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At least 415 records · Page 23

Investigating the Dynamic Compaction Behavior of Brittle Powders

The compression of porous materials to full density is a complex physical and mechanical process. Modeling efforts to capture the compaction behavior of distended solids (e.g., porous materials or powders) have produced a variety of compaction models, each functionalized to address specific compaction mechanisms. Of the available models, those suited for high-strain-rate deformation associated with shock loading have generally been developed for ductile materials. Compaction models for brittle materials have only been developed for quasistatic loading conditions. No predictive models currently exist that successfully capture the physical processes and mechanics for shock compaction of brittle distended materials, such as sand and oxide powders. The primary hindrance in producing a dynamic compaction model for brittle distended materials is the lack of availability of high-fidelity, high-precision shock compaction data for validation. The proposed work is aimed at alleviating this issue by developing a high-precision testing method in combination with high-fidelity analysis methods to accurately and precisely determine the shocked state of an ideal test material, ceria powder, under uniaxial strain dynamic compaction. The measured shock compaction data from these brittle powder experiments is then used in initial computations to design a cylindrically converging shock compaction experiment within situ time-resolved shocked density measurements, employing proton radiography. The preliminary experimental and computational results obtained from these uniaxial strain dynamic compaction and cylindrically converging shock compaction experiments are presented. An experimental and computational path forwards to predictive model development is then proposed.

36 MATERIALS SCIENCE↗

Development of the TREAT M2 experiment as a transient benchmark

A transient benchmark based on the Transient Reactor Test Facility (TREAT) M2 Calibration experiment (M2-CAL) is under development. TREAT, at Idaho National Laboratory, is a graphite moderated air-cooled research reactor which has been used extensively for fuel material testing under extreme and accident conditions. Accurate benchmark models are a beneficial component in the operations, experimental planning, and development of TREAT. In this work, we present a transient benchmark model for the M2-CAL experiment core loading. The benchmark model incorporates the coupling between the Monte Carlo Code SERPENT and the computational fluid dynamics code OpenFOAM to capture the temperature feedback mechanism. The M2-CAL transient 2580 was simulated in this work. A pre-transient analysis was performed to determine the optimum core composition and conditions before the beginning of the transient. The analysis was validated against historic TREAT kinetic measurements and the worth of the transient rod T-2. The axial power distribution in the flux wire for the M2-CAL experiment was determined and contrasted with the experiment. The model was then used to simulate the M2-CAL transient 2580 experiment based on the reported pre-transient and transient conditions. Several transient observables were calculated and compared to the experiment. The model shows good agreement with the experimental power traces, period, and average increase of power at the power ramp (time > 7.8 s). Inverse point kinetics analysis was performed during the period of the power ramp for the model and the experiment based on that, the temperature feedback component was isolated and contrasted with the experimental feedback. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

In-situ subaqueous capping of mercury-contaminated sediments in a fresh-water aquatic system, Part II-evaluation of sorption materials

The function and longevity of traditional, passive, isolation caps can be augmented through the use of more chemically active capping materials which have higher sorptive capacities, ideally rendering metals non-bioavailable. In the case of Hg, active caps also mitigate the rate and extent of methylation. This research examined low cost, readily available, capping materials for their ability to sequester Hg and MeHg. Furthermore, selected capping materials were evaluated to inhibit the methylation of Hg in an incubation study as well as the capacity of a selected capping material to inhibit translocation of Hg and MeHg with respect to ebullition-facilitated contaminant transport in a column study. Results indicated that bauxite had a better capacity for mercury sorption than the other test materials. However, bauxite as well as soil capping materials did not decrease methylation to a significant extent. Materials with larger surface areas, higher organic matter and acid volatile sulfide (AVS) content displayed a larger partitioning coefficient. In the incubation experiments, the presence of a carbon source (lactate), electron acceptor (sulfate) and the appropriate strains of SRB provided the necessary conditions for Hg methylation to occur. The column study showed effectiveness in sequestering Hg and MeHg and retarding transport to the overlying water column; however, disturbances to the soil capping material resulting from gas ebullition negated its effectiveness.

