Search NASA⌕ Search

SEARCH · Search NASA

Results for “Material Testing”

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 1,117 records · Page 62
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Designing for Big Area Additive Manufacturing

Additive manufacturing (AM), more commonly referred to as 3D printing, is revolutionizing the manufacturing industry. With any new technology comes new rules and guidelines for the optimal use of said technology. Big Area Additive Manufacturing (BAAM), developed by Cincinnati Incorporated and Oak Ridge National Laboratory’s Manufacturing Demonstration Facility, requires a host of new design parameters compared to small-scale 3D printing to create large-scale parts. However, BAAM also creates new possibilities in material testing and various applications in the manufacturing industry. Most of the design constraints of small-scale polymer 3D printers still apply to BAAM. Beyond those constraints, new rules and limitations exist because BAAM’s large-scale system significantly changes the thermal properties associated with small-scale AM. Here, this work details both physical and software-related design considerations for additive manufacturing. After reading this guide, one will have a better understanding of slicing software’s capabilities and limitations, different physical characteristics of design and how to apply them appropriately for AM, and how to take the inherent nature of AM into consideration during the design process.

36 MATERIALS SCIENCE↗
DOE OSTI2017

Fate and transport of hexahydro-1,3,5-trinitro-1,3,5-triazine (RDX) and its degradation products in sedimentary and volcanic rocks, Los Alamos, New Mexico

High-explosive compounds including hexahydro-1,3,5-trinitro-1,3,5-triazine (RDX) were used extensively in weapons research and testing at Los Alamos National Laboratory (LANL). Liquid effluents containing RDX were released to an outfall pond that flowed to Cañon de Valle at LANL's Technical Area 16 (TA-16), resulting in the contamination of the alluvial, intermediate and regional groundwater bodies. Monitoring of groundwater within Cañon de Valle has shown persistent RDX in the intermediate perched zone located between 225 and 311 m below ground surface. Monitoring data also show detectable levels of RDX putative anaerobic degradation products. Batch and column experiments were conducted to determine the extent of adsorption-desorption and transport of RDX and its degradation products (MNX, DNX, and TNX) in major rock types that are within the RDX plume. All experiments in this paper were performed in the dark using water obtained from a well located at the center of the plume, which is fairly oxic and has a neutral pH of 7.5. Retardation factors and partitioning coefficient (K d ) values for RDX were calculated from batch experiments. Additionally, retardation factors and K d values for RDX and its degradation products were calibrated from column experiments using a one-dimensional transport model with equilibrium sorption (linear isotherm). Results from the column and batch experiments showed little to no sorption of RDX to the aquifer materials tested, with retardation factors ranging from 1.0 to 1.8 and K d values varying from 0 to 0.70 L/kg. Finally, results also showed no measurable differences between the transport properties of RDX and its degradation products.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2016

Deformation and fracture of explosion-welded Ti/Al plates: A synchrotron-based study

Here, explosion-welded Ti/Al plates are characterized with energy dispersive spectroscopy and x-ray computed tomography, and exhibit smooth, well-jointed, interface. We perform dynamic and quasi-static uniaxial tension experiments on Ti/Al with the loading direction either perpendicular or parallel to the Ti/Al interface, using a mini split Hopkinson tension bar and a material testing system in conjunction with time-resolved synchrotron x-ray imaging. X-ray imaging and strain-field mapping reveal different deformation mechanisms responsible for anisotropic bulk-scale responses, including yield strength, ductility and rate sensitivity. Deformation and fracture are achieved predominantly in Al layer for perpendicular loading, but both Ti and Al layers as well as the interface play a role for parallel loading. The rate sensitivity of Ti/Al follows those of the constituent metals. For perpendicular loading, single deformation band develops in Al layer under quasi-static loading, while multiple deformation bands nucleate simultaneously under dynamic loading, leading to a higher dynamic fracture strain. For parallel loading, the interface impedes the growth of deformation and results in increased ductility of Ti/Al under quasi-static loading, while interface fracture occurs under dynamic loading due to the disparity in Poisson's contraction.

