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DOE OSTI2021

Force Reconstruction Using the Inverse of the Mode-Shape Matrix [Book Chapter]

A new technique for force reconstruction is developed. To estimate the externally applied force, this technique sums the weight-scaled acceleration signals, and is referred to as the Sum of Weighted Accelerations Technique (SWAT). To obtain the scalar weights the inverse of the mode shape matrix is used. Application of this technique is illustrated with both numerical calculations using a mass-spring model and experimental data from a structure impacting a rigid barrier.

055001 -- Nuclear Fuels-- Safeguards, Inspection, ↗

DOE OSTI2021

Load Deflection Data (LDDATA)

The current data set is load deflection data (Stress vs strain for compression stress and stress vs elongation for tension stress) from a material test. The goal of the test was to characterize this rubber-like material by looking at different variables: sample creation process, anisotropy, temperature, etc. The data will be used with the Texas A&M Online Research Experience for Undergraduates (O-REU) program. This program is joint with LANL. The student is not an official LANL student so they will not be able to access data and documents on the yellow. The load deflection data set is unclassified and will be sent to the student for analysis. Analysis will include curve fitting, scalar value analysis (stiffness, toughness, etc.) and analysis of variance to determine what variables statistically affect the response while correcting for covariates (temperature soak time and humidity). Data will be stored as Matlab structs and as HDF5 files. Same data, different formats.

36 MATERIALS SCIENCE↗

DOE OSTI2014

Characterization of basin concrete in support of structural integrity demonstration for extended storage

Concrete core samples from C basin were characterized through material testing and analysis to verify the design inputs for structural analysis of the L Basin and to evaluate the type and extent of changes in the material condition of the concrete under extended service for fuel storage. To avoid the impact on operations, core samples were not collected from L area, but rather, several concrete core samples were taken from the C Basin prior to its closure. C basin was selected due to its similar environmental exposure and service history compared to L Basin. The microstructure and chemical composition of the concrete exposed to the water was profiled from the water surface into the wall to evaluate the impact and extent of exposure. No significant leaching of concrete components was observed. Ingress of carbonation or deleterious species was determined to be insignificant. No evidence of alkali-silica reactions (ASR) was observed. Ettringite was observed to form throughout the structure (in air voids or pores); however, the sulfur content was measured to be consistent with the initial concrete that was used to construct the facility. Similar ettringite trends were observed in the interior segments of the core samples. The compressive strength of the concrete at the mid-wall of the basin was measured, and similar microstructural analysis was conducted on these materials post compression testing. The microstructure was determined to be similar to near-surface segments of the core samples. The average strength was 4148 psi, which is well-above the design strength of 2500 psi. The analyses showed that phase alterations and minor cracking in a microstructure did not affect the design specification for the concrete.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2016

Report on Status of Shipment of High Fluence Austenitic Steel Samples for Characterization and Stress Corrosion Crack Testing

The goal of the Mechanisms of Irradiation Assisted Stress Corrosion Cracking (IASCC) task in the LWRS Program is to conduct experimental research into understanding how multiple variables influence the crack initiation and crack growth in materials subjected to stress under corrosive conditions. This includes understanding the influences of alloy composition, radiation condition, water chemistry and metallurgical starting condition (i.e., previous cold work or heat treatments and the resulting microstructure) has on the behavior of materials. Testing involves crack initiation and growth testing on irradiated specimens of single-variable alloys in simulated Light Water Reactor (LWR) environments, tensile testing, hardness testing, microstructural and microchemical analysis, and detailed efforts to characterize localized deformation. Combined, these single-variable experiments will provide mechanistic understanding that can be used to identify key operational variables to mitigate or control IASCC, optimize inspection and maintenance schedules to the most susceptible materials/locations, and, in the long-term, design IASCC-resistant materials. In support of this research, efforts are currently underway to arrange shipment of “free” high fluence austenitic alloys available through Électricité de France (EDF) for post irradiation testing at the Oak Ridge National Laboratory (ORNL) and IASCC testing at the University of Michigan. These high fluence materials range in damage values from 45 to 125 displacements per atom (dpa). The samples identified for transport to the United States, which include nine, no-cost, 304, 308 and 316 tensile bars, were relocated from the Research Institute of Atomic Reactors (RIAR) in Dimitrovgrad, Ulyanovsk Oblast, Russia, and received at the Halden Reactor in Halden, Norway, on August 23, 2016. ORNL has been notified that a significant amount of work is required to prepare the samples for further shipment to Oak Ridge, Tennessee. The preliminary work for sample shipment between Halden and Oak Ridge includes fabrication of an inner cask sample container, decontamination and preparation of a Type A container, preparation of new activity calculations, all necessary paperwork, and handling. ORNL will continue to work to track progress of sample preparation and shipment status, and to work toward an agreement that covers material shipping costs between the Halden Reactor and the Oak Ridge National Laboratory.