54 ENVIRONMENTAL SCIENCES↗

Instrumentation for the In-Core Real-Time Mechanical Testing of Structural Materials (INCREASE) Project

Idaho National Laboratory (INL), in collaboration with the Electric Power Research Institute (EPRI), the Nuclear Regulatory Commission (NRC), the French Atomic and Alternative Energies Commission (CEA), the Joint Research Center (JRC), the Nuclear Research and Consultancy Group (NRG), and the Research Center Rez (CVR), started a Joint Experimental Program (JEEP) project that operates within the Nuclear Energy Agency’s Framework for Irradiation Experiments (FIDES II) program in order to develop capabilities for the in-core real-time mechanical testing of structural materials. This effort will focus on designing a shared capsule capable of housing a variety of in-core mechanical testing instrumentation allowing enhanced experiments for the material science community. The outcome of the project would be high-priority, stress relaxation data for stainless-steel-based materials provided by EPRI and CEA. Stress relaxation is a major phenomenon that contributes to material degradation in nuclear reactor components. Currently, nuclear material stress relaxation is assessed both before and after irradiation, using complex and costly post-irradiation examination (PIE) activities. In-situ data would support the development of precision modeling and simulation of this degradation phenomena and would provide validation and benchmarking for existing models using the PIE data. As part of the U.S. Department of Energy Advanced Sensor and Instrumentation (ASI) program, INL has fabricated and tested out-of-core mechanical test instrumentation. This instrumentation was designed for easy adaptation to the irradiation capsule proposed under this JEEP, and can be deployed to measure real-time stress relaxation under pressurized-water reactor (PWR) conditions. This initial effort will partially serve to replace the testing capabilities lost because of shutting down the Halden Boiling Water Reactor (HBWR). The project would provide these capabilities to the international community via a shared capsule design that is easily adaptable to additional material test reactors. The design features will incorporate expansion to PWR and non-light-water reactor (LWR) environments that will be developed in future work. The capsule and instrumentation will be demonstrated in the Massachusetts Institute of Technology Reactor (MITR) for phase I and the Petten High Flux Reactor (HFR) for phase II irradiations to deliver real-time stress relaxation data on high priority stainless steel structural materials.

36 MATERIALS SCIENCE↗

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

Synthesis of small-pore zeolitic materials with realistic loadings of transition metals atomically dispersed in the micropores has been a formidable challenge. Herein, we demonstrate a new simple and scalable route to high Pd and Pt loadings (>1 wt%) atomically dispersed in the micropores of a small-pore zeolite, SSZ-13. Unlike all previous methods that require addition of expensive transition metal precursor complexes into the synthesis gel and produce low loading of atomically dispersed species, this method does not require any complicated synthetic proce-dures. We determine the factors that govern the stabilization of M(II) species in the micropores of SSZ-13. Using a combined spectroscopic and DFT approach, we offer a unifying explana-tion for the discrepancies existing in available M-zeolite litera-ture. Moreover, we highlight the difference between the behavior of Pt and Pd during synthesis, and demonstrate that ionically dispersed Pd/SSZ-13 shows superior performance as a new class of environmentally important passive NOx Adsorber (PNA) material under practically relevant conditions. It thus represents the 2nd known metal-small pore zeolite material that has immediate industrial application, and the only known met-al-small pore zeolite material in which metal is close to 100 % utilized. In addition to NO x , the material performs remarkably for CO removal during cold start as well due to the formation of a mixed carbonyl-nitrosyl complex Pd(II)(CO)(NO).