36 MATERIALS SCIENCE↗
DOE OSTI2014

A cellular automaton method to simulate the microstructure and evolution of low-enriched uranium (LEU) U–Mo/Al dispersion type fuel plates

Low-enriched uranium (LEU) fuel plates for high power materials test reactors (MTR) are composed of nominally spherical uranium-molybdenum (U-Mo) particles within an aluminum matrix. Fresh U-Mo particles typically range between 10 and 100 mu m in diameter, with particle volume fractions up to 50%. As the fuel ages, reaction-diffusion processes cause the formation and growth of interaction layers that surround the fuel particles. The growth rate depends upon the temperature and radiation environment. The cellular automaton algorithm described in this paper can synthesize realistic random fuel-particle structures and simulate the growth of the intermetallic interaction layers. Examples in the present paper pack approximately 1000 particles into three-dimensional rectangular fuel structures that are approximately 1 mm on each side. The computational approach is designed to yield synthetic microstructures consistent with images from actual fuel plates and is validated by comparison with empirical data on actual fuel plates. (C) 2014 Elsevier B.V. All rights reserved.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Neutron energy spectrum influence on irradiation hardening and microstructural development of tungsten

We performed a neutron irradiation to single crystal pure tungsten in the mixed spectrum High Flux Isotope Reactor (HFIR). In order to investigate the influences of neutron energy spectrum, the microstructure and irradiation hardening were compared with previous data obtained from the irradiation campaigns in the mixed spectrum Japan Material Testing Reactor (JMTR) and the sodium-cooled fast reactor Joyo. The irradiation temperatures were in the range of ~90–~800 °C and fast neutron fluences were 0.02–9.00 × 10 25 n/m 2 (E > 0.1 MeV). Post irradiation evaluation included Vickers hardness measurements and transmission electron microscopy. Moreover, the hardness and microstructure changes exhibited a clear dependence on the neutron energy spectrum. The hardness appeared to increase with increasing thermal neutron flux when fast fluence exceeds 1 × 10 25 n/m 2 (E > 0.1 MeV). Finally, irradiation induced precipitates considered to be χ- and σ-phases were observed in samples irradiated to >1 × 10 25 n/m 2 (E > 0.1 MeV), which were pronounced at high dose and due to the very high thermal neutron flux of HFIR. Although the irradiation hardening mainly caused by defects clusters in a low dose regime, the transmutation-induced precipitation appeared to impose additional significant hardening of the tungsten.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Effects of Sodium Exposure on the Tensile Properties of Grade 91 steel

Two heats of G91 steel have been investigated in liquid sodium at 550- 650°C to understand their corrosion behaviour, microstructural evolution, and tensile properties. Sodium exposure experiments of subsized, sheet-type tensile specimens of G91 were performed in Argonne’s forced convection sodium materials testing loops, SMT-1 and SMT-2. Maximum exposure times of ~49,000 h at ~38,000 h, and ~20,000 h have been achieved at 550, 600, and 650°C, respectively. Thermal aging experiments of G91 steel were also conducted in parallel, and the results were compared with those of the sodium-exposed specimens to isolate the effect of thermal aging from the effect of sodium exposure. It was found that the yield stress and ultimate tensile strength of G91 were reduced by more than 50% after sodium exposures at 650°C due to decarburization. The reduction in tensile strength was accompanied by the dissolution of M 23 C 6 carbides and excessive grain growth. The effect of sodium at 550 and 600°C was insignificant for exposure times up to ~49,000 h and ~38,000 h, respectively. The effect of sodium on the microstructural stability and tensile properties of G91 is largely dependent on the carbon mass transfer in the alloy-sodium system.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

