Presenters will discuss the latest advances in plant and algal lipid metabolism, oil synthesis, lipid signaling, lipid visualization, lipid biotechnology and its applications, the physiological and developmental roles of lipids, and plant lipids in health. Sessions include: Producing Nutritional Lipids; Metabolic biochemistry in the next decade; Triacylglycerols: Metabolism, function, and as a target for engineering; Lipids in Protection, Reproduction, and Development; Genetic and Lipidomic Approaches to Understanding Lipid Metabolism and Signaling; Lipid Signaling in Stress Responses; New Insights on the Path to Triacylglycerols; Membrane Lipid Signaling; Lipid Visualization; Development of Biofuels and Industrial Lipids.

Reliability/safety analysis of stochastic dynamic systems (e.g., nuclear power plants, airplanes, chemical plants) is currently performed through a combination of Event-Tress and Fault-Trees. However, these conventional methods suffer from certain drawbacks: • Timing of events is not explicitly modeled • Ordering of events is preset by the analyst • The modeling of complex accident scenarios is driven by expert-judgment For these reasons, there is currently an increasing interest into the development of dynamic PRA methodologies since they can be used to address the deficiencies of conventional methods listed above.

Stripline BPM sensors contain inherent non-linearities, as a result of field distortions from the pickup elements. Many methods have been devised to facilitate corrections, often employing polynomial fitting. The cost of computation makes real-time correction difficult, particulalry when integer math is utilized. The application of neural-network technology, particularly the multi-layer perceptron algorithm, is proposed as an efficient alternative for electrode linearization. A process of supervised learning is initially used to determine the weighting coefficients, which are subsequently applied to the incoming electrode data. A non-linear layer, known as an “activation layer,” is responsible for the removal of saturation effects. Implementation of a perceptron in an FPGA-based software-defined radio (SDR) is presented, along with performance comparisons. In addition, efficient calculation of the sigmoidal activation function via the CORDIC algorithm is presented.

Article about the Formula One United States Grand Prix event held at Circuit of the Americas in Austin, Texas.

In 2007, the FAST wind turbine simulation tool, developed and maintained by the U.S. Department of Energy's (DOE's) National Renewable Energy Laboratory (NREL), was expanded to include capabilities that are suitable for modeling floating offshore wind turbines. In an effort to validate FAST and other offshore wind energy modeling tools, DOE funded the DeepCwind project that tested three prototype floating wind turbines at 1/50th scale in a wave basin, including a semisubmersible, a tension-leg platform, and a spar buoy. This paper describes the use of the results of the spar wave basin tests to calibrate and validate the FAST offshore floating simulation tool, and presents some initial results of simulated dynamic responses of the spar to several combinations of wind and sea states.

This report describes Charmonium mass splittings at the physical point.

Since 1981, the United States (U.S.) Nuclear Regulatory Commission (NRC) [1] requires a plant- specific simulator facility for use in training at U.S. nuclear power plants (NPPs). These training simulators are in near constant use for training and qualification of licensed NPP operators. In the early 1980s, the Halden Man-Machine Laboratory (HAMMLab) at the Halden Reactor Project (HRP) in Norway first built perhaps the most well known set of research simulators. The HRP offered a high- fidelity simulator facility in which the simulator is functionally linked to a specific plant but in which the human-machine interface (HMI) may differ from that found in the plant. As such, HAMMLab incorporated more advanced digital instrumentation and controls (I&C) than the plant, thereby giving it considerable interface flexibility that researchers took full advantage of when designing and validating different ways to upgrade NPP control rooms. Several U.S. partners—the U.S. NRC, the Electrical Power Research Institute (EPRI), Sandia National Laboratories, and Idaho National Laboratory (INL) – as well as international members of the HRP, have been working with HRP to run control room simulator studies. These studies, which use crews from Scandinavian plants, are used to determine crew behavior in a variety of normal …

