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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 415 records · Page 23
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Characterization and Testing of Amidoxime-Based Adsorbent Materials to Extract Uranium from Natural Seawater

Extraction of uranium (U) from seawater for use as a nuclear fuel is a significant challenge due to the low concentration of U in seawater (~3.3 ppb) and difficulties to selectively extract U from the background of major and trace elements in seawater. The Pacific Northwest National Laboratory (PNNL)’s Marine Sciences Laboratory (MSL) has been serving as a marine test site for determining performance characteristics (adsorption capacity, adsorption kinetics, and selectivity) of novel amidoxime-based polymeric adsorbents developed at Oak Ridge National Laboratory (ORNL) under natural seawater exposure conditions. This manuscript describes the performance of three formulations (38H, AF1, AI8) of amidoxime-based polymeric adsorbents produced at ORNL in MSL’s ambient seawater testing facility. The adsorbents were produced in two forms, fibrous material (40-100 mg samples) and braided material (5-10 g samples), and exposed to natural seawater using flow-through columns and recirculating flumes. All three formulations demonstrated high 56 day uranium adsorption capacity (>3 g U/kg adsorbent). The AF1 formulation had the best uranium adsorption performance, with a 56 day capacity of 3.9 g U/ kg adsorbent, a saturation capacity of 5.4 g U/kg adsorbent, and ~25 days half-saturation time. The two exposure methods, flowthrough columns and flumes, were demonstrated to produce similar performance results, providing confidence that the test methods were reliable, that scaling up from 10’s of mg quantities of exposure in flow-through columns to gram quantities in flumes produced similar results, and confirm that the manufacturing process produces a homogeneous adsorbent. Adsorption kinetics appear to be element specific, with half-saturation times ranging from minutes for the major cations in seawater, to 8-10 weeks for V and Fe. Reducing the exposure time provides a potential pathway to improve the adsorption capacity of U by reducing the V/U ratio on the adsorbent.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS2016

Post-Test Inspection of NASA's Evolutionary Xenon Thruster Long Duration Test Hardware: Ion Optics

A Long Duration Test (LDT) was initiated in June 2005 as a part of NASA's Evolutionary Xenon Thruster (NEXT) service life validation approach. Testing was voluntarily terminated in February 2014, with the thruster accumulating 51,184 hours of operation, processing 918 kg of xenon propellant, and delivering 35.5 MN-s of total impulse. The post-test inspection objectives for the ion optics were derived from the original NEXT LDT test objectives, such as service life model validation, and expanded to encompass other goals that included verification of in situ measurements, test issue root causes, and past design changes. The ion optics cold grid gap had decreased only by an average of 7% of pretest center grid gap, so efforts to stabilize NEXT grid gap were largely successful. The upstream screen grid surface exhibited a chamfered erosion pattern. Screen grid thicknesses were ≥ 86% of the estimated pretest thickness, indicating that the screen grid has substantial service life remaining. Deposition was found on the screen aperture walls and downstream surfaces that was primarily composed of grid material and back-sputtered carbon, and this deposition likely caused the minor decreases in screen grid ion transparency during the test. Groove depths had eroded through up to 35% of the accelerator grid thickness. Minimum accelerator aperture diameters increased only by about 5-7% of the pretest values and downstream surface diameters increased by about 24-33% of the pretest diameters. These results suggest that increasing the accelerator aperture diameters, improving manufacturing tolerances, and masking down the perforated diameter to 36 cm were successful in reducing the degree of accelerator aperture erosion at larger radii.

Ion Thruster↗
NASA NTRS2020

Characterization of the NASA Glenn Research Center 8- by 6-Foot Supersonic Wind Tunnel Supersonic Test Section (2020 Test)

Recent customer requirements for improved Mach number stability for sonic boom signature measurements has piqued interest in the use of the solid-wall supersonic test section in the NASA Glenn Research Center 8- by 6-Foot Supersonic Wind Tunnel. To supply customers with understanding of the operating conditions and flow quality in this region of the tunnel, characterization tests were performed in the solid-wall supersonic test section in January 2020. The test entry covered the operating range of the facility for three-drive-motor operation. Calibration models were generated along with an improved understanding of operational considerations for this region of the tunnel. The data set acquired included centerline axial and planar flow-field distributions to quantify test section flow quality. The collected data provided an improved understanding of operational considerations for this region of the tunnel as well as allowed for the development of detailed calibration models for the solid-wall supersonic test section.