36 MATERIALS SCIENCE↗

DOE OSTI2017

Cantilever testing of sintered-silver interconnects

Cantilever testing is an underutilized test method from which results and interpretations promote greater understanding of the tensile and shear failure responses of interconnects, metallizations, or bonded joints. The use and analysis of this method were pursued through the mechanical testing of sintered-silver interconnects that joined Ni/Au-plated copper pillars or Ti/Ni/Ag-plated silicon pillars to Ag-plated direct bonded copper substrates. Sintered-silver was chosen as the interconnect test medium because of its high electrical and thermal conductivities and high-temperature capability—attractive characteristics for a candidate interconnect in power electronic components and other devices. Deep beam theory was used to improve upon the estimations of the tensile and shear stresses calculated from classical beam theory. The failure stresses of the sintered-silver interconnects were observed to be dependent on test-condition and test-material-system. In conclusion, the experimental simplicity of cantilever testing, and the ability to analytically calculate tensile and shear stresses at failure, result in it being an attractive mechanical test method to evaluate the failure response of interconnects.

36 MATERIALS SCIENCE↗

DOE OSTI2011

New detector for use in fast neutron radiography

We have developed and tested a new type detector for use in the fast neutron (FN) imaging radiography applications. FN radiography is generally used for nondestructive material testing, medical and biology applications, border patrol, transportation and cargo screening tasks. It is complementary to other types of radiography, providing additional information on light element content of the material samples. Distinct from other FN imagers presently known, our device implements a neutron-sensitive scintillator attached to a position-sensitive photomultiplier tube (PSPMT), and operates in an event-by-event readout mode, acquiring energy, timing, and pulse shape information for all detected radiation events. The information is used to help separate events of FN interactions in the scintillator from the background events caused by the electronics noise and by the other types of background radiation. Selection of pure fast neutron events in the final image allows us to achieve ultimate image contrast and resolution, as compared with other types of FN imaging devices operating most commonly in an integration mode, in which the detector?s dark noise and radiation background dilute the images. The detector performance for FN imaging application was tested using D-D neutron generator, designed and manufactured by Adelphi Technology, Inc. This essentially point-like neutron source operates in continuous mode producing up to 109 of 2.5 MeV neutrons per second. Samples made of metals plastic and other material were used to measure the detector resolution, efficiency and uniformity. Results of these tests are presented and discussed. Fig. 1 shows one of the test FN radiographic images obtained using the sample made of 11 styrene plastic strips. All strips are squares 4.8 x 4.8 mm2 with six different lengths 10 to 60 mm with 10 mm increment. [A] [B] [C] Fig. 1. [A]-layout of the test sample; [B]-raw FN shadow image of the sample; [C]-map of the plastic strips as they appear on the image, with measured FN absorption factor shown in percent. Both X and Y position resolution of the FN imaging detector are estimated to be less than 0.5 mm (sigma) based on the analysis of the edges of the strip shadows in the plot. Because this detector shows the fraction-of-a-millimeter resolution desirable for most of FN applications, and is built using radiation hard materials, we believe it could be a good alternative to other FN imaging systems based on CCD or solid state detectors. In addition, having a sub-nanosecond timing resolution, it is suitable for the time-of-flight energy-resolved FN imaging.