This article analyzes dimensional changes due to irradiation of monolithic plate-type nuclear fuel and compares results with finite element analysis of the plates during fabrication and irradiation. Monolithic fuel plates tested in the Advanced Test Reactor (ATR) at Idaho National Lab (INL) are being used to benchmark proposed fuel performance for several high power research reactors. Post-irradiation metallographic images of plates sectioned at the midpoint were analyzed to determine dimensional changes of the fuel and the cladding response. A constitutive model of the fabrication process and irradiation behavior of the tested plates was developed using the general purpose commercial finite element analysis package, Abaqus. Using calculated burn-up profiles of irradiated plates to model the power distribution and including irradiation behaviors such as swelling and irradiation enhanced creep, model simulations allow analysis of plate parameters that are either impossible or infeasible in an experimental setting. The development and progression of fabrication induced stress concentrations at the plate edges was of primary interest, as these locations have a unique stress profile during irradiation. Additionally, comparison between 2D and 3D models was performed to optimize analysis methodology. In particular, the ability of 2D and 3D models account for out of plane stresses which …

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This article presents evaluation of the stress-str

As the level of wind penetration increases, wind turbine technology must move from merely generating power from wind to taking a role in supporting the bulk power system. Wind turbines should have the capability to provide inertial response and primary frequency (governor) response so they can support the frequency stability of the grid. To provide governor response, wind turbines should be able to generate less power than the available wind power and hold the rest in reserve, ready to be accessed as needed. This paper explores several ways to control wind turbine output to enable reserve-holding capability. This paper focuses on fixed-speed (also known as Type 1) and variable-slip (also known as Type 2) turbines.

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This paper presents a description of The Fukushima Daiichi Accident Study Information Portal. The Information Portal was created by the Idaho National Laboratory as part of joint NRC and DOE project to assess the severe accident modeling capability of the MELCOR analysis code. The Fukushima Daiichi Accident Study Information Portal was created to collect, store, retrieve and validate information and data for use in reconstructing the Fukushima Daiichi accident. In addition to supporting the MELCOR simulations, the Portal will be the main DOE repository for all data, studies and reports related to the accident at the Fukushima Daiichi nuclear power station. The data is stored in a secured (password protected and encrypted) repository that is searchable and accessible to researchers at diverse locations.

Grain boundary chemistry in an X750 Ni alloy was analyzed by atom probe tomography in an effort to clarify the possible roles of elemental segregation and carbide presence on the stress corrosion cracking behavior of Ni alloys. Two types of cracks are observed: straight cracks along twin boundaries and wavy cracks at general boundaries. It was found that carbides (M23C6 and TiC) are present at both twin and general boundaries, with comparable B and P segregation for all types of grain boundaries. Twin boundaries intercept ?’ precipitates while the general boundaries wave around the ?’ and carbide precipitates. Near a crack tip, oxidation takes place on the periphery of carbide precipitate.

INTRODUCTION A series of experiments was carried out in 1962-65 at the Oak Ridge National Laboratory Critical Experiments Facility (ORCEF) for use in space reactor research programs. A core containing 93.2 wt% enriched UO2 fuel rods was used in these experiments. The first part of the experimental series consisted of 252 tightly-packed fuel rods (1.27-cm triangular pitch) with graphite reflectors [1], the second part used 252 graphite-reflected fuel rods organized in a 1.506-cm triangular-pitch array [2], and the final part of the experimental series consisted of 253 beryllium-reflected fuel rods in a 1.506-cm-triangular-pitch configuration and in a 7-tube-cluster configuration [3]. Fission rate distribution and cadmium ratio measurements were taken for all three parts of the experimental series. Reactivity coefficient measurements were taken for various materials placed in the beryllium reflected core. All three experiments in the series have been evaluated for inclusion in the International Reactor Physics Experiment Evaluation Project (IRPhEP) [4] and the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbooks, [5]. The evaluation of the first experiment in the series was discussed at the 2011 ANS Winter meeting [6]. The evaluations of the second and third experiments are discussed below. These experiments are of interest as benchmarks because …

Uncertainty and sensitivity studies are essential elements of the reactor simulation code verification and validation process. Although several international uncertainty quantification activities have been launched in recent years in the LWR, BWR and VVER domains (e.g. the OECD/NEA BEMUSE program [1], from which the current OECD/NEA LWR Uncertainty Analysis in Modelling (UAM) benchmark [2] effort was derived), the systematic propagation of uncertainties in cross-section, manufacturing and model parameters for High Temperature Reactor (HTGR) designs has not been attempted yet. This paper summarises the scope, objectives and exercise definitions of the IAEA Coordinated Research Project (CRP) on HTGR UAM [3]. Note that no results will be included here, as the HTGR UAM benchmark was only launched formally in April 2012, and the specification is currently still under development.