WIRE-21 Sensor Irradiation Experiment Ready for HFIR Insertion

The ability to deploy new nuclear fuels for current or future reactor concepts requires a wealth of data regarding fuel performance during normal operation, anticipated operational occurrences, and design-basis accidents. Most of these data have historically been collected during experiments in materials test reactors, ideally with online instrumentation to collect as much data as possible. However, advanced instrumentation could also allow for in situ monitoring of fuel operating conditions during commercial reactor operation to maximize fuel utilization, reduce unnecessary conservativism in design margins, and improve operator understanding of limiting peaking factors. The latter approach would complicate fuel handling, particularly during refueling, unless the instrumentation could be placed inside the fuel rods and transmitted wirelessly to a receiver located outside the fuel’s primary pressure boundary. To this end, Westinghouse Electric Company (WEC) developed wireless sensors based on inductive coupling that can transmit information regarding fuel centerline temperatures and rod internal pressures wirelessly from within a fuel rod to a nearby instrument thimble. After testing these sensors in lower-power university research reactors, the next step is to perform high neutron fluence testing to characterize the performance of these wireless sensors under conditions that are more representative of the intended application—in this case, light-water reactors (LWRs). The removable Be (RB) positions of the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) provide the neutron flux, experiment volume, and access to instrument leads required to achieve these sensor testing goals. This report summarizes the design, analysis, and assembly of the Wireless Instrumented RB Experiment 2021 (WIRE-21). This is the most highly instrumented irradiation experiment ever performed in HFIR. The experiment will use seven different sensing techniques to measure temperature, pressure, neutron flux, and neutron fluence during reactor operation. In addition to WEC’s wireless temperature and pressure sensors, WIRE-21 includes an array of thermocouples, self-powered neutron detectors, spatially distributed fiber optic temperature sensors, passive SiC temperature monitors, and flux wires. The design of WIRE-21 and the cabling that was installed in HFIR also provide the infrastructure to enable accelerated, economical testing of advanced sensor technologies while leveraging the extremely high neutron flux that is available in HFIR. The containment for WIRE-21 is similar to previous RB irradiation vehicles but includes a few modifications, most notably the use of integrated compression seals to pass a larger number of sensor leads through the experiment’s pressure boundary. In addition to the sensor leads, inert gas lines are passed into the experiment to enable active temperature control and the ability to pneumatically actuate a bellows-driven pressure sensor. WIRE-21 is targeting component temperatures (300–350°C) and neutron fluence levels (~10 22 n/cm 2 ) that would be expected in the plenum region of LWR fuels, except for the active sensing region of the wireless temperature sensor, which is targeting LWR fuel centerline temperatures (~800–1,100°C). WIRE-21 was successfully assembled, passed all nondestructive examination, and was delivered to HFIR for insertion during upcoming cycle 498 (April 2022).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Power transient analyses of experimental in-reflector devices during safety shutdown in Jules Horowitz Reactor (JHR)

The Jules Horowitz Reactor (JHR) is designed to be a 100 MW material testing reactor (MTR) and it is expected to become the reference facility in the framework of European nuclear research activity. As the core neutron spectrum is quite fast, several experimental devices concerning fuel studies have been conceived to be placed in the reflector in order to exploit a proper thermal neutron flux irradiation. Since the core power is relatively high, the neutronic coupling between the reactor core and the reflector devices has to be taken into account for different rod insertions. In fact the thermal power produced within the fuel samples is considerable. Heat removal during shutdown is a main topic in nuclear safety and it is worth to analyse thermal power transients in fuel samples as well. Here a thermal hydraulic model for JHR core is proposed aiming at a simple and representative description as far as reactivity feedbacks are concerned. Then it is coupled with a neutronic pointwise kinetics analysis by means of the DULCINEE code to compute core power transient calculations. Moreover, some reflector-core coupling evaluations are performed through Monte Carlo method using the TRIPOLI 4.7 code. The JHR equilibrium cycle is considered with respect to four fuel compositions namely Beginning of Cycle (BOC), Xenon Saturation Point (XSP), Middle of Cycle (MOC) and End of Cycle (EOC). Then thermal power transients in the experimental reflector devices are evaluated during safety shutdowns and they are verified for all these cycle steps. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS2023

Uncertainty Analysis of Slug Calorimeters in the HyMETS Arc-Jet Facility

The objective of this work is to perform an uncertainty analysis of the deduced stagnation heat flux environment on a slug calorimeter for conditions that span the performance envelope of the Hypersonic Materials Environmental Test System arc-jet facility located at NASA Langley Research Center. Analytical solutions are developed for boundary-value problems on the slug element accounting for non-ideal effects, including spatial variation in the slug heat flux, multi-dimensional thermal conduction, and back-face losses, which departs from the state-of-the-art method derived from the American Society of Testing and Materials. Boundary-value problem definitions are informed by preliminary finite element thermal analysis of the slug calorimeter assembly (including both slug and housing) and just the slug element. The analytical solutions are presented in a general sense and in a truncated form from error analysis. Results are shown in optimizing and validating the analytical models against available slug back-face thermal data. The optimization results indicate that the appropriate epistemic uncertainty of the deduced stagnation heat flux on the slug calorimeter is at most±2.5% for both a high-and low-enthalpy test condition. In addition, a numerical approach is used to determine the aleatory (probabilistic) uncertainty component in the slug stagnation heat flux by applying a marching least-squares slope routine through the steady-state portion of the slug back-face thermal response. Results indicate a compromise between the number of samples and the filter frequency of slug back-face thermal data points when evaluating the standard deviation of the deduced stagnation heat flux statistics. When combining the mixed uncertainty, both aleatory and epistemic, the interval of uncertainty in the deduced stagnation heat flux is determined to be up to ±4%, which is at least a 60% reduction from the standard uncertainty used in the state-of-the-art method.

