Mandelli, Diego; Rabiti, Cristian; Cogliati, Joshua Joseph; Kinoshita, Robert Arthur; Alfonsi, Andrea; Sen, Ramazan Sonat
RAVEN is a generic software framework to perform parametric and probabilistic analy- sis based on the response of complex system codes. The initial development was aimed to provide dynamic risk analysis capabilities to the Thermo-Hydraulic code RELAP-7, currently under development at the Idaho National Laboratory (INL). Although the initial goal has been fully accomplished, RAVEN is now a multi-purpose probabilistic and uncer- tainty quantification platform, capable to agnostically communicate with any system code. This agnosticism includes providing Application Programming Interfaces (APIs). These APIs are used to allow RAVEN to interact with any code as long as all the parameters that need to be perturbed are accessible by inputs files or via python interfaces. RAVEN is capable of investigating the system response, and investigating the input space using Monte Carlo, Grid, or Latin Hyper Cube sampling schemes, but its strength is focused to- ward system feature discovery, such as limit surfaces, separating regions of the input space leading to system failure, using dynamic supervised learning techniques. The development of RAVEN has started in 2012, when, within the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program, the need to provide a modern risk evaluation framework became stronger. RAVEN principal assignment is to provide the necessary software and algorithms in order to employ the concept developed by the Risk Informed Safety Margin Characterization (RISMC) program. RISMC is one of the pathways defined within the Light Water Reactor Sustainability (LWRS) program. In the RISMC approach, the goal is not just the individuation of the frequency of an event potentially leading to a system failure, but the closeness (or not) to key safety-related events. Hence, the approach is in- terested in identifying and increasing the safety margins related to those events. A safety margin is a numerical value quantifying the probability that a safety metric (e.g. for an important process such as peak pressure in a pipe) is exceeded under certain conditions. The initial development of RAVEN has been focused on providing dynamic risk assess- ment capability to RELAP-7, currently under develop-ment at the INL and, likely, future replacement of the RELAP5-3D code. Most the capabilities that have been implemented having RELAP-7 as principal focus are easily deployable for other system codes. For this reason, several side activates are currently ongoing for coupling RAVEN with soft- ware such as RELAP5-3D, etc. The aim of this document is the explaination of the input requirements, focalizing on the input structure.
Cogliati, Joshua J.; Rabiti, Cristian; Permann, Cody J.
The milestones have been achieved. RAVEN has been migrated to Gitlab which adds new abilities for code review and management. Standalone RAVEN framework packages have been created for OSX and two Linux distributions.
Biofuels International Corporation Jump to: navigation, search Name: Raven Biofuels International Corporation Place: Paramus, New Jersey Zip: 07652-1236 Sector: Biofuels Product:...
A. Alfonsi; C. Rabiti; D. Mandelli; J. Cogliati; R. A. Kinoshita; A. Naviglio
Conventional Event-Tree (ET) based methodologies are extensively used as tools to perform reliability and safety assessment of complex and critical engineering systems. One of the disadvantages of these methods is that timing/sequencing of events and system dynamics is not explicitly accounted for in the analysis. In order to overcome these limitations several techniques, also know as Dynamic Probabilistic Risk Assessment (D-PRA), have been developed. Monte-Carlo (MC) and Dynamic Event Tree (DET) are two of the most widely used D-PRA methodologies to perform safety assessment of Nuclear Power Plants (NPP). In the past two years, the Idaho National Laboratory (INL) has developed its own tool to perform Dynamic PRA: RAVEN (Reactor Analysis and Virtual control ENvironment). RAVEN has been designed in a high modular and pluggable way in order to enable easy integration of different programming languages (i.e., C++, Python) and coupling with other application including the ones based on the MOOSE framework, developed by INL as well. RAVEN performs two main tasks: 1) control logic driver for the new Thermo-Hydraulic code RELAP-7 and 2) post-processing tool. In the first task, RAVEN acts as a deterministic controller in which the set of control logic laws (user defined) monitors the RELAP-7 simulation and controls the activation of specific systems. Moreover, RAVEN also models stochastic events, such as components failures, and performs uncertainty quantification. Such stochastic modeling is employed by using both MC and DET algorithms. In the second task, RAVEN processes the large amount of data generated by RELAP-7 using data-mining based algorithms. This paper focuses on the first task and shows how it is possible to perform the analysis of dynamic stochastic systems using the newly developed RAVEN DET capability. As an example, the Dynamic PRA analysis, using Dynamic Event Tree, of a simplified pressurized water reactor for a Station Black-Out scenario is presented.
