In 2007, the United States Department of Energy designated the Advanced Test Reactor (ATR), located at Idaho National Laboratory, as a National Scientific User Facility (NSUF). This designation made test space within the ATR and post-irradiation examination (PIE) equipment at INL available for use by researchers via a proposal and peer review process. The goal of the ATR NSUF is to provide researchers with the best ideas access to the most advanced test capability, regardless of the proposer's physical location. Since 2007, the ATR NSUF has expanded its available reactor test space, and obtained access to additional PIE equipment. Recognizing that INL may not have all the desired PIE equipment, or that some equipment may become oversubscribed, the ATR NSUF established a Partnership Program. This program enables and facilitates user access to several university and national laboratories. So far, seven universities and one national laboratory have been added to the ATR NSUF with capability that includes reactor-testing space, PIE equipment, and ion beam irradiation facilities. With the addition of these universities, irradiation can occur in multiple reactors and post-irradiation exams can be performed at multiple universities. In each case, the choice of facilities is based on the user's technical needs. …

The Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) is one of the world’s premier test reactors for studying the effects of intense neutron radiation on reactor materials and fuels. The ATR began operation in 1967, and has operated continuously since then, averaging approximately 250 operating days per year. The combination of high flux, large test volumes, and multiple experiment configuration options provide unique testing opportunities for nuclear fuels and material researchers. The ATR is a pressurized, light-water moderated and cooled, beryllium-reflected highly-enriched uranium fueled, reactor with a maximum operating power of 250 MWth. The ATR peak thermal flux can reach 1.0 x1015 n/cm2-sec, and the core configuration creates five main reactor power lobes (regions) that can be operated at different powers during the same operating cycle. In addition to these nine flux traps there are 68 irradiation positions in the reactor core reflector tank. The test positions range from 0.5” to 5.0” in diameter and are all 48” in length, the active length of the fuel. The INL also has several hot cells and other laboratories in which irradiated material can be examined to study material radiation effects. In 2007 the US Department of Energy (DOE) designated …

This is a report for public consumption, for the AVTA website, detailing the testing and analysis of the benchmark testing conducted on the Chevrolet Volt on-board charger.

This report summarizes the work performed by an Ames-led project team under a 4-year DOE-ITP sponsored project titled, 'Advanced Wear-resistant Nanocomposites for Increased Energy Efficiency.' The Report serves as the project deliverable for the CPS agreement number 15015. The purpose of this project was to develop and commercialize a family of lightweight, bulk composite materials that are highly resistant to degradation by erosive and abrasive wear. These materials, based on AlMgB{sub 14}, are projected to save over 30 TBtu of energy per year when fully implemented in industrial applications, with the associated environmental benefits of eliminating the burning of 1.5 M tons/yr of coal and averting the release of 4.2 M tons/yr of CO{sub 2} into the air. This program targeted applications in the mining, drilling, machining, and dry erosion applications as key platforms for initial commercialization, which includes some of the most severe wear conditions in industry. Production-scale manufacturing of this technology has begun through a start-up company, NewTech Ceramics (NTC). This project included providing technical support to NTC in order to facilitate cost-effective mass production of the wear-resistant boride components. Resolution of issues related to processing scale-up, reduction in energy intensity during processing, and improving the quality and …

NREL's National Wind Technology Center provides the world's only dedicated turbine controls testing platforms.

Idaho National Laboratory (INL) operates a mass separator that is currently producing high purity isotopes for use as internal standards for high precision isotope dilution mass spectrometry (IDMS). In 2008, INL began the revival of the vintage 1970’s era instrument. Advancements thus far include the successful upgrading and development of system components such as the vacuum system, power supplies, ion-producing components, and beam detection equipment. Progress has been made in the separation and collection of isotopic species including those of Ar, Kr, Xe, Sr, and Ba. Particular focuses on ion source improvements and developments have proven successful with demonstrated output beam currents of over 10 micro-amps 138Ba and 350nA 134Ba from a natural abundance source charge (approximately 2.4 percent 134Ba). In order to increase production and collection of relatively high quantities (mg levels) of pure isotopes, several advancements have been made in ion source designs, source material introduction, and beam detection and collection. These advancements and future developments will be presented.

This was a poster displayed at the Symposium. Advances on previous high throughput screening of biomass recalcitrance methods have resulted in improved conversion and replicate precision. Changes in plate reactor metallurgy, improved preparation of control biomass, species-specific pretreatment conditions, and enzymatic hydrolysis parameters have reduced overall coefficients of variation to an average of 6% for sample replicates. These method changes have improved plate-to-plate variation of control biomass recalcitrance and improved confidence in sugar release differences between samples. With smaller errors plant researchers can have a higher degree of assurance more low recalcitrance candidates can be identified. Significant changes in plate reactor, control biomass preparation, pretreatment conditions and enzyme have significantly reduced sample and control replicate variability. Reactor plate metallurgy significantly impacts sugar release aluminum leaching into reaction during pretreatment degrades sugars and inhibits enzyme activity. Removal of starch and extractives significantly decreases control biomass variability. New enzyme formulations give more consistent and higher conversion levels, however required re-optimization for switchgrass. Pretreatment time and temperature (severity) should be adjusted to specific biomass types i.e. woody vs. herbaceous. Desalting of enzyme preps to remove low molecular weight stabilizers and improved conversion levels likely due to water activity impacts on enzyme structure and …

This fact provides information on the Strategic Technical Assistance Response Team (START) Program, a U.S. Department of Energy Office of Indian Energy Policy and Programs (DOE-IE) initiative to provide technical expertise to support the development of next-generation energy projects in Indian Country.