uncertainty↗
DOE OSTI2012

A Multi-Methods Approach to HRA and Human Performance Modeling: A Field Assessment

The Advanced Test Reactor (ATR) is a research reactor at the Idaho National Laboratory is primarily designed and used to test materials to be used in other, larger-scale and prototype reactors. The reactor offers various specialized systems and allows certain experiments to be run at their own temperature and pressure. The ATR Canal temporarily stores completed experiments and used fuel. It also has facilities to conduct underwater operations such as experiment examination or removal. In reviewing the ATR safety basis, a number of concerns were identified involving the ATR canal. A brief study identified ergonomic issues involving the manual handling of fuel elements in the canal that may increase the probability of human error and possible unwanted acute physical outcomes to the operator. In response to this concern, that refined the previous HRA scoping analysis by determining the probability of the inadvertent exposure of a fuel element to the air during fuel movement and inspection was conducted. The HRA analysis employed the SPAR-H method and was supplemented by information gained from a detailed analysis of the fuel inspection and transfer tasks. This latter analysis included ergonomics, work cycles, task duration, and workload imposed by tool and workplace characteristics, personal protective clothing, and operational practices that have the potential to increase physical and mental workload. Part of this analysis consisted of NASA-TLX analyses, combined with operational sequence analysis, computational human performance analysis (CHPA), and 3D graphical modeling to determine task failures and precursors to such failures that have safety implications. Experience in applying multiple analysis techniques in support of HRA methods is discussed.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Microstructure of Neutron-Irradiated Al 3 Hf-Al Thermal Neutron Absorber Materials

A thermal neutron-absorbing metal matrix composite (MMC) comprised of Al 3 Hf particles in an aluminum matrix was developed to filter out thermal neutrons and create a fast flux environment for material testing in a mixed-spectrum nuclear reactor. Intermetallic Al 3 Hf particles capture thermal neutrons and are embedded in a highly conductive aluminum matrix that provides conductive cooling of the heat generated due to thermal neutron capture by the hafnium. These Al 3 Hf-Al MMCs were fabricated using powder metallurgy via hot pressing. The specimens were neutron-irradiated to between 1.12 and 5.38 dpa and temperatures ranging from 286 °C to 400 °C. The post-irradiation examination included microstructure characterization using transmission electron microscopy (TEM) and energy-dispersive X-ray spectroscopy. This study reports the microstructural observations of four irradiated samples and one unirradiated control sample. All the samples showed the presence of oxide at the particle–matrix interface. The irradiated specimens revealed needle-like structures that extended from the surface of the Al 3 Hf particles into the Al matrix. An automated segmentation tool was implemented based on a YOLO11 computer vision-based approach to identify dislocation lines and loops in TEM images of the irradiated Al-Al 3 Hf MMCs. This work provides insight into the microstructural stability of Al 3 Hf-Al MMCs under irradiation, supporting their consideration as a novel neutron absorber that enables advanced spectral tailoring.

36 MATERIALS SCIENCE↗
NASA NTRSDate not supplied

NASA-GRC High Voltage Materials Development and Test Capabilities Portfolio

This presentation provides the background information on NASA-GRC high voltage (HV) materials team research efforts towards electrified propulsion systems since 2016 . Additionally, it covers polymer and ceramic filler materials development for HV electrical insulation composites, copper/ carbon nanotube hybrid conductors, modeling efforts, HV test capabilities, and future material processing capabilities.