A. Alfonsi; C. Rabiti; D. Mandelli; J. Cogliati; R. Kinoshita
The Reactor Analysis and Virtual control ENviroment (RAVEN) code is a software tool that acts as the control logic driver and post-processing engine for the newly developed Thermal-Hydraulic code RELAP-7. RAVEN is now a multi-purpose Probabilistic Risk Assessment (PRA) software framework that allows dispatching different functionalities: Derive and actuate the control logic required to simulate the plant control system and operator actions (guided procedures), allowing on-line monitoring/controlling in the Phase Space Perform both Monte-Carlo sampling of random distributed events and Dynamic Event Tree based analysis Facilitate the input/output handling through a Graphical User Interface (GUI) and a post-processing data mining module
Swiler, Laura Painton; Mandelli, Diego; Rabiti, Cristian; Alfonsi, Andrea
This report summarizes the result of a NEAMS project focused on the use of reliability methods within the RAVEN and RELAP-7 software framework for assessing failure probabilities as part of probabilistic risk assessment for nuclear power plants. RAVEN is a software tool under development at the Idaho National Laboratory that acts as the control logic driver and post-processing tool for the newly developed Thermal-Hydraulic code RELAP-7. Dakota is a software tool developed at Sandia National Laboratories containing optimization, sensitivity analysis, and uncertainty quantification algorithms. Reliability methods are algorithms which transform the uncertainty problem to an optimization problem to solve for the failure probability, given uncertainty on problem inputs and a failure threshold on an output response. The goal of this work is to demonstrate the use of reliability methods in Dakota with RAVEN/RELAP-7. These capabilities are demonstrated on a demonstration of a Station Blackout analysis of a simplified Pressurized Water Reactor (PWR).
Cristian Rabiti; Andrea Alfonsi; Joshua Cogliati; Diego Mandelli; Robert Kinoshita
RAVEN is a generic software driver to perform parametric and probabilistic analysis of code simulating complex systems. Initially developed to provide dynamic risk analysis capabilities to the RELAP-7 code  is currently being generalized with the addition of Application Programming Interfaces (APIs). These interfaces are used to extend RAVEN capabilities to any software as long as all the parameters that need to be perturbed are accessible by inputs files or directly via python interfaces. RAVEN is capable to investigate the system response probing the input space using Monte Carlo, grid strategies, or Latin Hyper Cube schemes, but its strength is its focus toward system feature discovery like limit surfaces separating regions of the input space leading to system failure using dynamic supervised learning techniques. The paper will present an overview of the software capabilities and their implementation schemes followed by same application examples.
Cristian Rabiti; Paul Talbot; Andrea Alfonsi; Diego Mandelli; Joshua Cogliati
RAVEN, under the support of the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program, has been tasked to provide the necessary software and algorithms to enable the application of the conceptual framework developed by the Risk Informed Safety Margin Characterization (RISMC)  path. RISMC is one of the paths defined under the Light Water Reactor Sustainability (LWRS) DOE program.
Andrea Alfonsi; Cristian Rabiti; Diego Mandelli; Joshua Cogliati; Robert Kinoshita; Antonio Naviglio
RAVEN is a generic software framework to perform parametric and probabilistic analysis based on the response of complex system codes. The initial development was aimed to provide dynamic risk analysis capabilities to the Thermo-Hydraulic code RELAP-7 , currently under development at the Idaho National Laboratory. Although the initial goal has been fully accomplished, RAVEN is now a multi-purpose probabilistic and uncertainty quantification platform, capable to agnostically communicate with any system code. This agnosticism has been employed by providing Application Programming Interfaces (APIs). These interfaces are used to allow RAVEN to interact with any code as long as all the parameters that need to be perturbed are accessible by inputs files or via python interfaces. RAVEN is capable to investigate the system response, investigating the input space using Monte Carlo, Grid, or Latin Hyper Cube sampling schemes, but its strength is focused toward system feature discovery, such as limit surfaces, separating regions of the input space leading to system failure, using dynamic supervised learning techniques. The paper presents an overview of the software capabilities and their implementation schemes followed by some application examples.
