geometry’s excess reactivity. Future work will include calculating a power coupling factor in the experiment, determining power peaking factors throughout the core, and assessing how much reactivity would be added by replacing the Inconel-625 cladding with silicon carbide.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2020

Impact of Grain Size on Performance Degradation of TREAT LEU

We argue that radiation damage induced degradation of thermal conductivity does not set a lower limit on fuel grain sizes for the low enriched uranium fuel design of the Transient Reactor Test Facility (TREAT). Earlier work reports that smaller grains cause a larger degradation of thermal conductivity than larger grains constraining the smallest feasible size of fuel grains. This work assesses TREAT’s transient performance in the presence of radiation damage. The difference between the two studies is in treating damaged and fresh graphite as serial (this work) or parallel (previous) heat resistors. We use a multiphysics model of TREAT fuel grains to compute the reduction in transient capability measured by the total deposited energy as a function of irradiation dose. We find that radiation damage has a negligible effect on energy deposition.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Iodine capture with mechanically robust heat-treated Ag-Al-Si-O xerogel sorbents

Silver-loaded heat-treated aluminosilicate xerogels (Ag-HTX) were evaluated as sorbents for iodine [I2(g)] capture. The heat treatment step was performed to help increase the mechanical integrity of the gels. The synthesized xerogels were characterized using powder X-ray diffraction, scanning electron microscopy, energy-dispersive X-ray spectroscopy, transmission electron microscopy, Brunauer-Emmett-Teller analysis, gravimetric iodine loading, nanoindentation, and dynamic mechanical analysis. The structural and chemical analyses of Ag-HTX showed uniform distribution of Ag throughout the gel network after Ag-exchange. After I2(g) capture, the AgI crystallites were observed in the sorbent verifying chemisorption as the iodine capture mechanism. Iodine loading of this xerogel was 0.43 g g-1 at 150°C over 1 d and 0.52 g g-1 at 22°C over 33 d. The specific surface area of Ag-HTX was 202 m2 g-1 and decreased to 87 m2 g-1 after iodine loading. The hardness of the heat-treated xerogel was 160 times higher than heat-treated aerogel of the same composition. The heat-treatment process increased the modulus value to 40.77 MPa from 6.99 MPa of as-made xerogel, demonstrating the need for this added step in the synthesis process. These results show that Ag-HTX is a promising sorbent for I2(g) capture with good loading capacity and mechanical stability.

xerogels, aerogels, iodine capture, hardness testi↗
DOE OSTI2022

Desalinating a Real Hyper-Saline Pre-Treated Produced Water via Direct-Heat Vacuum Membrane Distillation

Membrane distillation (MD) is an emerging thermal desalination technology capable of desalinating waters of any salinity. During typical MD processes, the saline feedwater is heated and acts as the thermal energy carrier; however, temperature polarization (as well as thermal energy loss) contributes to low distillate fluxes, low single-pass water recovery and poor thermal efficiency. An alternative approach is to integrate an extra thermal energy carrier as part of the membrane and/or module assembly, which can channel externally provided heat directly to the membrane-feedwater interface and/or along the feed channel length. This direct-heat delivery has been demonstrated to increase single-pass water recovery and enhance the overall thermal efficiency. We developed a bench-scale direct-heated vacuum MD (DHVMD) process to desalinate pre-treated oil and gas "produced water" with an initial total dissolved solids of 115,500 ppm at a feed temperature ranging between 24 and 32 degrees C. We evaluated both water flux and specific energy consumption (SEC) as a function of water recovery. The system achieved a 50% water recovery without significant scaling, with an average flux >6 kg m-2 hr-1 and a SEC as low as 2,530 kJ kg-1. The major species of mineral scales (i.e., NaCl, CaSO4, and SrSO4) that limited the water recovery to 68% were modeled in terms of thermodynamics and identified by scanning electron microscopy and energy-dispersive X-ray spectroscopy. In addition, we further developed and employed a physics-based process model to estimate temperature, salinity, water transport and energy flows for full-scale vacuum MD and DHVMD modules. Model results show that a direct-heat input rate of 3,600 W can increase single-pass water recovery from 2.1% to 3.1% while lowering the thermal SEC from 7,800 kJ kg-1 to 6,517 kJ kg-1 in an unoptimized module. Finally, the scaling up potential of DHVMD process is briefly discussed.

brine↗
DOE OSTI2020

The Generality of the GUGA MRCI Approach in COLUMBUS for Treating Complex Quantum Chemistry