Supersonic↗
NASA NTRS2020

In-flight Testing of the Injection of the LISA Pathfinder Test Mass into a Geodesic

LISA Pathfinder is a technology demonstrator space mission, aimed at testing key technologies for detecting gravitational waves in space. The mission is the precursor of LISA, the first space gravitational waves observatory, whose launch is scheduled for 2034. The LISA Pathfinder scientific payload includes two gravitational reference sensors (GRSs), each one containing a test mass (TM), which is the sensing body of the experiment. A mission critical task is to set each TM into a pure geodesic motion, i.e. guaranteeing an extremely low acceleration noise in the sub-Hertz frequency bandwidth. The grabbing positioning and release mechanism (GPRM), responsible for the injection of the TM into a geodesic trajectory, was widely tested on ground, with the limitations imposed by the 1-g environment. The experiments showed that the mechanism, working in its nominal conditions, is capable of releasing the TM into free-fall fulfilling the very strict constraint imposed on the TM residual velocity, in order to allow its capture on behalf of the electrostatic actuation. However, the first in-flight releases produced unexpected residual velocity components, for both the TMs. Moreover, all the residual velocity components were greater than maximum value set by the requirements. The main suspect is that unexpected contacts took place between the TM and the surroundings bodies. As a consequence, ad hoc manual release procedures had to be adopted for the few following injections performed during the nominal mission. These procedures still resulted in non compliant TM states which were captured only after impacts. However, such procedures seem not practicable for LISA, both for the limited repeatability of the system and for the unmanageable time lag of the telemetry/telecommand signals (about 4400 s). For this reason, at the end of the mission, the GPRM was deeply tested in-flight, performing a large number of releases, according to different strategies. The tests were carried out in order to understand the unexpected dynamics and limit its effects on the final injection. Some risk mitigation maneuvers have been tested aimed at minimizing the vibration of the system at the release and improving the alignment between the mechanism and the TM. However, no overall optimal release strategy to be implemented in LISA could be found, because the two GPRMs behaved differently.

LISA↗
DOE OSTI2019

Test Plan for Reirradiation and 1400°C Heating Test of AGR-3/4 Compact 10-1

Post irradiation examination (PIE) and heating tests of tristructural isotropic (TRISO)-coated fuel particles, fuel compacts, and graphite from Advanced Gas Reactor (AGR) irradiation experiment AGR 3/4 are in progress. The AGR-3/4 test train consisted of twelve separate capsules irradiated in the northeast flux trap of the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) from December 14, 2011, to April 12, 2014 (Collin 2015b). This test plan describes the reirradiation and subsequent heating test of AGR-3/4 Compact 10-1. One of the main goals of this test is to assess the retention of short-lived fission products I-131 (t1/2 = 8.02 d) and Xe-133 (t1/2 = 5.2 d) in fuel kernels. To generate short-lived fission products, the compact will first be reirradiated in the Neutron Radiography Reactor (NRAD). Following reirradiation, the compact will be removed from NRAD, gamma scanned on the Precision Gamma Scanner (PGS), and then loaded into the Fuel Accident Condition Simulator (FACS) furnace for heating at 1400°C. Fission gases and condensable fission products released from the compact will be collected in the Fission Gas Monitoring System (FGMS) and the FACS condensation plates, respectively.