Popov, V.↗

DOE OSTI2026

Interconnect: Cooperative Research and Development Final Report, CRADA Number CRD-13-00507 (Project 4)

This CRADA modification involves analyses on a variety of CdTe-PV related materials, test structures, solar cells, and modules produced at FSLR and/or NLR and adds analysis related to module degradation and reliability. Materials will be provided by FSLR, NLR, and/or by interleaving FSLR and NLR layers and processes. Module reliability activities will include technical risk assessment, materials characterization, module and test structure characterization, modeling, accelerated test development, and outdoor testing.

14 SOLAR ENERGY↗

DOE OSTI2014

Ultrasonic detection technology based on joint robot on composite component with complex surface

Some components have complex surface, such as the airplane wing and the shell of a pressure vessel etc. The quality of these components determines the reliability and safety of related equipment. Ultrasonic nondestructive detection is one of the main methods used for testing material defects at present. In order to improve the testing precision, the acoustic axis of the ultrasonic transducer should be consistent with the normal direction of the measured points. When we use joint robots, automatic ultrasonic scan along the component surface normal direction can be realized by motion trajectory planning and coordinate transformation etc. In order to express the defects accurately and truly, the robot position and the signal of the ultrasonic transducer should be synchronized.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗

DOE OSTI2024

Progress towards subsurface spectroscopy inside heterogeneous materials using four wave mixing phase conjugation

Herein, we report preliminary results on subsurface spectroscopy inside heterogeneous materials using four wave mixing phase conjugation. For our test material we use an inert plastic bonded explosive simulant that contains fluorescent guidestar particles. Using phase conjugation, we are able to obtain spectral enhancements approximately equal to the phase conjugate reflectivity. Additionally, we discuss future directions to improve this technique to obtain greater signal enhancements.

47 OTHER INSTRUMENTATION↗

DOE OSTI2021

Performance of three-dimensional printed nasopharyngeal swabs for COVID-19 testing

At the start of the COVID-19 pandemic, the US faced nationwide shortages of nasopharyngeal swabs due to both overwhelmed supply chains and an increase in demand. To address this shortfall, multiple 3D printed swabs were ultimately produced and sold for COVID-19 testing. In this work, we present a framework for mechanical and functional bench-testing of nasopharyngeal swabs using standard and widely available material testing equipment. Using this framework, we offer a comprehensive, quantitative comparison of the 3D printed swabs to benchmark their performance against traditional flocked swabs. The test protocols were designed to emulate the clinical use of the nasopharyngeal swabs and to evaluate potential failure modes. Overall, the 3D printed swabs performed comparably to, or outperformed, the traditional swabs in all mechanical tests. While traditional swabs outperformed some of the new 3D printed swabs in terms of sample uptake and retention, similar amounts of RNA were recovered from both 3D printed and traditional swabs.

3D printing↗

DOE OSTI2021

Nuclear Requalification of the New ATR Core

The Advanced Test Reactor (ATR) creates a unique high neutron flux materials testing environment that subsequently causes neutron embrittlement damage to its beryllium reflector eventually requiring a replacement during a reactor Outage. Concurrently, a piece-for-piece core replacement is conducted to maintain reactor functionality called a core internals changeout (CIC). Once changeout is complete, ATR Reactor Engineering utilizes previous CIC data to model and predict criticality scenarios for the new unirradiated core. However, CIC data has been inconsistently reported over the years making it difficult to find data trends that could be useful for the modeling and prediction process. So nuclear requalification testing is conducted using different flux and irradiation tests to accurately verify the models and to recalibrate nuclear monitoring instrumentation in the ATR core allowing for safe re-start of normal operations.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2024