Boron Nitride↗
DOE OSTI2020

Evaluation of Changes to the MTR Fuel Dissolution Flowsheet to Increase the Number of L-Bundles Charged to a Dissolver

Material Test Reactor (MTR) fuels are dissolved in the H-Canyon facility using a Hg-catalyzed, HNO₃ dissolution flowsheet. The flowsheet is based on experimental work in which laboratory-scale experiments were used to provide bounding values for both offgas and H₂ generation rates. The generation of H₂ gas, primarily from the dissolution of Al, is controlled by limiting the surface area exposed to the solution. The number of L-Bundles charged and their submergence level defines the exposed surface area which is directly proportional to the H₂ generation rate. For the MTR fuel dissolution flowsheet, the maximum number of L-Bundles which can be charged to a dissolver was specified as a function of the Al concentration in the solution and the submergence level of the bundle. The MTR fuels are currently dissolved in the 6.1D dissolver. To reach the desired terminal Al concentration, a total of 12 L-Bundles of fuel are dissolved per batch. Depending on the submergence level being used in the dissolver, the 12 bundles are dissolved in three charges of four bundles per charge or in a sequence of three, four, and five bundles. However, if six L-Bundles could be dissolved in the initial charge, a dissolver batch could be completed in two dissolutions instead of three. To assess the feasibility of dissolving six bundles in the first charge, the Savannah River National Laboratory evaluated changes to the MTR flowsheet which increase the number of bundles which can be charged to a dissolver. Since the development of the existing MTR fuel dissolution flowsheet, improved offgas measurement methods have been developed to more accurately calculate the offgas composition and generation rate during laboratory-scale dissolutions. In the first part of this study, improvements in the calibration model for the Raman spectrometer (used for offgas characterization) and a more effective data smoothing routine were applied to existing data. The uncertainty in the offgas composition data used to define the MTR fuel dissolution flowsheet were subsequently re-evaluated which resulted in an approximate 50% reduction in the uncertainty of the H₂ gas concentration measurements. The reduced uncertainty was used in new calculations to define the maximum number of L-Bundles which can be charged to either the 6.1D or 6.4D dissolver as a function of the Al concentration. With the reduced uncertainty in the H₂ concentration measurements, a flowsheet was defined which supported initially charging 6 bundles to a dissolver. This flowsheet is based on the use of 0.002 M Hg, an L-Bundle immersion depth of 54 in., an air sparge/purge flowrate of 40 scfm, an iodine reactor operating temperature of 200 °C, and not exceeding 60% of the H₂ LFL in the dissolver offgas. In the second phase of this study, calculations were performed to assess the impact of changing various operating parameters to increase the number of L-Bundles charged to an H-Canyon dissolver. Applying the improved offgas measurement techniques including the reduced uncertainty in the H₂ concentration measurements to the existing data, the number of bundles of MTR fuel which can be charged to either the 6.1D or 6.4D dissolver were calculated based on varying the bundle immersion depth (44-60 in.), air sparge/purge flowrate (40-60 scfm), iodine reactor operating temperature (150-200 °C), and the percentage of the H₂ LFL which must not be exceeded (60-76%). Qualitatively, the percentage of the LFL had the largest influence on the maximum number of bundles which can be charged to a dissolver followed by the air sparge/purge flowrate, L-Bundle immersion depth, and the iodine reactor operating temperature. During the evaluation of each variable, the other variables were held at the baseline flowsheet conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Investigation Into LiCl-LiF+Li 2 O as a Low Temperature Electrolyte in Direct Electrolytic Reduction of UO 2

Direct electrolytic reduction (DER) of UO 2 in a binary LiCl-Li2O (1–3 wt%) at 650 °C has been widely reported. The process temperature can be reduced to 550 °C in theory by using a near eutectic ternary LiCl-LiF+Li 2O (∼2 wt%) electrolyte, which will result in substantial reduction in rate of salt vaporization. The feasibility of using the ternary electrolyte for the process was tested via material compatibility testing, electrochemical polarization testing, and controlled potential or current DER experiments. MgO, Pt, and 625 nickel samples were each contacted with the ternary salt for 168 h. Negligible changes to dimensions and mass of the samples were measured, and concentrations of corrosion products in the salt were 102 ppm or less. O2−oxidation was observed to occur at a lower potential than Cl-or F-on a Pt anode. DER of UO2was performed in six trials. Process variables included type of reference electrode, electrochemical control method, type of cathode basket, and electrolyte composition. UO 2conversion ranged from 28.2 to 99.9 % , and current efficiency ranged from 38 to 52%. Overall, results indicate that the use of the ternary salt is feasible for DER.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Achieving Atomic Dispersion of Highly Loaded Transition Metals in Small‐Pore Zeolite SSZ‐13: High‐Capacity and High‐Efficiency Low‐Temperature CO and Passive NO x Adsorbers