This fact sheet provides information on Tribes in the lower 48 states selected to receive assistance from the Strategic Technical Assistance Response Team (START) Program, a U.S. Department of Energy Office of Indian Energy Policy and Programs (DOE-IE) initiative to provide technical expertise to support the development of next-generation energy projects in Indian Country.

This fact sheet provides information on the Alaska Native governments selected to receive assistance from the Strategic Technical Assistance Response Team (START) Program, a U.S. Department of Energy Office of Indian Energy Policy and Programs (DOE-IE) initiative to provide technical expertise to support the development of next-generation energy projects in Indian Country.

This report will summarize the home energy improvements performed in the Atlanta, GA area. In total, nine homes were retrofitted with eight of the homes having predicted source energy savings of approximately 30% or greater based on simulated energy consumption.

This is a final report of the Advancing Residential Retrofits in the Mixed Humid Climate to Achieve Deep Energy Savings.

Test data are reported that demonstrate the deposition from a spray dispersion system (Illinois Tool Works inductively charging rotary atomization nozzle) for application of decontamination solution to various surfaces in the passenger cabin of a Boeing 737 aircraft. The decontamination solution (EnviroTru) was tagged with a known concentration of fluorescein permitting determination of both airborne decontaminant concentration and surface deposited decontaminant solution so that the effective deposition rates and surface coverage could be determined and correlated with the amount of material sprayed. Six aerosol dispersion tests were conducted. In each test, aluminum foil deposition coupons were set out throughout the passenger area and the aerosol was dispersed. The aerosol concentration was measured with filter samplers as well as with optical techniques Average aerosol deposition ranged from 3 to 15 grams of decontamination solution per square meter. Some disagreement was observed between various instruments utilizing different measurement principles. These results demonstrate a potentially effective method to disperse decontaminant to interior surfaces of a passenger aircraft.

The Advanced Test Reactor (ATR) Full size plate In center flux trap Position (AFIP) experiment AFIP-3 was designed to evaluate the performance of monolithic fuels at a prototypic scale of 2.25 inches x 21.5 inches x 0.050 inches (5.75 cm x 54.6 cm x 0.13cm). The AFIP-3 experiment was fabricated by hot isostatic pressing (HIP) and consists of two plates, one with a zirconium (Zr) diffusion barrier and one with a silicon (Si) enhanced fuel/clad interface1,2. The following report summarizes the life of the AFIP-3 experiment through end of irradiation, including a brief description of the safety analysis, as-run neutronic analysis results, hydraulic testing results, and thermal analysis results.

The Advanced Test Reactor (ATR) Full size plate In center flux trap Position (AFIP) experiment AFIP-4 was designed to evaluate the performance of monolithic uranium-molybdenum (U-Mo) fuels at a scale prototypic of research reactor fuel plates. The AFIP-4 test further examine the fuel/clad interface and its behavior under extreme conditions. After irradiation, fission gas retention measurements will be performed during post irradiation (PIE)1,2. The following report summarizes the life of the AFIP-4 experiment through end of irradiation, including a brief description of the safety analysis, as-run neutronic analysis results, hydraulic testing results, and thermal analysis results.

Due to the incompletion of the Advanced Test React

The first fuel plate frame assembly of the AFIP-6 MKII experiment was irradiated as planned from December, 2011 through February, 2012 in the center flux trap of the Advanced Test Reactor during cycle 151A. Following irradiation in this cycle and while reconfiguring the experiment in the ATR canal, a non-fueled component (the bottom plate) of the first fuel plate frame assembly became separated from the rail sides. There is no evidence that the fueled region of the fuel plate frame assembly was compromised by this incident or the irradiation conditions. The separation of this component was determined to have been caused by flow induced vibrations, where vortex shedding frequencies were resonant with a natural frequency of the bottom plate component. This gave way to amplification, fracture, and separation from the assembly. Although parallel flow induced vibrations were analyzed, vortex shedding flow induced vibrations was an unfamiliar failure mode that was difficult to identify. Both the once-irradiated first fuel plate and un-irradiated second fuel plate frame assemblies were planned for irradiation in the subsequent cycle 151B. The AFIP-6 MKII experiment was excluded from irradiation in cycle 151B because non-trivial design modifications would be needed to mitigate this type of incident during …

The Advanced Test Reactor (ATR) Full size plate In

The NGNP Graphite R&D program is currently establishing the safe operating envelope of graphite core components for a Very High Temperature Reactor (VHTR) design. The program is generating quantitative data necessary for predicting the behavior and operating performance of the new nuclear graphite grades. To determine the in-service behavior of the graphite for pebble bed and prismatic designs, the Advanced Graphite Creep (AGC) experiment is underway. This experiment is examining the properties and behavior of nuclear grade graphite over a large spectrum of temperatures, neutron fluences and compressive loads. Each experiment consists of over 400 graphite specimens that are characterized prior to irradiation and following irradiation. Six experiments are planned with the first, AGC-1, currently being irradiated in the Advanced Test Reactor (ATR) and pre-irradiation characterization of the second, AGC-2, completed. This data package establishes the readiness of 512 specimens for assembly into the AGC-2 capsule.