Nuclear Requalification of the ATR Core

The Advanced Test Reactor (ATR) creates a unique, high neutron flux materials testing environment that subsequently causes neutron embrittlement damage to its beryllium reflector. The reflector is eventually required to be replaced during a reactor Outage. Concurrently, a piece-for-piece core replacement is conducted to maintain reactor functionality called a core internals changeout (CIC). Once changeout is complete, ATR Reactor Engineering utilizes previous CIC data, notably 1977-1994 CIC data, to conduct predictive criticality analysis on the new, unirradiated core. However, CIC data has been inconsistently reported over the years making it difficult, and in some cases impossible, to utilize. These inconsistencies arise in various forms of not using correct units, inconsistent diagram usage, and not portraying the same kind of data similarly across the reports. To work around this inconsistency issue, Reactor Engineering is implementing the use of several new neutronic modeling codes that utilize modern computation speeds and methods to model the new core in different criticality environments. Further validation is then conducted by nuclear testing and bringing the reactor critical and conducting various irradiation tests. Upon completion and certification that the reactor passed all testing requirements, ATR can once again re-start normal operations.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2019

Ion Fluxes and Neutral Gas Ionization Efficiency of the 100-kW Light-Ion Helicon Plasma Source Concept for the Material Plasma Exposure eXperiment

Linear plasma devices are cost-effective alternatives for testing materials under reactor-relevant divertor plasma conditions. In this work, we describe how an intense radio-frequency (RF) plasma source concept for the Material Plasma Exposure eXperiment (MPEX) is under development at Oak Ridge National Laboratory. The source concept, Proto-MPEX, aims to produce high-density background deuterium helicon plasmas ( n e > 3 × 10 19 m - 3 ) that are subsequently heated with additional RF and microwave systems to deliver reactor-relevant conditions for studies on plasma-material interaction. In this work, we focus on the plasma-producing stage and its effectiveness in converting input neutral gas into plasma, namely, the neutral gas ionization efficiency. We provide a direct quantitative measurement of the ionization efficiency by measuring the total ion flux arriving at the target region relative to the neutral gas injected at the source. Using 80 kW at 13.56 MHz and a source magnetic field of 0.05 T, the helicon plasma source delivers ion fluxes up to 1 × 10 24 m - 2 s - 1 and heat fluxes greater than 1 M W m - 2 to a target plate located 2 m away from the source. Under these conditions, we observe that the plasma source converts ~89% of the input neutral gas into plasma that arrives at the target as ion flux at a rate of 5 × 10 20 s - 1 . We demonstrate that because of the large pumping capacity of the plasma, neutral gas pumping systems are required only in the target region to maintain optimal plasma operation.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2020

Augmented Monitoring and Condition Assessment Program for SNF Wet Storage Life Extension - 20489

Approximately 27 MTHM of spent nuclear fuel (SNF) owned and managed by the U.S. Department of Energy, Office of Environmental Management is stored in the L Basin at the Savannah River Site (SRS). This 'DOE SNF' is comprised of approximately 12,000 aluminum-clad, aluminum-based fuel assemblies (∼7 MTHM), and approximately 2000 non-aluminum fuel assemblies (∼20 MTHM). A program is in progress to perform Non-Destructive Examination (NDE) of the fuel and their storage containers to characterize the materials' condition, and to evaluate the effects of service to enable continued safe wet storage of the SNF. The Augmented Monitoring and Condition Assessment Program (AMCAP), a two-part program to develop and implement remote underwater Non-Destructive Examination of the aluminum SNF, and of the containers of the non-aluminum SNF, respectively, is aimed at the characterization and evaluation of corrosion degradation of aluminum fuel and container materials. The predominant design of the aluminum SNF (ASNF) stored in L Basin is the Materials Test Reactor (MTR) equivalent design, a plate fuel design. These SNF are stored in 5'' diameter tubes or 5'' x 5'' squares (called bundles) that are nominally 12' or up to 14' long. The SNF is de-bundled and inspected using a custom-designed MTR Fuel Inspection Table. The inspection table provides for indexed fuel positioning for a video camera examination with controlled lighting. A total of 10 of the SNF originating from foreign research and test reactors were selected for examination based on burnup, enrichment, and prior damage caused by service/storage history. A special inspection campaign of these 10 assemblies is in progress. The observed corrosion damage included minor to moderate attack from general corrosion, pitting, crevice, end grain and galvanic corrosion. Example results from the completed inspections are shown and discussed. The focus is a comparison of the as-received condition versus the as-found current storage condition that will serve to validate Water Chemistry and Corrosion Monitoring Programs. The non-aluminum SNF (NASNF) stored in L Basin are of diverse design that includes various geometries with claddings of stainless steel, Zircaloy, and Hastelloy. The fuel core materials include uranium alloys, oxides/mixed oxides, and carbides. These SNF materials, originating from early experimental and test reactors, are in various physical forms including single fuel elements and cut pieces. The fuel is stored in L Basin in various configurations including in direct bundled storage in aluminum tubes and in isolation cans that are in the bundles or in a larger over-size storage container (OSC). Concern with inside-out corrosion and the potential loss of configuration control and ability to handle the storage containers prompted the development of remote NDE methods that include visual and UT technologies to assess the condition of the containers. Candidate stored materials were selected for a special inspection campaign; the development of the NDE methods for the inspection for galvanic, crevice and sediment-induced corrosion (inside-out) are discussed. A summary of the SNF storage in L Basin at the SRS, and an overview of the AMCAP to enable continued safe storage of DOE SNF are described. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2022