Cristian Rabiti; Andrea Alfonsi; Joshua Cogliati; Diego Mandelli; Robert Kinoshita
RAVEN is a complex software tool that will have tasks spanning from being the RELAP-7 user interface, to using RELAP-7 to perform Risk Informed Safety Characterization (RISMC), and to controlling RELAP-7 calculation execution. The goal of this document is to: 1. Highlight the functional requirements of the different tasks of RAVEN 2. Identify shared functions that could be aggregate in modules so to obtain a minimal software redundancy and maximize software utilization. RAVEN is in fact a software framework that will allow exploiting the following functionalities: • Derive and actuate the control logic required to: o Simulate the plant control system o Simulate the operator (procedure guided) actions o Perform Monte Carlo sampling of random distributed events o Perform event three based analysis • Provide a GUI to: o Input a plant description to RELAP-7 (component, control variable, control parameters) o Concurrent monitoring of Control Parameters o Concurrent alteration of control parameters • Provide Post Processing data mining capability based on o Dimensionality reduction o Cardinality reduction In this document it will be shown how an appropriate mathematical formulation of the control logic and probabilistic analysis leads to have most of the software infrastructure leveraged between the two main tasks. Further, this document will go through the development accomplished this year, including simulation results, and priorities for the next years development
diego mandelli; curtis smith; cristian rabiti; andrea alfonsi
This paper presents the solution of the space propulsion problem using a PRA code currently under development at Idaho National Laboratory (INL). RAVEN (Reactor Analysis and Virtual control ENviroment) is a multi-purpose Probabilistic Risk Assessment (PRA) software framework that allows dispatching different functionalities. It is designed to derive and actuate the control logic required to simulate the plant control system and operator actions (guided procedures) and to perform both Monte- Carlo sampling of random distributed events and Event Tree based analysis. In order to facilitate the input/output handling, a Graphical User Interface (GUI) and a post-processing data-mining module are available. RAVEN allows also to interface with several numerical codes such as RELAP5 and RELAP-7 and ad-hoc system simulators. For the space propulsion system problem, an ad-hoc simulator has been developed and written in python language and then interfaced to RAVEN. Such simulator fully models both deterministic (e.g., system dynamics and interactions between system components) and stochastic behaviors (i.e., failures of components/systems such as distribution lines and thrusters). Stochastic analysis is performed using random sampling based methodologies (i.e., Monte-Carlo). Such analysis is accomplished to determine both the reliability of the space propulsion system and to propagate the uncertainties associated to a specific set of parameters. As also indicated in the scope of the benchmark problem, the results generated by the stochastic analysis are used to generate risk-informed insights such as conditions under witch different strategy can be followed.
Rabiti, Cristian; Alfonsi, Andrea; Mandelli, Diego; Cogliati, Joshua; Kinoshita, Robert
RAVEN, under the support of the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program , is advancing its capability to perform statistical analyses of stochastic dynamic systems. This is aligned with its mission to provide the tools needed by the Risk Informed Safety Margin Characterization (RISMC) path-lead  under the Department Of Energy (DOE) Light Water Reactor Sustainability program . In particular this task is focused on the synergetic development with the RELAP-7  code to advance the state of the art on the safety analysis of nuclear power plants (NPP). The investigation of the probabilistic evolution of accident scenarios for a complex system such as a nuclear power plant is not a trivial challenge. The complexity of the system to be modeled leads to demanding computational requirements even to simulate one of the many possible evolutions of an accident scenario (tens of CPU/hour). At the same time, the probabilistic analysis requires thousands of runs to investigate outcomes characterized by low probability and severe consequence (tail problem). The milestone reported in June of 2013  described the capability of RAVEN to implement complex control logic and provide an adequate support for the exploration of the probabilistic space using a Monte Carlo sampling strategy. Unfortunately the Monte Carlo approach is ineffective with a problem of this complexity. In the following year of development, the RAVEN code has been extended with more sophisticated sampling strategies (grids, Latin Hypercube, and adaptive sampling). This milestone report illustrates the effectiveness of those methodologies in performing the assessment of the probability of core damage following the onset of a Station Black Out (SBO) situation in a boiling water reactor (BWR). The first part of the report provides an overview of the available probabilistic analysis capabilities, ranging from the different types of distributions available, possible sampling strategies, and post processing analysis capabilities. The first part of the report provides an extensive description of two major developments introduced this year: adaptive sampling for limit surface sampling and multi variate distributions. The document concludes with a description of the demo case (BWR-SBO) and a discussion of the results obtained.
