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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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All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Development and characterization of an aerosol-generated multi-method analytical particle test material

Ceria (CeO 2 ) particles with low to ultra-low loading of nickel dopant were produced using an aerosol-based, droplet-to-particle synthesis via an in-line calcination technique. This aerosol-based synthesis method enables the production of particles with a monodisperse size distribution. These produced and well-characterized, multi-element, ceria-based particles demonstrate a material exemplar for multi-method analytical testing. They were prepared from a cerium nitrate feedstock where low loading nickel dopant was added at target Ni/(Ni + Ce) atomic percents of 1 %, 0.1 %, and 0.01 %, using a nickel nitrate spike. This methodology proved to produce ceria particles doped with a dynamic range of low to ultra-low loadings of nickel over a 24-h period, with consistent size distribution, morphology, and composition. The successful incorporation of nickel was demonstrated with bulk and single particle inductively coupled plasma mass spectroscopy and revealed notable particle-to-particle elemental homogeneity. X-ray photoelectron spectroscopy demonstrated the presence of a high concentration of nickel dopant incorporated preferentially toward the surface of the particles, and that this dopant aided oxidation of surface Ce(III) atoms to Ce(IV). These particle test materials were then validated through X-ray absorption near edge spectroscopy, comparing the ultra-low 0.01 % Ni and low 1 % Ni-doped ceria samples. This revealed a more-reduced oxidation state of the nickel with an increase in dopant concentration. Finally, this work demonstrates a synthesis and systematic characterization scheme to produce multi-method analytical test particulates.

36 MATERIALS SCIENCE↗
DOE OSTI2010

Cryogenic RF Material Testing with a High-Q Copper Cavity

An X-band RF cryogenic material testing system has been developed in the past few years. This system employs a high-Q copper cavity with an interchangeable flat bottom working under a TE{sub 013} like mode. By measuring the cavity Qs with a network analyzer, the system can characterize the surface resistance of different samples at different temperatures. Using a 50 MW 2{mu}s pulsed klystron, the system can measure the quenching H field for superconducting samples, up to 300-400 mT. In this paper, we will present the most recent developments of the system and testing results.

36 MATERIALS SCIENCE↗
DOE OSTI2015

Impact Study on the Methodology used for Photon- Heating Determination in Material-Testing Reactors

Determination of photon heating by calculation is an important issue for the Jules Horowitz Reactor (JHR), the next international Material-Testing Reactor (MTR) under construction at CEA Cadarache research center in the south of France. Accurate knowledge of photon heating in structure materials and irradiation devices is necessary for JHR design and safety studies. In this paper, we quantify the impact of different photon-heating calculation routes by comparing absorbed dose and KERMA calculations (Kinetic Energy Released per Mass) from two different Monte Carlo codes, TRIPOLI-4.9{sup R} and MCNP (Monte Carlo N-Particle transport code). These calculations are carried out in JHR-representative geometries with the nuclear-data library JEFF3.1.1 and the photon-data library EPDL97. Discrepancies amounting to up to 18% between absorbed dose and KERMA are found in JHR irradiation devices and are linked to charged-particle transport effects taking place in heterogeneous materials of small dimensions. In a JHR assembly cell, discrepancies of about 1% on photon KERMA and of about 3% on absorbed dose are highlighted between the two Monte Carlo codes. These latter discrepancies are small compared to typical sources of uncertainty for Monte Carlo calculation (for instance, nuclear data uncertainty) and are supposed to be due to differences in the processing of gamma-production data by neutron interactions and to differences in electromagnetic-shower models and implementation between the two codes. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Report on the completion of long-lead components procurement of new sodium materials test loop