Multi-Modal Characterization of Nuclear Fuels and Materials at the Idaho National Laboratory Materials & Fuels Complex

The Idaho National Laboratory (INL) leads cutting-edge research pertaining to the advancement of nuclear reactor technologies, including nuclear fuels and materials. The INL Irradiated Materials Characterization Laboratory (IMCL), Electron Microscopy Laboratory (EML), and future Sample Preparation Laboratory (SPL) are available to the nuclear research community to assess the behavior of nuclear fuels and materials, efficiently and comprehensively characterizing from the engineering to atomistic scale. The IMCL is a unique, 12,000-square-foot facility located at the INL Materials and Fuels Complex designed for analysis of irradiated materials. The facility operates advanced characterization instruments that are sensitive to vibration, temperature, and electromagnetic interference in modular radiological shielding and confinement systems, granting researchers the ability to assess the microstructural, chemical, mechanical and thermophysical properties of nuclear materials, especially irradiated fuels. The EML is dedicated to advanced characterization of materials with optical and electron microscopy tools, including scanning electron microscopy/focused-ion beam (SEM/FIB) and transmission electron microscopy (TEM). Upon construction, the SPL will be a 3 story, 49,000 sq. ft facility, that is the most modern reactor structural materials testing and analysis facility in the world, designed to investigate reactor structural materials in support of life-extension programs and development of advanced reactor concepts, including mechanical testing and advanced characterization capabilities. This presentation will showcase some of the main capabilities available at both IMCL, EML, and SPL, specifically illustrating how these characterization techniques are incorporated into multi-modal characterization work scopes to elucidate the degradation of nuclear structural materials and irradiated fuels.

36 MATERIALS SCIENCE↗

DOE OSTI2021

An initial framework for the rapid qualification of long-term creep rupture strength via microstructural modeling

This report describes the development and testing of a new method for extrapolating short-term creep rupture test data to predict long-term rupture strength. The goal of this work is to reduce the time required to qualify new materials for nuclear service by reducing the lead time required for dedicated, long-term material testing to establish key long-term material properties. The new approach described here uses a physics-based model to predict the long-term creep rupture strength of 316H stainless steel using only short-term test data. The key idea is to use Bayesian inference to find the statistical distribution of the model parameters that best explain the short-term rupture data. Because the model is physics-based these parameters are all microstructural quantities that can be measured through detailed material characterization experiments. The Bayesian prior distributions provide a means for incorporating this characterization data into the final model to improve the accuracy of the long-term model predictions. However, where such data is not available the process still produces an accurate model based on an uniformed prior. Our hypothesis is that this approach more accurately extrapolates the short-term test data when compared to current, empirical methods. The report proves this hypothesis using actual long-term rupture data available for 316H, including tests with rupture times greater than 200,000 hours. The general approach developed here could be applied to other materials and other time-dependent material properties. Applying this new technique to develop long-term qualified material properties, potentially in conjunction with other accelerated qualification approaches like staggered qualification test programs, could greatly reduce the time required to qualify new materials for nuclear service.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