The core part of the program system COLUMBUS allows highly efficient calculations using variational multireference (MR) methods in the framework of configuration interaction with single and double excitations (MR-CISD) and averaged quadratic coupled-cluster calculations (MR-AQCC), based on uncontracted sets of configurations and the graphical unitary group approach (GUGA). The availability of analytic MR-CISD and MR-AQCC energy gradients and analytic nonadiabatic couplings for MR-CISD enables exciting applications including, e.g., investigations of π-conjugated biradicaloid compounds, calculations of multitudes of excited states, development of diabatization procedures, and furnishing the electronic structure information for on-the-fly surface nonadiabatic dynamics. With fully variational uncontracted spin-orbit MRCI, COLUMBUS provides a unique possibility of performing high-level calculations on compounds containing heavy atoms up to lanthanides and actinides. Crucial for carrying out all of these calculations effectively is the availability of an efficient parallel code for the CI step. Configuration spaces of several billion in size now can be treated quite routinely on standard parallel computer clusters. Emerging developments in COLUMBUS, including the all configuration mean energy multiconfiguration self-consistent field method and the graphically contracted function method, promise to allow practically unlimited configuration space dimensions. Spin density based on the GUGA approach, analytic spin-orbit energy gradients, possibilities for local electron correlation MR calculations, development of general interfaces for nonadiabatic dynamics, and MRCI linear vibronic coupling models conclude this overview.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2019

Chemical oxygen demand (COD) removal from bio-treated coking wastewater by hydroxyl radicals produced from a reduced clay mineral

Removal of chemical oxygen demand (COD) from wastewater has been of considerable interest for several decades, but detailed mechanisms are poorly known. Here we present an environmental friendly method for removing COD from coking wastewater by using chemically reduced nontronite (rNAu-2). Hydroxyl radicals (·OH) produced from oxidation of structural Fe(II) in rNAu-2 in phosphate buffer at pH 6.5, oxidized various compounds in wastewater and thus significantly decreased COD (from 199.10 ± 0.90 to 80.4 ± 1.97 mg/L) within 120 h. The amount of COD removal was positively correlated with the amount of ·OH production, with removal efficiency depending on pH and buffer type. Mass spectrometric analysis of dissolved organic matter (DOM) suggested an oxidative pathway for COD removal. Excitation-emission matrix (EEM) fluorescence spectroscopy showed that aromatic protein-like compounds accounted for a large fraction of DOM in initial wastewater, but after rNAu-2 treatment, their proportion decreased. A sequential treatment of wastewater by rNAu-2 and a common wastewater bacterium Alcaligenes faecalis demonstrated that different orders of chemical and biological treatments affected COD removal efficiency, likely because chemical and microbial treatments targeted different compounds in coking wastewater.

Zhou, Ziqi↗
DOE OSTI2018

ATF Transient Prescription Tests in TREAT

EXECUTIVE SUMMARY The Transient Reactor Test facility (TREAT) was constructed in the late 1950’s and provided thousands of transient irradiations before being placed in standby mode in 1994. In 2017, it resumed operation in order to reclaim its crucial role in nuclear-heated safety research. TREAT is a transient reactor where inherently-safe core physics, a nimble transient rod drive system, and a philosophy of continual facility improvement work together to enable flexible power maneuvers for fuel safety research. While the facility’s flexible design and multi-mission nature saw historic experiments for numerous transient types, TREAT was not specifically adapted to optimize or demonstrate its full performance capability for Light Water Reactor (LWR) postulated accidents including Hot Zero Power Reactivity Initiated Accidents (HZP-RIA) and Loss of Coolant Accidents (LOCA). The physics of HZP-RIA drive very brief pulses with fairly large energy excursions. TREAT is currently capable of approaching HZP-RIA pulse widths through transients initiated on a large step insertion of reactivity and terminated shortly thereafter by rapid re-insertion of the transient rods (“clipping”). LOCAs are postulated to occur quite differently, starting from normal power operations followed by a long period of lower power decay heat. TREAT is able to simulate LOCA events with an initial flattop transient for tens of seconds, followed by a sizeable power reduction, to a longer segment where specimen fission heating represents internal decay heat generation. Two reactor physics testing campaigns were undertaken to address the HZP-RIA type transients, neither of which had been performed in any great measure historically to address data gaps in the timing, repeatability of clipping and effect of dynamic transient rod motions on core-to-specimen fission power coupling. The first campaign, referred to as the Accident Tolerant Fuel (ATF) transient prescription tests, employed dosimeters in the core to characterize the effect of rod motions on specimen power coupling for clipped, unclipped, and low-level steady state operations each with approximately the same moderate core energy release. This campaign was accompanied by efforts to measure dosimetry and compare results to analytic predictions. The other HZP-RIA campaign, referred to as Narrow Pulse Width (NPW) transient prescriptions, employed higher step insertions with increased core energy release to demonstrate TREAT’s capability in achieving the narrowest-possible pulse width. Additionally, a series of LOCA transient prescriptions were also performed to demonstrate how well TREAT could perform long-duration shaped transients. Both NPW and LOCA transient prescriptions did not include dosimeters but used existing plant power instrumentation to produce transient power data. All three transient prescription campaigns were successfully carried out in TREAT. The ATF-3 transient prescriptions showed that TREAT can perform clipped transients with remarkable repeatability. Measurement and modeling both showed that the power coupling in small, fissile dosimeters does not change by more than 8% between clipped, unclipped, and steady state runs while uncertainty sources in both the modeling and measurement methods were identified and found to be on the order of 20% in total. Overall, the ATF-3 transient prescription tests were successful in determining repeatability and consistency of dosimeter power coupling, developing processes for modeling and measuring this crucial parameter, and demonstrating the need for continued efforts in transient prescription work at TREAT in support of future experiment design and post test data interpretation. The NPW and LOCA transient prescriptions were also successfully performed to exhibit TREAT’s capability for these important LWR accident categories. The NPW transients showed pulse widths of ~90 ms, which were the narrowest ever achieved in TREAT to date and relevant for the upper end range of HZP-RIA simulation. This effort provided crucial data for calibrating kinetic models to be used in enhanced pulse-narrowing strategies in upcoming irradiations. Similarly, The LOCA transient prescriptions showed that TREAT can simulate the needed nuclear heating and event time scales for LOCA fuel safety research and provided essential data for future experiment design.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Exosomes derived from thymic stromal lymphopoietin-treated dendritic cells regulate T helper 17/regulatory T cell differentiation via miR-21/Smad7 axis