Khivantsev, Konstantin↗

Recent progress and tests of radiation resistant impregnation materials for Nb{sub 3}Sn coils

Fermilab is collaborating with Lawrence Berkeley National Laboratory (LBNL) and Brookhaven National Laboratory (BNL) (US-LARP collaboration) to develop a large-aperture Nb{sub 3}Sn superconducting quadrupole for the Large Hadron Collider (LHC) luminosity upgrade. An important component of this work is the development of materials that are sufficiently radiation resistant for use in critical areas of the upgrade. This paper describes recent progress in characterization of materials, including the baseline CTD101K epoxy, cyanate ester blends, and Matrimid 5292, a bismaleimide-based system. Structural properties of “ten stacks” of cable impregnated with these materials are tested at room and cryogenic temperatures and compared to the baseline CT-101K. Experience with potting 1 and 2 meter long coils with Matrimid 5292 are described. Test results of a single 1-m coil impregnated with Matrimid 5292 are reported and compared to similar coils impregnated with the traditional epoxy.

43 PARTICLE ACCELERATORS↗

Changes to the morphology and coefficient of thermal expansion in HDPE and UHMWPE following irradiation-based crosslinking

This work describes recent findings of the Los Alamos National Laboratory concerning the alterations to polyethylene’s (PE’s) mechanical and chemical properties after crosslinking via gamma irradiation. Three materials were examined as part of this study, including one HDPE and two UHMWPE specimens. Changes to degree of crystallinity (X C ) were qualitatively determined via measurements of density, and results suggest that X C is not necessarily correlated to crosslink density, depending on polymer manufacturing process and treatment following irradiation. TMA data suggests an anisotropic response in linear thermal expansion, and that the response may be more correlated to X C than crosslink density. Finally, FTIR analysis has led to the conclusion that oxidation occurs as a result of irradiation in open air, the degree of oxidation is proportional to the maximum radiation dose received, and that oxidation is largely a surface effect, penetrating no deeper than 1.5 mm. Due to time constraints, a full dataset could not be generated for changes in creep and thermal properties as a result of irradiation, and TMA analysis is only complete for one material tested in this study. These measurements are ongoing and will provide more context to the hypotheses drawn in this work.

36 MATERIALS SCIENCE↗

Assessment of Cementitious Composites for High-Temperature Geothermal Wells

High-temperature (HT) geothermal wells can provide green power 24 hours a day, 7 days a week. Under harsh environmental and operational conditions, the long-term durability requirements of such wells require special cementitious composites for well construction. This paper reports a comprehensive assessment of geothermal cement composites in cyclic pressure function laboratory tests and field exposures in an HT geothermal well (300–350 °C), as well as a numerical model to complement the experimental results. Performances of calcium–aluminate cement (CAC)-based composites and calcium-free cement were compared against the reference ordinary Portland cement (OPC)/silica blend. The stability and degradation of the tested materials were characterized by crystalline composition, thermo-gravimetric and elemental analyses, morphological studies, water-fillable porosity, and mechanical property measurements. All CAC-based formulations outperformed the reference blend both in the function and exposure tests. The reference OPC/silica lost its mechanical properties during the 9-month well exposure through extensive HT carbonation, while the properties of the CAC-based blends improved over that period. The Modified Cam-Clay (MCC) plasticity parameters of several HT cement formulations were extracted from triaxial and Brazilian tests and verified against the experimental results of function cyclic tests. These parameters can be used in well integrity models to predict the field-scale behavior of the cement sheath under geothermal well conditions.

15 GEOTHERMAL ENERGY↗

Nitrile O-ring Cracking: A Case of Vacuum Flange O-ring Failures

A review of recent nitrile O-ring failures in ISO-KF vacuum flange connections in glovebox applications is presented. An investigation of a single “isolated” o-ring failure leads to the discovery of cracked nitrile o-rings in a glovebox atmospheric control unit. The initial cause of the o-ring failure is attributed to ozone degradation. However, additional investigation reveals nitrile o-ring cracking on multiple gloveboxes and general purpose piping, roughly 85% of the nitrile o-rings removed for inspection show evidence of visible cracking after being in service for 18 months or less. The results of material testing and ambient air testing is presented, elevated ozone levels are not found. The contributing factors of o-ring failure, including nitrile air sensitivity, inadequate storage practices, and poor installation techniques, are discussed. A discussion of nitrile o-ring material properties, the benefits and limitations, and alternate materials are discussed. Considerations for o-ring material selection, purchasing, storage, and installation are presented in the context of lessons learned from the nitrile o-ring cracking investigation. This paper can be presented in 20 minutes and does not require special accommodations or special audio visual devices.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