The Graphite Technology Development Program will run a series of six experiments to quantify the effects of irradiation on nuclear grade graphite. The second Advanced Graphite Creep (AGC) experiment (AGC-2) began with Advanced Test Reactor (ATR) Cycle 149A on April 12, 2011, and ended with ATR Cycle 151B on May 5, 2012. The purpose of this report is to qualify AGC-2 irradiation monitoring data following INL Management and Control Procedure 2691, Data Qualification. Data that are Qualified meet the requirements for data collection and use as described in the experiment planning and quality assurance documents. Data that do not meet the requirements are Failed. Some data may not quite meet the requirements, but may still provide some useable information. These data are labeled as Trend. No Trend data were identified for the AGC-2 experiment. All thermocouples functioned throughout the AGC-2 experiment. There was one instance where spurious signals or instrument power interruption resulted in a recorded temperature value being well outside physical reality. This value was identified and labeled as Failed data. All other temperature data are Qualified. All helium and argon gas flow data are within expected ranges. Total gas flow was approximately 50 sccm through the capsule. Helium …

The implementation of automated regulatory control has been around since the middle of the last century through analog means. It has allowed engineers to operate the plant more consistently by focusing on overall operations and settings instead of individual monitoring of local instruments (inside and outside of a control room). A similar approach is proposed for cyber security, where current border-protection designs have been inherited from information technology developments that lack consideration of the high-reliability, high consequence nature of industrial control systems. Instead of an independent development, however, an integrated approach is taken to develop a holistic understanding of performance. This performance takes shape inside a multiagent design, which provides a notional context to model highly decentralized and complex industrial process control systems, the nervous system of critical infrastructure. The resulting strategy will provide a framework for researching solutions to security and unrecognized interdependency concerns with industrial control systems.

Destructive post-irradiation examination was performed on fuel Compact 6-3-2, which was irradiated in the AGR-1 experiment to a final compact average burnup of 11.3% FIMA and a time-average, volume-average temperature of 1070°C. The analysis of this compact was focused on characterizing the extent of fission product release from the particles and examining particles to determine the condition of the kernels and coating layers. The work included deconsolidation of the compact and leach-burn-leach analysis, visual inspection and gamma counting of individual particles, measurement of fuel burnup by several methods, metallurgical preparation of selected particles, and examination of particle cross-sections with optical microscopy. A single particle with a defective SiC layer was identified during deconsolidation-leach-burn-leach analysis, which is in agreement with previous measurements showing elevated cesium in the Capsule 6 graphite fuel holder associated with this fuel compact. The fraction of the compact europium inventory released from the particles and retained in the matrix was relatively high (approximately 6E-3), indicating release from intact particle coatings. The Ag-110m inventory in individual particles exhibited a very broad distribution, with some particles retaining =80% of the predicted inventory and others retaining less than 25%. The average degree of Ag-110m retention in 60 gamma counted particles …

This document presents the as-run analysis of the AGR-1 irradiation experiment. AGR-1 is the first of eight planned irradiations for the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. Funding for this program is provided by the US Department of Energy (DOE) as part of the Next-Generation Nuclear Plant (NGNP) project. The objectives of the AGR-1 experiment are: 1. To gain experience with multi-capsule test train design, fabrication, and operation with the intent to reduce the probability of capsule or test train failure in subsequent irradiation tests. 2. To irradiate fuel produced in conjunction with the AGR fuel process development effort. 3. To provide data that will support the development of an understanding of the relationship between fuel fabrication processes, fuel product properties, and irradiation performance. In order to achieve the test objectives, the AGR-1 experiment was irradiated in the B-10 position of the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) for a total duration of 620 effective full power days of irradiation. Irradiation began on December 24, 2006 and ended on November 6, 2009 spanning 13 ATR cycles and approximately three calendar years. The test contained six independently controlled and monitored capsules. Each capsule contained …

The PARFUME modeling code was used to predict failure probability of TRISO-coated fuel particles and diffusion of fission products through these particles during safety tests following the first irradiation test of the Advanced Gas Reactor program (AGR-1). These calculations support the AGR-1 Safety Testing Experiment, which is part of the PIE effort on AGR-1. Modeling of the AGR-1 Safety Test Predictions includes a 620-day irradiation followed by a 300-hour heat-up phase of selected AGR-1 compacts. Results include fuel failure probability, palladium penetration, and fractional release of fission products. Results show that no particle failure is predicted during irradiation or heat-up, and that fractional release of fission products is limited during irradiation but that it significantly increases during heat-up.