Progress Towards Developing Neutron Tolerant Magnetostrictive and Piezoelectric Transducers

Current generation light water reactors (LWRs), sodium cooled fast reactors (SFRs), small modular reactors (SMRs), and next generation nuclear plants (NGNPs) produce harsh environments in and near the reactor core that can severely tax material performance and limit component operational life. To address this issue, several Department of Energy Office of Nuclear Energy (DOE-NE) research programs are evaluating the long duration irradiation performance of fuel and structural materials used in existing and new reactors. In order to maximize the amount of information obtained from Material Testing Reactor (MTR) irradiations, DOE is also funding development of enhanced instrumentation that will be able to obtain in-situ, real-time data on key material characteristics and properties, with unprecedented accuracy and resolution. Such data are required to validate new multi-scale, multi-physics modeling tools under development as part of a science-based, engineering driven approach to reactor development. It is not feasible to obtain high resolution/microscale data with the current state of instrumentation technology. However, ultrasound-based sensors offer the ability to obtain such data if it is demonstrated that these sensors and their associated transducers are resistant to high neutron flux, high gamma radiation, and high temperature. To address this need, the Advanced Test Reactor National Scientific User Facility (ATR-NSUF) is funding an irradiation, led by PSU, at the Massachusetts Institute of Technology Research Reactor to test the survivability of ultrasound transducers. As part of this effort, PSU and collaborators have designed, fabricated, and provided piezoelectric and magnetostrictive transducers that are optimized to perform in harsh, high flux, environments. Four piezoelectric transducers were fabricated with either aluminum nitride, zinc oxide, or bismuth titanate as the active element that were coupled to either Kovar or aluminum waveguides and two magnetostrictive transducers were fabricated with Remendur or Galfenol as the active elements. Pulse-echo ultrasonic measurements of these transducers are made in-situ. This paper will present an overview of the test design including selection criteria for candidate materials and optimization of test assembly parameters, data obtained from both out-of-pile and in-pile testing at elevated temperatures, and an assessment based on initial data of the expected performance of ultrasonic devices in irradiation conditions

Reinhardt1, Brian↗

DOE OSTI2024

Moving toward automated µFTIR spectra matching for microplastic identification: addressing false identifications and improving accuracy

Abstract Infrared spectroscopy is a widely used tool for studying microplastics and identifying microparticles. Researchers rely on spectral libraries to differentiate between synthetic and natural materials. Unfortunately, spectral library matching is not perfect, and best practices require researchers to use time consuming, manual peak matching to assess spectral matches. Moving toward automated matching requires increased confidence in the matching process. Using spectra matching software may increase the efficiency of particle identification, however some matching strategies may confuse natural materials such as cotton, silk, and plant matter with common classes of synthetics such as polyesters and polyamides. In this experiment, we prepared 22 pristine sample materials from natural and synthetic sources and measured micro-Fourier transform infrared (µFTIR) spectra in transmission mode for each sample using a Thermo Nicolet iN10 MX instrument. The collected spectra were then input into two spectral library matching systems (Omnic Picta and Open Specy), using a total of five identification routines. Next, we placed a subset of four pristine microplastic materials in a biologically active river system for two weeks to simulate environmental samples. These simulated environmental samples were processed using 10% hydrogen peroxide for 24 h to remove organic contamination and then identified using the strongest performing library. We found that libraries with fewer sample spectra produced lower correlation matches and that using derivative correction greatly reduced the number of inaccuracies in identifying materials as either natural or synthetic. We also found that environmental fouling reduced the correlation value of library matches when compared to pristine particles, however the effect was not consistent across the four materials tested. Overall, we found that the accuracy of automated library matching in the tested systems and processing routines varied from 64.1 to 98.0% for distinguishing between natural and synthetic materials, and that a high Hit Quality Index (HQI) did not always correlate with accuracy. These results are important for the microplastic field, demonstrating a need to rigorously test spectral libraries and processing routines with known materials to ensure identification accuracy.

Kozloski, Rachel↗