This report provides an update on the completion of long-lead components procurement of new sodium materials test loop. The loop, designated as SMT-3, is a forced-convection sodium loop to be used for the evaluation of sodium compatibility of advanced Alloy 709 with emphasis on long term exposures of tensile, creep, fatigue, creep fatigue, and fracture toughness ASTM-size specimens in support of ASME Code qualification and NRC licensing. The report is a deliverable (Level 4) in FY18 (M4NT-18AN050503013), under the Work Package NT-18AN05050301, “A709 Sodium Compatibility” performed by Argonne National Laboratory (ANL), as part of the Advanced Materials Program for the Advanced Reactor Technology. This work package enables the development of advanced structural materials by providing corrosion, microstructure, and mechanical property data from the standpoint of sodium compatibility of advanced structural alloys. The first sodium loop (SMT-1) with a single tank was constructed in 2011 at ANL and has been in operation for exposure of subsize sheet specimens of advanced alloys at a single temperature. The second sodium loop with dual tanks (SMT-2) was constructed in 2013 and has been in operation for the exposure of subsize sheet specimens of advanced alloys at two different temperatures. The current loop (SMT-3) has been designed to incorporate sufficient chamber capacity to expose a large number of ASTM-size specimens to evaluate the sodium effects on tensile, creep, fatigue, creep-fatigue, and fracture toughness properties, in support of ASME Code Qualification and USNRC Licensing. Three major long lead items were identified: 1) Selection and approval of location for the new loop and installation of the necessary infrastructure, such as transformer, power supply panels, and wiring, that is adequate for the long term operation of the loop, 2) Perform a detailed piping analysis that meets the ANL QA and Safety guidelines as well as complying with ASME Code requirements, and 3) Design and fabrication of the specimen exposure vessels that are capable of at least 42 tensile, creep, and fatigue specimens in each vessel and is compliant and compatible with ASME Code and the piping design analysis. The SMT-3 loop will be located in Building 206 adjacent to the currently operating SMT-2 loop. In addition, we have removed the aged power supply system in Building 206 and installed a new transformer, wiring, and power panels for the new loop. The preliminary loop design was sent to an outside vendor for a detailed piping analysis, to ensure compliance with the ASME Code guidelines. The piping design was completed in FY2017. The construction of components such as cold trap, economizers, piping arrangement etc. will be performed in the central shops at ANL. About 150 liters of sodium for the loop will be procured in early FY2018. The loop system is designed to circulate sodium through the sample tanks and he associated loop without an operator for an extended period of time. With the three sodium loops (i.e., single-tank, dual-tank and four–tank), materials can be tested at different sodium temperatures, and large tensile, creep, fatigue, creep-fatigue, and fracture toughness specimens can be exposed to sodium for extended periods of time and generate data on mechanical properties in support of ASME Code Qualification and USNRC Licensing of advanced Alloy 709 for use as a structural material in SFRs.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2011

CALMOS: Innovative device for the measurement of nuclear heating in material testing reactors

An R and D program has been carried out since 2002 in order to improve gamma heating measurements in the 70 MWth OSIRIS Material Testing Reactor operated by CEA's Nuclear Energy Div. at the Saclay research center. Throughout this program an innovative calorimetric probe associated to a specific handling system has been designed in order to make measurements both along the fissile height and on the upper part of the core, where nuclear heating rates still remain high. Two mock-ups of the probe were manufactured and tested in 2005 and 2009 in ex-core area of OSIRIS reactor for the process validation, while a displacement system has been especially designed to move the probe axially. A final probe has been designed thanks to modeling results and to preliminary measurements obtained with mock-ups irradiated to a heating level of 2W/g, This paper gives an overview of the development, describes the calorimetric probe, and expected advantages such as the possibility to use complementary methods to get the nuclear heating measurement. Results obtained with mock-ups irradiated in ex-core area of the reactor are presented and discussed. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2026

Structure–Property Relationships of Recycled Lithium-Ion Battery Cathodes: Microstructure Optimization Using Virtual Materials Testing

The increasing demand for sustainable battery technologies requires effective recycling strategies for end-of-life lithium-ion battery cathodes. In this study, virtual materials testing, a well-established framework for modeling conventionally manufactured NMC-based cathodes, is applied to partially recycled cathodes. To this end, virtual cathodes consisting of mixtures of pristine and recycled NMC particles are utilized to systematically analyze structure–property relationships depending on mixing ratios and different spatial arrangement strategies. For this purpose, a stochastic 3D model is developed that is capable of generating virtual cathodes with arbitrary volume fractions of active materials and mixing ratios of pristine and recycled NMC particles. Particularly, the stochastic 3D model can mimic the different size distributions of pristine and recycled particles that are observed in image data. Additionally, the model allows the structuring of pristine and recycled NMC either uniformly mixed or layer-wise arranged, mimicking single- and dual-layer cathodes. Subsequently, a systematic computational analysis is conducted to assess the influence of increasing active material ratios of recycled particles, ranging from 0 % to 100 %, while maintaining a constant overall active material volume fraction. The impact of particle mixing on cathode performance is evaluated by examining transport-relevant geometrical descriptors and effective properties, such as geodesic tortuosity, specific surface area, and tortuosity factor.