Apparatus and method for fatigue testing of a material specimen in a high-pressure fluid environment

The invention provides fatigue testing of a material specimen while the specimen is disposed in a high pressure fluid environment. A specimen is placed between receivers in an end cap of a vessel and a piston that is moveable within the vessel. Pressurized fluid is provided to compression and tension chambers defined between the piston and the vessel. When the pressure in the compression chamber is greater than the pressure in the tension chamber, the specimen is subjected to a compression force. When the pressure in the tension chamber is greater than the pressure in the compression chamber, the specimen is subjected to a tension force. While the specimen is subjected to either force, it is also surrounded by the pressurized fluid in the tension chamber. In some examples, the specimen is surrounded by hydrogen.

36 MATERIALS SCIENCE↗

The Innovation of I-Loop

The closing and decommissioning of Halden Boiling Water Reactor in Norway resulted in the loss of many nuclear irradiation testing facilities, notably those providing boiling water reactor (BWR) environments. To lighten the loss of such an important asset, the Advanced Test Reactor (ATR) is being called upon. The recent overhaul of the ATR’s core and the installation of the new Top Head Closure Plate now makes it possible to utilize the ATR’s I-loop locations for I-Loop Tube test facilities. Similar to the current in-pile tube facilities, these new I-Loop facilities will capitalize on the available space and neutron flux the ATR generates in this region of the core for fuels and materials testing. Current efforts are underway to add new PWR and BWR I-Loop Tube facilities to the I-Loop positions for these purposes.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Cluster 1: commercializing additive manufacturing—hurdles in materials characterization and testing

A major challenge in the commercialization of additive manufactured (AM) materials and processes is the ability to achieve acceptance of processes and products. There has been some progress towards acceptance has been made by adapting legacy qualification paradigms to match with the very limited process control and monitoring offered by AM machines. The opportunity for in-situ measurement can provide process monitoring and control perhaps changing the way we qualify parts however it is limited by lack of adequate process measurement methods. New measurement techniques, sensors and correlations to relevant phenomena are needed that enable process control and monitoring for consistently producing high quality articles. Beyond process data we need to characterize uncertainties of performance in all aspects of material, process and final part. These are prerequisites to achieving articles that are indeed worthy of materials characterization efforts that establish a microstructural reference of desirable performance through process-structure-property relations. Only then can industry apply physics based understanding of the material, part and process to probabilistically predict performance of an AM part. Our paper provides a brief overview, discussion of hurdles and key areas where R&D investment is needed.

36 MATERIALS SCIENCE↗

A fusion nuclear science facility for a fast-track path to DEMO

An accelerated fusion energy development program, a fast-track approach, requires proceeding with a nuclear and materials testing program in parallel with research on burning plasmas, ITER. A Fusion Nuclear Science Facility (FNSF) would address many of the key issues that need to be addressed prior to DEMO, including breeding tritium and completing the fuel cycle, qualifying nuclear materials for high fluence, developing suitable materials for the plasma-boundary interface, and demonstrating power extraction. The Advanced Tokamak (AT) is a strong candidate for an FNSF as a consequence of its mature physics base, capability to address the key issues, and the direct relevance to an attractive target power plant. The standard aspect ratio provides space for a solenoid, assuring robust plasma current initiation,and for an inboard blanket, assuring robust tritium breeding ratio (TBR) >1 for FNSF tritium self-sufficiency and building of inventory needed to start up DEMO. An example design point gives a moderate sized Cu-coil device with R/a = 2.7 m/0.77 κ = 2.3, B T = 5.4 T, I P = 6.6 MA, β N = 2.75, P fus = 127 MW. The modest bootstrap fraction of f BS = 0.55 provides an opportunity to develop steady state with sufficient current drive for adequate control. Lastly, proceeding with a FNSF in parallel with ITER provides a strong basis to begin construction of DEMO upon the achievement of Q ~ 10 in ITER.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Improved Accelerated Stress Tests Based on Fuel Cell Vehicle Data