Highlights: • Exosomes from TSLP-treated DCs promote Th17 differentiation. • Exosomes from TSLP-treated DCs inhibit Tregs differentiation. • MiR-21 is highly expressed in exosomes from TSLP-treated DCs. • Exosomal miR-21 regulates Th17/Treg differentiation by targeting Smad7. Thymic stromal lymphopoietin (TSLP) is associated with fungal keratitis. This work aims to investigate whether TSLP can regulate T helper (Th) 17 and regulatory T cell (Treg) differentiation. We separated dendritic cells (DCs) from peripheral blood of healthy volunteers. DCs were treated with TSLP to activate DCs, and exosomes were obtained. CD{sup +} T cells were incubated with exosomes from TSLP-treated DCs. We found that exosomes from TSLP-treated DCs notably promoted the proportions of Th17 cells and inhibited the proportions of Tregs in the CD4{sup +} T cells. Moreover, exosomes from TSLP-treated DCs enhanced the expression of retinoid-related orphan receptor γt (RORγt) and interleukin 17 (IL-17), and repressed the expression of forkhead box protein P3 (Foxp3) and interleukin 10 (IL-10) in the CD4{sup +} T cells. Furthermore, miR-21 was highly expressed in exosomes from TSLP-treated DCs. Exosomes from TSLP-treated miR-21-silenced DCs promoted Treg differentiation and suppressed Th17 differentiation. Smad7 up-regulation repressed Th17 differentiation and enhanced Treg differentiation, which was abolished by miR-21 overexpression. Smad7 overexpression rescued the effect of exosomes from TSLP-treated DCs on Th17/Treg differentiation. In conclusion, our article confirms that TSLP induces DCs to deliver miR-21 by secreting exosomes, and thus miR-21 regulates Th17/Treg differentiation by inhibiting Smad7. Thus, this work further reveals the biological role of miR-21 in fungal keratitis.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2018

Treated fillers, compositions containing same, and articles prepared therefrom

The present invention includes a process for producing treated filler that includes (a) treating a slurry that includes untreated filler where the untreated filler has not been previously dried, with a treating composition that includes a treating agent, thereby forming a treated filler slurry, and (b) drying the treated filler slurry to produce treated filler. The treating agent can include an unsaturated fatty acid, derivative of an unsaturated fatty acid, or salt thereof. The present invention also is directed to treated filler prepared by the process, as well as rubber compounding compositions and tires including the treated filler.

Votruba-Drzal, Peter Lawrence↗
DOE OSTI2018

Nuclear characterization of a general-purpose instrumentation and materials testing location in TREAT

The Transient Reactor Test facility (TREAT) was constructed in the late 1950s, provided thousands of transient irradiations before being placed in standby in 1994, and resumed operations in 2017 in order to reclaim its crucial role in nuclear-heated safety research. The latter half of TREAT’s historic operation was best known for integral-scale testing of fuel specimens under postulated reactor plant accident conditions, while TREAT’s earlier history included extensive simpler phenomena identification tests that elucidated fundamental behaviors and paved the way for these integral-scale tests. Advances in modern computational capabilities and a resurgence of interest in novel reactor technology have created an opportunity for emphasizing modernized science-based and separate effects test capabilities once again at TREAT. An innovative approach to this type of testing has been developed to leverage minor radioactivity built-in during brief TREAT irradiations in low activation hardware to facilitate handling for materials and instrumentation testing. This capability, termed the Minimal Activation Retrievable Capsule Holder (MARCH) irradiation vehicle system, will be used for inaugural fueled experiments in TREAT’s modern era as well as novel approaches to study materials undergoing neutron irradiation and instrumentation development and qualifications. This paper describes a comprehensive nuclear characterization, obtained via computational modeling, of the test position in the MARCH system. Though directly applicable to the MARCH system, the results also provide general quantification of the nuclear performance of the reactor and potential test materials, crucial for evaluating potential experiment design and response in TREAT. The neutron and photon flux environment was calculated via MCNP with ENDF/B-VII.1 nuclear data. Wire heating rate and DPA calculations were also performed and evaluated using historic TREAT test data. Furthermore, these calculations were performed to provide steady-state baseline reference values typical in both half- and full-slotted TREAT core configurations, enabling design scoping analysis prior to development of more specific design testing needs and transient testing experimentation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