Rabiti, C.; Mandelli, D.; Alfonsi, A.; Cogliati, J.; Kinoshita, R.
RAVEN (Reactor Analysis and Virtual control Environment) is a software code under development at Idaho National Laboratory aimed at performing probabilistic risk assessment and uncertainty quantification using RELAP-7, for which it acts also as a simulation controller. In this paper we will present the equations characterizing a dynamic stochastic system and we will then discuss the behavior of each stochastic term and how it is accounted for in the RAVEN software design. Moreover we will present preliminary results of the implementation. (authors)
C. Rabiti; D. Mandelli; J. Cogliati; R. Kinoshita
RAVEN (Reactor Analysis and Virtual control Environment) is a software code under development at Idaho National Laboratory aimed at performing probabilistic risk assessment and uncertainty quantification using RELAP-7, for which it acts also as a simulation controller. In this paper we will present the equations characterizing a dynamic stochastic system and we will then discuss the behavior of each stochastic term and how it is accounted for in the RAVEN software design. Moreover we will present preliminary results of the implementation.
Alfonsi, A.; Rabiti, C.; Mandelli, D.; Cogliati, J. J.; Kinoshita, R. A.
RAVEN is a software tool under development at the Idaho National Laboratory (INL) that acts as the control logic driver and post-processing tool for the newly developed Thermal-Hydraulic code RELAP-7. The scope of this paper is to show the software structure of RAVEN and its utilization in connection with RELAP-7. A short overview of the mathematical framework behind the code is presented along with its main capabilities such as on-line controlling/ monitoring and Monte-Carlo sampling. A demo of a Station Black Out PRA analysis of a simplified Pressurized Water Reactor (PWR) model is shown in order to demonstrate the Monte-Carlo and clustering capabilities. (authors)
Alfonsi Andrea; Mandelli Diego; Rabiti Cristian; Joshua Cogliati; Robert Kinoshita
RAVEN is a software tool under development at the Idaho National Laboratory (INL) that acts as the control logic driver and post-processing tool for the newly developed Thermo-Hydraylic code RELAP- 7. The scope of this paper is to show the software structure of RAVEN and its utilization in connection with RELAP-7. A short overview of the mathematical framework behind the code is presented along with its main capabilities such as on-line controlling/monitoring and Monte-Carlo sampling. A demo of a Station Black Out PRA analysis of a simplified Pressurized Water Reactor (PWR) model is shown in order to demonstrate the Monte-Carlo and clustering capabilities.
C. Rabiti; D. Mandelli; A. Alfonsi; J. Cogliati; R. Kinoshita; D. Gaston; R. Martineau; C. Curtis
Increases in computational power and pressure for more accurate simulations and estimations of accident scenario consequences are driving the need for Dynamic Probabilistic Risk Assessment (PRA)  of very complex models. While more sophisticated algorithms and computational power address the back end of this challenge, the front end is still handled by engineers that need to extract meaningful information from the large amount of data and build these complex models. Compounding this problem is the difficulty in knowledge transfer and retention, and the increasing speed of software development. The above-described issues would have negatively impacted deployment of the new high fidelity plant simulator RELAP-7 (Reactor Excursion and Leak Analysis Program) at Idaho National Laboratory. Therefore, RAVEN that was initially focused to be the plant controller for RELAP-7 will help mitigate future RELAP-7 software engineering risks. In order to accomplish this task, Reactor Analysis and Virtual Control Environment (RAVEN) has been designed to provide an easy to use Graphical User Interface (GUI) for building plant models and to leverage artificial intelligence algorithms in order to reduce computational time, improve results, and help the user to identify the behavioral pattern of the Nuclear Power Plants (NPPs). In this paper we will present the GUI implementation and its current capability status. We will also introduce the support vector machine algorithms and show our evaluation of their potentiality in increasing the accuracy and reducing the computational costs of PRA analysis. In this evaluation we will refer to preliminary studies performed under the Risk Informed Safety Margins Characterization (RISMC) project of the Light Water Reactors Sustainability (LWRS) campaign . RISMC simulation needs and algorithm testing are currently used as a guidance to prioritize RAVEN developments relevant to PRA.