42 ENGINEERING↗
DOE OSTI2018

Test Plan for the Reirradiation Heating Test of Fuel Particles from AGR 2 Compact 5-4-2

Post irradiation examination (PIE) and high-temperature heating tests of tristructural isotropic (TRISO)-coated fuel particles and compacts from the second Advanced Gas Reactor (AGR) irradiation experiment, AGR 2, are in progress. This test plan describes the reirradiation and subsequent high-temperature testing of loose fuel particles obtained from AGR 2 Compact 5-4-2. To simulate exposed kernels, four of these particles have been cracked at Oak Ridge National Laboratory (ORNL) to produce localized defects that extend through the TRISO layers (outer-pyrolytic carbon [OPyC], silicon carbide [SiC], and inner-pyrolytic carbon [IPyC]) while keeping these layers in-place and maintaining the shape and integrity of the kernel. To generate short-lived fission products, these particles will be reirradiated in the Neutron Radiography (NRAD) reactor prior to heating testing in the Fuel Accident Condition Simulator (FACS) furnace. FACS testing will be at 1600°C, and fission products released from the reirradiated particles will be measured. The intention of the test is to assess the retention of short-lived fission products I-131 (t1/2 = 8.02 days) and Xe-133 (t1/2 = 5.2 days) in the fuel kernels at accident temperatures.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Initial results from safety testing of US AGR-2 irradiation test fuel - 18574

Two cylindrical compacts containing tri-structural isotropic (TRISO)-coated particles with kernels that contained a mixture of uranium carbide and uranium oxide (UCO) and two compacts with UO{sub 2}-kernel TRISO particles have undergone 1600 deg. C safety testing. These compacts were irradiated in the US Advanced Gas Reactor Fuel Development and Qualification Program's second irradiation test (AGR-2). The time-dependent releases of several radioisotopes ({sup 110m}Ag, {sup 134}Cs, {sup 137}Cs, {sup 154}Eu, {sup 155}Eu, {sup 90}Sr, and {sup 85}Kr) were monitored while heating the fuel specimens to 1600 deg. C in flowing helium for 300 h. The UCO compacts behaved similarly to previously reported 1600 deg. C-safety-tested UCO compacts from the AGR-1 irradiation. No failed TRISO or failed SiC were detected (based on krypton and cesium release), and cesium release through intact SiC was very low. Release behavior of silver, europium, and strontium appeared to be dominated by inventory originally released through intact coating layers during irradiation but retained in the compact matrix until it was released during safety testing. Both UO{sub 2} compacts exhibited cesium release from multiple particles whose SiC failed during the safety test. Europium and strontium release from these two UO{sub 2} compacts appeared to be dominated by release from the particles with failed SiC. Silver release was characteristically similar to the release from the UCO compacts in that an initial release of the majority of silver trapped in the matrix occurred during ramping to 1600 deg. C. However, additional silver release was observed later in the safety testing due to the UO{sub 2} TRISO with failed SiC. Failure of the SiC layer in the UO{sub 2} fuel may have been dominated by CO corrosion, as opposed to the palladium degradation observed in AGR-1 UCO fuel. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Simplifying the Advanced Test Reactor LOWE Element Test Series

The Advanced Test Reactor (ATR) is one of six remaining high performance research reactors in the United States scheduled to be converted from 93% high-enriched uranium (HEU) fuel to 19.75% low-enriched uranium (LEU) fuel as part of the Department of Energy (DOE), National Nuclear Security Administration (NNSA), Office of Material Management and Minimization (MMM) reactor conversion efforts(1). The specific LEU fuel element design for the ATR is called the Low Enriched (LOWE) element. A series of fuel qualification tests are scheduled to occur over the next decade before the ATR is fully converted. In addition to several experiments of demonstration plates, a series of Element Tests (ETs) are planned for conversion, in which full sized LOWE elements are placed in ATR driver positions. The purpose of these ETs was generally to sequentially increase the amount of LEU and core power to full power, therefore creating a representative safety profile in which LEU could operate for the duration of the ATR lifetime. The planned element tests were (2): ET-1: one LOWE element at low power for once cycle ET-2: ~8 LOWE elements at low/medium power for multiple cycles, and ET-3: >8 LOWE elements for a full lobe of elements at high power for multiple cycles. Given recent improvements in modeling fidelity, scheduling considerations, and an opportunity to combine later ETs, the LEU conversion program successfully defined the operational requirements for the “ET-ATR” test, which combines the needed information collected from ET-2 and ET-3 into a single test.