Abstract The majority of harmful atmospheric CO and NO x emissions are from vehicle exhausts. Although there has been success addressing NO x emissions at temperatures above 250 °C with selective catalytic reduction technology, emissions during vehicle cold start (when the temperature is below 150 °C), are a major challenge. Herein, we show we can completely eliminate both CO and NO x emissions simultaneously under realistic exhaust flow, using a highly loaded (2 wt %) atomically dispersed palladium in the extra‐framework positions of the small‐pore chabazite material as a CO and passive NO x adsorber. Until now, atomically dispersed highly loaded (>0.3 wt %) transition‐metal/SSZ‐13 materials have not been known. We devised a general, simple, and scalable route to prepare such materials for Pt II and Pd II . Through spectroscopy and materials testing we show that both CO and NO x can be simultaneously completely abated with 100 % efficiency by the formation of mixed carbonyl‐nitrosyl palladium complex in chabazite micropore.

Khivantsev, Konstantin↗
DOE OSTI2021

Quantifying adherence of oxide scales on steels exposed to high temperature and pressure steam

Oxide scale exfoliation is a major concern in fossil fuel power generation because it can cause tube blockages and erode valves and steam turbine components downstream. There is still considerable scientific and commercial interest to improve the mechanistic understanding of oxide failures by developing models to predict exfoliation and the extent of tube blockage as a function of operating conditions and component geometries. Tensile testing inside a scanning electron microscope was conducted on ferritic–martensitic and austenitic steel specimens with the steam side (Fe,Cr)-rich oxides grown after exposures for up to 1000 h in steam with ~100 ppb O 2 at 276 bar and 550°C. Multiple oxide layer cracks and delamination events were observed and analyzed in detail during the tests. Results from the testing agreed well with earlier observations that had identified the failure location at the outer–inner oxide layer for all tested materials. Calculated adhesion energies identified the outer–inner oxide interface of alloy 347HFG as the weakest interface.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Corrosion evaluation of Ti–6Al–4V manufactured by electron beam melting in Ringer’s physiological solution: an in vitro study of the passive film

We report that electrochemical characteristics and semiconducting behavior of additively manufactured electron beam melted (EBM) and wrought (WR) Ti–6Al–4V (Ti-G5) are compared in Ringer’s physiological solution. X-ray diffraction (XRD) and field emission scanning electron microscopy (FE-SEM) confirmed the α+β structure of the tested materials, with two different microstructure types of “bimodal” and “basket-weave” for WR and EBM, respectively. Potentiodynamic polarization (PDP) revealed that the corrosion current density for EBM (i corr =0.27±0.06 μA cm -2 ) is less than the WR (i corr =0.70±0.05 μA cm -2 ). Moreover, potentiostatic polarization (PS) that was employed to form the passive layers at three different potentials of 300, 500, and 700 mV Ag/AgCl , showed that the passive films on the EBM sample are thinner. This finding was confirmed by electrochemical impedance spectroscopy (EIS). Furthermore, through Mott–Schottky (M–S) analysis, donor densities on WR passive films were found to be ~1.5 times larger than EBM. Although PS and EIS confirmed that the passive layer on EBM is thinner, it provides higher corrosion resistance than WR. The passive layer on both samples were found to have n-type characteristics with a duplex structure.

36 MATERIALS SCIENCE↗
DOE OSTI2016

The Preparation of Non-radioactive Glassy Surrogate Nuclear Explosion Debris (SNED) Loaded with an Isotopically Altered Noble Gas

The measurement of the Kr and Xe isotope ratios in nuclear explosion debris is an important forensic tool as the method requires very little sample preparation. Fragments of debris are simply crushed or heated to release the trapped gases Kr and Xe arising from fission product decay. As suitable test materials are not readily available we have been investigating a method to incorporate isotopically enriched 129Xe in glassy materials that mimic nuclear explosion debris. The approach used to prepare these materials will be described along with some of the example results obtained.

Trinitite, Nuclear forensics, Fission Xe, Surrogat↗