UTC will led a top-tier team of industry and national laboratory participants to update and improve DOE’s Accelerated Stress Tests (AST’s) for hydrogen fuel cells. This in-depth investigation will focused on critical fuel cell components (e.g. membrane electrode assemblies - MEA) whose durability represented barriers for widespread commercialization of hydrogen fuel cell technology. UTC had access to MEA materials that had accrued significant load time under real-world conditions in PureMotion® 120 power plant used in transit buses. These materials are referred to as end-of-life (EOL) components in the rest of this document. Advanced characterization techniques were used to evaluate degradation mode progress using these critical cell components extracted from both bus power plants and corresponding materials tested using the DOE AST’s. These techniques were applied to samples at beginning-of-life (BOL) to serve as a baseline. These comparisons advised the progress of the various failure modes that these critical components were subjected to, such as membrane degradation, catalyst support corrosion, platinum group metal dissolution, and others. Gaps in the existing ASTs predicted the degradation observed in the field in terms of these modes were outlined. Using the gaps, new AST’s were recommended and tested to better reflect the degradation modes seen in field operation. Also, BOL components were degraded in a test vehicle at UTC designed to accelerate the bus field operation.

08 HYDROGEN↗

Addressing Research and Development Gaps for Plasma-Material Interactions with Linear Plasma Devices

Plasma-material interactions in future fusion reactors have been identified as a knowledge gap to be dealt with before any next step device past ITER can be built. The challenges are manifold. They are related to power dissipation so that the heat fluxes to the plasma-facing components can be kept at technologically feasible levels; maximization of the lifetime of divertor plasma-facing components that allow for steadystate operation in a reactor to reach the neutron fluence required; the tritium inventory (storage) in the plasma-facing components, which can lead to potential safety concerns and reduction in the fuel efficiency; and it is related to the technology of the plasma-facing components itself, which should demonstrate structural integrity under the high temperatures and high neutron fluence. While the dissipation of power exhaust can and should be addressed in high power toroidal devices, the interaction of the plasma with the materials can be best addressed in dedicated linear devices due to their cost effectiveness and ability to address urgent research and development gaps more timely. However, new linear plasma devices are needed to investigate the PMI under fusion reactor conditions and test novel plasma-facing components. Existing linear devices are limited either in their flux, their reactor-relevant plasma transport regimes in front of the target, their fluence, or their ability to test material samples a priori exposed to high neutron fluence. The proposed Material Plasma Exposure eXperiment (MPEX) is meant to address those deficiencies and will be designed to fulfill the fusion reactor-relevant plasma parameters as well as the ability to expose a priori neutron activated materials to plasmas.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Bacterial Membranes Are More Perturbed by the Asymmetric Versus Symmetric Loading of Amphiphilic Molecules

Characterizing the biophysical properties of bacterial membranes is critical for understanding the protective nature of the microbial envelope, interaction of biological membranes with exogenous materials, and designing new antibacterial agents. Presented here are molecular dynamics simulations for two cationic quaternary ammonium compounds, and the anionic and nonionic form of a fatty acid molecule interacting with a Staphylococcus aureus bacterial inner membrane. The effect of the tested materials on the properties of the model membranes are evaluated with respect to various structural properties such as the lateral pressure profile, lipid tail order parameter, and the bilayer’s electrostatic potential. Conducting asymmetric loading of molecules in only one leaflet, it was observed that anionic and cationic amphiphiles have a large impact on the Staphylococcus aureus membrane’s electrostatic potential and lateral pressure profile as compared to a symmetric distribution. Nonintuitively, we find that the cationic and anionic molecules induce a similar change in the electrostatic potential, which points to the complexity of membrane interfaces, and how asymmetry can induce biophysical consequences. Finally, we link changes in membrane structure to the rate of electroporation for the membranes, and again find a crucial impact of introducing asymmetry to the system. Understanding these physical mechanisms provides critical insights and viable pathways for the rational design of membrane-active molecules, where controlling the localization is key.