42 ENGINEERING↗
DOE OSTI2022

Scoping Study of Effects on the Advanced Test Reactor’s Center Flux Trap Due to Various Loadings in the H and Inner A Test Positions

This study determined the effects experienced by the In-Pile Tube, of the Advanced Test Reactor’s center flux trap, due to various loading configurations of the H and Inner A positions. The effects of concern included: the axial profile of the linear heat generation rate of each fuel rodlet, the axial flux shape, and spectrum, in the fuel, and cladding, of each rodlet, and the total fission heating rate of the entire In-Pile Tube. Additionally, the reactivity worth of each configuration, relative to an ATR configuration with water filling all concerning test positions, was determined. The test positions were filled with commonly used test specimens, backup test specimens, and homogeneous substances. The study was performed using the modeling software MCNP6.2.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Experimental test of entangled histories

Entangled histories arise when a system partially decoheres in such a way that its past cannot be described by a sequence of states, but rather a superposition of sequences of states. Such entangled histories have not been previously observed. We propose and demonstrate the first experimental scheme to create entangled history states of the Greenberger–Horne–Zeilinger (GHZ) type. In our experiment, the polarization states of a single photon at three different times are prepared as a GHZ entangled history state. We define a GHZ functional which attains a maximum value 1 on the ideal GHZ entangled history state and is bounded above by 1∕16 for any three-time history state lacking tripartite entanglement. We have measured the GHZ functional on a state we have prepared experimentally, yielding a value of 0.656±0.005, clearly demonstrating the contribution of entangled histories.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2015

Water-gas shift: Characterization and testing of nanoscale YSZ supported Pt catalysts

Nano-scale Y-doped zirconium oxide materials were prepared with high surface areas (150-200 m(2)/g) and small nano-crystallites (<8 nm). A combination of XANES and EXAFS was used to show that ZrO2 exhibited the tetragonal phase, while the Zr0.5Y0.5O1.75 support displayed the cubic phase. A comparison with undoped zirconia suggests that the Zr0.9Y0.1O1.95 support was tetragonal in structure. A slight increase in d-spacing observed in HR-TEM for the Zr0.9Y0.1O1.95 support relative to undoped ZrO2, along with a shift to lower 2 theta in XRD, provide evidence that Y-doping caused macrostrain. STEM imaging confirmed that the Pt clusters ranged from 0.5 to 2 nm over all three supports. Catalyst reducibility was explored by H-2-TPR, XANES at the Zr K-edge, and TPR-XANES at the Pt L-III edge. A higher concentration of surface defects for the 0.5%Pt/Zr0.9Y0.1O1.95 catalyst relative to 0.5%Pt/ZrO2 was confirmed by DRIFTS of adsorbed CO, while a greater surface mobility of surface formate was suggested based on forward formate decomposition experiments in steam. The Y-doped Pt promoted catalysts displayed higher water-gas-shift activity relative to the 0.5%Pt/ZrO2 catalyst when the Y content was at or below 50%, with the best catalyst being 0.5%Pt/Zr0.9Y0.1O1.95. (C) 2015 Elsevier B.V. All rights reserved.

Pt/YSZ↗
DOE OSTI2025

Demonstration Test Report: Vacuum Thermal Desorption (VTD) Testing to Support West Area Waste Feed Test 2

Approximately 54 million gallons of radioactive and hazardous waste are stored in underground storage tanks at the U.S. Department of Energy’s (DOE’s) Hanford Site in southeastern Washington State. The Hanford Waste Treatment and Immobilization Plant (WTP) is being constructed to immobilize the waste in glass waste forms. The baseline configuration of the WTP includes a Pretreatment Facility to separate the wastes into a small volume of high-level waste (HLW) containing most of the radioactivity and a larger volume of low-activity waste (LAW) containing most of the non-radioactive chemicals. The HLW will be converted to glass in the HLW vitrification facility for ultimate disposal at a federal repository.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