Mandelli, Diego; Rabiti, Cristian; Cogliati, Joshua; Alfonsi, Andrea; Askin Guler; Tunc Aldemir
Passive system, structure and components (SSCs) will degrade over their operation life and this degradation may cause to reduction in the safety margins of a nuclear power plant. In traditional probabilistic risk assessment (PRA) using the event-tree/fault-tree methodology, passive SSC failure rates are generally based on generic plant failure data and the true state of a specific plant is not reflected realistically. To address aging effects of passive SSCs in the traditional PRA methodology  does consider physics based models that account for the operating conditions in the plant, however,  does not include effects of surveillance/inspection. This paper represents an overall methodology for the incorporation of aging modeling of passive components into the RAVEN/RELAP-7 environment which provides a framework for performing dynamic PRA. Dynamic PRA allows consideration of both epistemic and aleatory uncertainties (including those associated with maintenance activities) in a consistent phenomenological and probabilistic framework and is often needed when there is complex process/hardware/software/firmware/ human interaction . Dynamic PRA has gained attention recently due to difficulties in the traditional PRA modeling of aging effects of passive components using physics based models and also in the modeling of digital instrumentation and control systems. RAVEN (Reactor Analysis and Virtual control Environment)  is a software package under development at the Idaho National Laboratory (INL) as an online control logic driver and post-processing tool. It is coupled to the plant transient code RELAP-7 (Reactor Excursion and Leak Analysis Program) also currently under development at INL , as well as RELAP 5 . The overall methodology aims to: • Address multiple aging mechanisms involving large number of components in a computational feasible manner where sequencing of events is conditioned on the physical conditions predicted in a simulation environment such as RELAP-7. • Identify the risk-significant passive components, their failure modes and anticipated rates of degradation • Incorporate surveillance and maintenance activities and their effects into the plant state and into component aging progress. • Asses aging affects in a dynamic simulation environment 1. C. L. SMITH, V. N. SHAH, T. KAO, G. APOSTOLAKIS, “Incorporating Ageing Effects into Probabilistic Risk Assessment –A Feasibility Study Utilizing Reliability Physics Models,” NUREG/CR-5632, USNRC, (2001). 2. T. ALDEMIR, “A Survey of Dynamic Methodologies for Probabilistic Safety Assessment of Nuclear Power Plants, Annals of Nuclear Energy, 52, 113-124, (2013). 3. C. RABITI, A. ALFONSI, J. COGLIATI, D. MANDELLI and R. KINOSHITA “Reactor Analysis and Virtual Control Environment (RAVEN) FY12 Report,” INL/EXT-12-27351, (2012). 4. D. ANDERS et.al, "RELAP-7 Level 2 Milestone Report: Demonstration of a Steady State Single Phase PWR Simulation with RELAP-7," INL/EXT-12-25924, (2012).
Diego Mandelli; Curtis Smith; Thomas Riley; John Schroeder; Cristian Rabiti; Aldrea Alfonsi; Joe Nielsen; Dan Maljovec; Bie Wang; Valerio Pascucci
The existing fleet of nuclear power plants is in the process of extending its lifetime and increasing the power generated. In order to evaluate the impact of these two factors on the safety of the plant, the Risk Informed Safety Margin Characterization (RISMC) project aims to provide insight to decision makers through a series of simulations of the plant dynamics for different initial conditions (e.g., probabilistic analysis and uncertainty quantification). This report focuses, in particular, on the impact of power uprate on the safety of a boiled water reactor system. The case study considered is a loss of off-site power followed by the loss of diesel generators, i.e., a station black out (SBO) event. Analysis is performed by using a thermo-hydraulic code, i.e. RELAP-5, and a stochastic analysis tool currently under development at INL, i.e. RAVEN. Starting from the event tree models contained in SAPHIRE, we built the input file for RELAP-5 that models in great detail system dynamics under SBO conditions. We also interfaced RAVEN with RELAP-5 so that it would be possible to run multiple RELAP-5 simulation runs by changing specific keywords of the input file. We both employed classical statistical tools, i.e. Monte-Carlo, and more advanced machine learning based algorithms to perform uncertainty quantification in order to quantify changes in system performance and limitations as a consequence of power uprate. We also employed advanced data analysis and visualization tools that helped us to correlate simulation outcome such as maximum core temperature with a set of input uncertain parameters. Results obtained gave a detailed overview of the issues associated to power uprate for a SBO accident scenario. We were able to quantify how timing of safety related events were impacted by a higher reactor core power. Such insights can provide useful material to the decision makers to perform risk-infomed safety margins management.