25 ENERGY STORAGE↗
DOE OSTI2011

New measurement system for on line in core high-energy neutron flux monitoring in materials testing reactor conditions

Flux monitoring is of great interest for experimental studies in material testing reactors. Nowadays, only the thermal neutron flux can be monitored on line, e.g., using fission chambers or self-powered neutron detectors. In the framework of the Joint Instrumentation Laboratory between SCK-CEN and CEA, we have developed a fast neutron detector system (FNDS) capable of measuring on line the local high-energy neutron flux in fission reactor core and reflector locations. FNDS is based on fission chambers measurements in Campbelling mode. The system consists of two detectors, one detector being mainly sensitive to fast neutrons and the other one to thermal neutrons. On line data processing uses the CEA depletion code DARWIN in order to disentangle fast and thermal neutrons components, taking into account the isotopic evolution of the fissile deposit. The first results of FNDS experimental test in the BR2 reactor are presented in this paper. Several fission chambers have been irradiated up to a fluence of about 7 x 10{sup 20} n/cm{sup 2}. A good agreement (less than 10% discrepancy) was observed between FNDS fast flux estimation and reference flux measurement.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Analysis of 440 GeV proton beam–matter interaction experiments at the High Radiation Materials test facility at CERN

In a previous paper [Schmidt et al., Phys. Plasmas 21, 080701 (2014)], we presented the first results on beam–matter interaction experiments that were carried out at the High Radiation Materials test facility at CERN. In these experiments, extended cylindrical targets of solid copper were irradiated with beam of 440 GeV protons delivered by the Super Proton Synchrotron (SPS). The beam comprised of a large number of high intensity proton bunches, each bunch having a length of 0.5 ns with a 50 ns gap between two neighboring bunches, while the length of this entire bunch train was about 7 μs. These experiments established the existence of the hydrodynamic tunneling phenomenon the first time. Detailed numerical simulations of these experiments were also carried out which were reported in detail in another paper [Tahir et al., Phys. Rev. E 90, 063112 (2014)]. Excellent agreement was found between the experimental measurements and the simulation results that validate our previous simulations done using the Large Hadron Collider (LHC) beam of 7 TeV protons [Tahir et al., Phys. Rev. Spec. Top.--Accel. Beams 15, 051003 (2012)]. According to these simulations, the range of the full LHC proton beam and the hadronic shower can be increased by more than an order of magnitude due to the hydrodynamic tunneling, compared to that of a single proton. This effect is of considerable importance for the design of machine protection system for hadron accelerators such as SPS, LHC, and Future Circular Collider. Recently, using metal cutting technology, the targets used in these experiments have been dissected into finer pieces for visual and microscopic inspection in order to establish the precise penetration depth of the protons and the corresponding hadronic shower. This, we believe will be helpful in studying the very important phenomenon of hydrodynamic tunneling in a more quantitative manner. The details of this experimental work together with a comparison with the numerical simulations are presented in this paper.

36 MATERIALS SCIENCE↗
DOE OSTI2015

Thick Films acoustic sensors devoted to MTR environment measurements. Thick Films acoustic sensors devoted to Material Testing Reactor environment measurements

The development of advanced instrumentation for in-pile experiments in Material Testing Reactor constitutes a main goal for the improvement of the nuclear fuel behavior knowledge. An acoustic method for fission gas release detection was tested with success during a first experiment called REMORA 3 in 2010 and 2011, and the results were used to differentiate helium and fission gas release kinetics under transient operating conditions. This experiment was lead at OSIRIS reactor (CEA Saclay, France). The maximal temperature on the sensor during the irradiation was about 150 deg. C. In this paper we present a thick film transducer produce by screen printing process. The screen printing of piezoelectric offers a wide range of possible applications for the development of acoustic sensors and piezoelectric structure for measurements in high temperature environment. We firstly produced a Lead Zirconate Titanate (PZT) based paste composed of Pz27 powder from Ferroperm, CF7575 glass, and organic solvent ESL 400. Likewise a Bismuth Titanate based paste synthesized in our laboratory was produced. With these inks we produced thick film up to 130 μm by screen printing process. Material properties characterizations of these thick-film resonators are essential for device design and applications. The piezoelectric coefficients d33 and pyro-electric P(T) coefficient are investigated. The highest P(T) and d33 are respectively 80 μC.m{sup -2}.K{sup -1} and 130 μC.N{sup -1} for the PZT transducer -which validates the fabrication process-. In view of the development of this transducer oriented for high temperature and irradiation environment, we investigated the electrical properties of the transducers for different ranges of frequencies and temperature - from 20 Hz up to 40 MHz between 30 and 400 deg. C. We highlight the evolution of the impedance response and piezoelectric parameters of screen printed piezoelectric structures on alumina. Shortly an irradiation will be realized in order to investigate their reliability towards neutron fluence. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

A Cryogenic RF Material Testing Facility at SLAC

The authors have developed an X-band SRF testing system using a high-Q copper cavity with an interchangeable flat bottom for the testing of different materials. By measuring the Q of the cavity, the system is capable to characterize the quenching magnetic field of the superconducting samples at different power level and temperature, as well as the surface resistivity. This paper presents the most recent development of the system and testing results.