59 BASIC BIOLOGICAL SCIENCES↗

United States Advanced Ultra-Supercritical Component Test Facility for 760°C Steam Power Plants ComTest Project

Following the successful completion of a 15-year effort to develop and test materials that would allow coal-fired power plants to be operated at advanced ultra-supercritical (A-USC) steam conditions, a United States-based consortium is presently engaged in a project to build an A-USC component test facility (ComTest). A-USC steam cycles have the potential to improve cycle efficiency, reduce fuel costs, and reduce greenhouse gas emissions. Current development and demonstration efforts are focused on enabling the construction of A-USC plants, operating with steam temperatures as high as 1400°F (760°C) and steam pressures up to 5000 psi (35 MPa), which can potentially increase cycle efficiencies to 47% HHV (higher heating value), or approximately 50% LHV (lower heating value), and reduce CO 2 emissions by roughly 25%, compared to today’s U.S. fleet. A-USC technology provides a lower-cost method to reduce CO 2 emissions, compared to CO 2 capture technologies, while retaining a viable coal option for owners of coal generation assets. Among the goals of the ComTest facility are to validate that components made from advanced nickel-based alloys can operate and perform under A-USC conditions, to accelerate the development of a U.S.-based supply chain for the full complement of A-USC components, and to decrease the uncertainty of cost estimates for future A-USC power plants. The configuration of the ComTest facility would include the key A-USC technology components that were identified for expanded operational testing, including a gas-fired superheater, high-temperature steam piping, steam turbine valve, and cycling header component. Membrane walls in the superheater have been designed to operate at the full temperatures expected in a commercial A-USC boiler, but at a lower (intermediate) operating pressure. This superheater has been designed to increase the temperature of the steam supplied by the host utility boiler up to 1400°F (760°C). The steam turbine stop and control valve component has been designed to operate at full A-USC temperatures, and would be tested both in throttling operation and to accumulate accelerated, repetitive stroke cycles. A cycling header component has been designed to confirm the suitability of new high-temperature nickel alloys to cycling operation, expected of future coal-fired power plants. Current test plans would subject these components to A-USC operating conditions for at least 8,000 hours by September 2020. The ComTest project is managed by Energy Industries of Ohio, and technically directed by the Electric Power Research Institute, Inc., with General Electric designing the A-USC components. This consortium is completing the Detailed Engineering phase of the project, with procurement scheduled to begin in late 2017. The effort is primarily funded by the U.S. Department of Energy, through the National Energy Technology Laboratory, along with the Ohio Development Services Agency. This presentation outlines the motivation for the project, explains the project’s structure and schedule, and provides technical details on the design of the ComTest facility.

54 ENVIRONMENTAL SCIENCES↗

CORROSION OF ALUMINUM CLAD SPENT NUCLEAR FUEL IN THE 70 TON CASK DURING TRANSFER FROM L AREA TO H-CANYON

Aluminum-clad spent nuclear fuel will be transported for processing in the 70-ton nuclear fuel element cask from L Basin to H-canyon. During transport these fuels would be expected to experience high temperature aqueous corrosion from the residual L Basin water that will be present in the cask. Cladding corrosion losses during transport were calculated for material test reactor (MTR) and high flux isotope reactors (HFIR) fuels using literature and site information on aqueous corrosion at a range of time/temperature conditions. Calculations of the cladding corrosion loss were based on Arrhenius relationships developed for aluminum alloys typical of cladding material with the primary assumption that an adherent passive film does not form to retard the initial corrosion rate. For MTR fuels a cladding thickness loss of 33 % was found after 1 year in the cask with a maximum temperature of 260 {degrees}C. HFIR fuels showed a thickness loss of only 6% after 1 year at a maximum temperature of 180 {degrees}C. These losses are not expected to impact the overall confinement function of the aluminum cladding.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