Physically measuring the dynamic responses of wind turbine support structures enables the calculation of the applied loads using an inverse procedure. In this process, inverse means deriving the inputs/forces from the outputs/responses. This paper presents results of a numerical verification of such an inverse load calculation. For this verification, the comprehensive simulation code FAST is used. FAST accounts for the coupled dynamics of wind inflow, aerodynamics, elasticity and turbine controls. Simulations are run using a 5-MW onshore wind turbine model with a tubular tower. Both the applied loads due to the instantaneous wind field and the resulting system responses are known from the simulations. Using the system responses as inputs to the inverse calculation, the applied loads are calculated, which in this case are the rotor thrust forces. These forces are compared to the rotor thrust forces known from the FAST simulations. The results of these comparisons are presented to assess the accuracy of the inverse calculation. To study the influences of turbine controls, load cases in normal operation between cut-in and rated wind speed, near rated wind speed and between rated and cut-out wind speed are chosen. The presented study shows that the inverse load calculation is capable of …

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The potential for unintended consequences of biofuels--competition for land and water--necessitates that sustainable biofuel expansion considers the complexities of resource requirements within specific context (e.g., technology, feedstock, supply chain, local resource availability).

A prismatic gas-cooled very high temperature reactor (VHTR) is being developed under the next generation nuclear plant program (NGNP) of the U.S. Department of Energy, Office of Nuclear Energy. In the design of the prismatic VHTR, hexagonal shaped graphite blocks are drilled to allow insertion of fuel pins, made of compacted TRISO fuel particles, and coolant channels for the helium coolant. One of the concerns for the reactor design is the effects of a loss of flow accident (LOFA) where the coolant circulators are lost for some reason, causing a loss of forced coolant flow through the core. In such an event, it is desired to know what happens to the (reduced) heat still being generated in the core and if it represents a problem for the fuel compacts, the graphite core or the reactor vessel (RV) walls. One of the mechanisms for the transport of heat out of the core is by the natural circulation of the coolant, which is still present. That is, how much heat may be transported by natural circulation through the core and upwards to the top of the upper plenum? It is beyond current capability for a computational fluid dynamic (CFD) analysis to perform …

This article reports on the development and comparison of two maps of Texas related to Rocky Mountain spotted fever.

The U.S. Nuclear Regulatory Commission’s (USNRC) Standardized Plant Analysis Risk (SPAR) Level 2 models for U.S. commercial nuclear power plants has historically used a partitioning approach for plant damage state (PDS) binning and model quantification since late 1990s [1]. While this approach has the advantage to be able to identify the details of the severe accident sequences with one or more individual PDS vector characters, the Level 2 model quantification process is tedious and error-prone with multiple steps involved. A new approach to quantify Level 2 SPAR models was recently developed and implemented in the latest SAPHIRE Version 8 [2]. The new approach removes the partition rules and greatly simplifies the quantification process.

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The International Reactor Physics Experiment Evaluation Project (IRPhEP) was established to preserve integral reactor physics experimental data, including separate or special effects data for nuclear energy and technology applications. Numerous experiments that have been performed worldwide, represent a large investment of infrastructure, expertise, and cost, and are valuable resources of data for present and future research. These valuable assets provide the basis for recording, development, and validation of methods. If the experimental data are lost, the high cost to repeat many of these measurements may be prohibitive. The purpose of the IRPhEP is to provide an extensively peer-reviewed set of reactor physics-related integral data that can be used by reactor designers and safety analysts to validate the analytical tools used to design next-generation reactors and establish the safety basis for operation of these reactors. Contributors from around the world collaborate in the evaluation and review of selected benchmark experiments for inclusion in the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook) [1]. Several new evaluations have been prepared for inclusion in the March 2012 edition of the IRPhEP Handbook.