43 PARTICLE ACCELERATORS↗
DOE OSTI2022

Granta:MI Record Review Procedure for Materials Testing Projects

This document is intended to serve as a guide for the review of records in LANL’s Weapons Materials Database. Specifically, this guide provides step-by-step instructions for the review of records associated with a materials testing project. This is the type of dataset that testing organizations such as MST-7 & MST-8 at LANL commonly produce for the weapons complex and want to archive in Granta:MI.

45 MILITARY TECHNOLOGY, WEAPONRY, AND NATIONAL DEF↗
DOE OSTI2014

Adaptation of Crack Growth Detection Techniques to US Material Test Reactors

A key component in evaluating the ability of Light Water Reactors to operate beyond 60 years is characterizing the degradation of materials exposed to radiation and various water chemistries. Of particular concern is the response of reactor materials to Irradiation Assisted Stress Corrosion Cracking (IASCC). Some materials testing reactors (MTRs) outside the U.S., such as the Halden Boiling Water Reactor (HBWR), have deployed a technique to measure crack growth propagation during irradiation. This technique incorporates a compact loading mechanism to stress the specimen during irradiation. A crack in the specimen is monitored using the Direct Current Potential Drop (DCPD) method. A project is underway to develop and demonstrate the performance of a similar type of test rig for use in U.S. MTRs. The first year of this three year project was devoted to designing, analyzing, fabricating, and bench top testing a mechanism capable of applying a controlled stress to specimens while they are irradiated in a pressurized water loop (simulating PWR reactor conditions). During the second year, the mechanism will be tested in autoclaves containing high pressure, high temperature water with representative water chemistries. In addition, necessary documentation and safety reviews for testing in a reactor environment will be completed. In the third year, the assembly will be tested in the Massachusetts Institute of Technology Reactor (MITR) and Post Irradiation Examinations (PIE) will be performed.

36 MATERIALS SCIENCE↗
DOE OSTI2026

Overview of advanced plasma-facing materials testing for Fusion Pilot Plants at DIII-D

Characterization and testing of advanced plasma-facing materials (PFMs) for Fusion Pilot Plants (FPP) is being conducted at the DIII-D National Fusion Facility through the ongoing two-year FPP Candidate Materials Thrust. Year one tested 17 novel materials utilizing the Divertor Materials Evaluation System (DiMES), with samples analyzed pre- and post-experiment via SEM, EDS, and confocal microscopy. Repeatable reference discharges were developed to ensure uniformity between experiments, including a new strike-point rastering scenario to provide more uniform heat/particle flux across DiMES during ELMing H-mode discharges. Various sample geometries and temperatures were used to achieve FPP-relevant conditions, including samples angled 10° towards the incident plasma flux and pre-heating up to 500 °C. The first exposure of liquid lithium (Li) capillary porous structures in a tokamak demonstrated uniform emission of Li vapor and suppression of Li droplets in H-mode when preheated to 350 °C. Dispersoid-strengthened W with 1 wt% TaC, TiC, and ZrC exposed to H-mode showed cracking and dispersoid ejection for all varieties except TiC, providing a clear down-selection. Ultra-high temperature ceramic materials TiB 2 and ZrB 2 showed minimal degradation under L-mode exposure. Silicon carbide (SiC) fiber composites showed arcing along edges, while CVD SiC remained pristine. Atmospheric plasma-sprayed W and SiC coatings endured H-mode exposure without macroscopic delamination; SiC exhibited granular ejection, while W showed increased outgassing. Additional W-based alloys were stress tested in H-mode, including Ni-based W heavy alloys, W f SiC f /W composites, W multi-principle element alloys, and functionally-graded W/SiC, to varying degrees of success.

DIII-D↗
DOE OSTI2012

ORNL Automated-In-Motion Evaluation Environment (AIMVEE) User Training and Testing Materials - U.S. Copyright TXu 1-797-273

The AIMVEE/WIM system electronically retrieves deployment information, identifies vehicle automatically, and determines total weight, individual wheel weight, individual axle weights, axle spacing, and center-of-balance for any wheeled vehicle in motion. The AIMVEE/WIM system can also perform these functions statically for both wheel vehicles and cargo with information. The AIMVEE/WIM system incorporates digital images and applies cubing algorithms to determine length, width, height for cubic dimensions of both vehicle and cargo. Once all this information is stored, it electronically links to data collection and dissemination systems to provide actual weight and measurement information for planning, deployment, and in-transit visibility. The Static Scale Conversion (SSC) system is an unique enhancement to the AIMVEE/WIM system. It enables a SSC to weigh and measure vehicles and cargo dynamically (i.e., as they pass over the large scale and is included in the AIMVEE computer code base. The material copyrighted is the ORNL Automated-In-Motion Vehicle Evaluation Environment (AIMVEE)/Weigh-In-Motion User Training and Testing material. It includes instructional material in the set-up, operation and tear-down of the AIMVEE/WIM system. It also includes a final exam associated with the training.

Automated-in-Motion Evaluation Environment (AIMVEE↗
DOE OSTI2015

ADAPTATION OF CRACK GROWTH DETECTION TECHNIQUES TO US MATERIAL TEST REACTORS

A key component in evaluating the ability of Light Water Reactors to operate beyond 60 years is characterizing the degradation of materials exposed to radiation and various water chemistries. Of particular concern is the response of reactor materials to Irradiation Assisted Stress Corrosion Cracking (IASCC). Some test reactors outside the United States, such as the Halden Boiling Water Reactor (HBWR), have developed techniques to measure crack growth propagation during irradiation. The basic approach is to use a custom-designed compact loading mechanism to stress the specimen during irradiation, while the crack in the specimen is monitored in-situ using the Direct Current Potential Drop (DCPD) method. In 2012 the US Department of Energy commissioned the Idaho National Laboratory and the MIT Nuclear Reactor Laboratory (MIT NRL) to take the basic concepts developed at the HBWR and adapt them to a test rig capable of conducting in-pile IASCC tests in US Material Test Reactors. The first two and half years of the project consisted of designing and testing the loader mechanism, testing individual components of the in-pile rig and electronic support equipment, and autoclave testing of the rig design prior to insertion in the MIT Reactor. The load was applied to the specimen by means of a scissor like mechanism, actuated by a miniature metal bellows driven by pneumatic pressure and sized to fit within the small in-core irradiation volume. In addition to the loader design, technical challenges included developing robust connections to the specimen for the applied current and voltage measurements, appropriate ceramic insulating materials that can endure the LWR environment, dealing with the high electromagnetic noise environment of a reactor core at full power, and accommodating material property changes in the specimen, due primarily to fast neutron damage, which change the specimen resistance without additional crack growth. The project culminated with an in-pile demonstration at the MIT Reactor. The test rig and associated support equipment were used to apply loads to a representative Compact Tensile specimen during one MITR operating cycle, while measuring crack growth using the DCPD method. Although the test period was short (approximately 70 days), and the accumulated neutron dose relatively small, successful operation of the test rig was demonstrated. The specimen was cycled more than 8000 times (more than would be typical for a long term IASCC test), which was sufficient to propagate a crack of over 2 mm.

Crack Growth↗
DOE OSTI2025

Computational materials assessment of the D/Li-stripping neutron source as a prototypical facility for fusion materials testing

As the US fusion materials community awaits the selection and design of a fusion prototypical neutron source (FPNS), a risk reduction exercise has been conducted to (i) provide an updated materials performance evaluation using state-of-the-art computational materials modeling, (ii) expand on legacy analysis based on pure Fe to other relevant fusion structural materials types, and (iii) ensure that materials response under FPNS operational conditions is consistent with referential fusion reactor conditions. The current paper describes the efforts undertaken to assemble a comprehensive computational methodology that includes neutronics, primary damage calculations, atomistic simulations of displacement cascades, chemical inventory evolution calculations, and a computational thermodynamic analysis of emerging phases during irradiation. Our work extends existing studies in pure Fe to reduced-activation ferritic/martensitic steels, tungsten, silicon carbide, and vanadium alloys. We focus on the single-beam deuteron/lithium-stripping neutron source behind the IFMIF-DONES concept, which we assess against ITER, two DEMO designs, and an ideal pure 14-MeV flux. Our analysis indicates that, within standard uncertainties inherent to the models employed, the DONES concept adequately captures fusion conditions in the four materials analyzed. Our work is intended as a comprehensive irradiation damage analysis of fusion-representative neutron sources, to be used for further neutron source evaluation and fusion facility operation.

Marian, Jaime [Univ. of California, Los Angeles, C↗