technology known as "AC-Direct," which seeks to overcome the limitations of inverters and synchronous generators for mobile, off-grid, and distributed power applications....
Denver, Colorado Zip: 80205 Region: Rockies Area Sector: Buildings Product: Efficient window and daylighting systems Website: www.ravenbrick.com Coordinates: 39.754373,...
C. Rabiti; A. Alfonsi; D. Mandelli; J. Cogliati; R. Martineau
The Next Generation of System Analysis Code (NGSAC)  aims to model and simulate the Nuclear Power Plant (NPP) thermo-hydraulic behavior with high level of accuracy. In this respect, Idaho National Laboratory (INL) is developing a NGSAC (known as RELAP-7) which will allow to model NPP responses for a set of accident scenarios (e.g., loss of off-site power).
Reveals the Assembly and Evolution of Complex Metalloenzymes summary written by Raven Hanna The potential for using biological enzymes to make hydrogen to use as a...
Fuel Cell Costs by Changing the Structure and Reactivity of Platinum summary written by Raven Hanna Hydrogen fuel cells are a green alternative to fossil fuels for powering...
Virginia Castlewood, Virginia Cleveland, Virginia Honaker, Virginia Lebanon, Virginia Raven, Virginia St. Paul, Virginia Retrieved from "http:en.openei.orgw...
Burlington, North Carolina Elon, North Carolina Gibsonville, North Carolina Glen Raven, North Carolina Graham, North Carolina Green Level, North Carolina Haw River, North...
Bluefield, Virginia Cedar Bluff, Virginia Claypool Hill, Virginia Pocahontas, Virginia Raven, Virginia Richlands, Virginia Tazewell, Virginia Retrieved from "http:en.openei.org...
Hycrete Public Energy Solutions Pure Energy Corporation PEC RLR Consultants LLC Raven Biofuels International Corporation Resource Energy Systems LLC Reunion Power LLC...
Registered Energy Companies in Paramus, New Jersey Pure Energy Corporation PEC Raven Biofuels International Corporation References US Census Bureau Incorporated place...
the official policy or position of Los Alamos National Laboratory. Resources To learn more about Phil and Monica Noll, visit the Phil and Monica Noll's Raven Mountain Images...
County, Maine.1 Registered Energy Companies in Brunswick, Maine Independence Wind LLC Raven Technology References US Census Bureau Incorporated place and minor civil division...
Potential Effects of Aerosol Source on Ocean Photosynthesis summary written by Raven ... Soluble iron can reach oceans by water currents carrying dusts or by air currents carrying ...
and nesting habitat for species such as Common Ravens, Ferruginous Hawks, and other raptors. Uncommon birds recorded along this route included a Sage Thrasher, and a...
5th congressional district BGA Engineering LLC Building Performance Equipment, Inc. HERA USA Inc formerly Ergenics Inc Pure Energy Corporation PEC Raven Biofuels International...
LabSpaces New Research Puts Theory of Water Structure on Thin Ice summary written by Raven Hanna Water has unusual and complex properties that make it especially well suited to...
Ologeez SciLink LabSpaces Finding the Crystal Structure of P-gp: A Protein that Makes Cancer Cells Resistant to Chemotherapy summary written by Raven Hanna Medications can be...
Cao, Y; Li, R; Chi, Z; Zhu, S
Purpose: Different treatment planning systems (TPS) use different treatment optimization and leaf sequencing algorithms. This work compares cervical carcinoma IMRT plans optimized with four commercial TPSs to investigate the plan quality in terms of target conformity and delivery efficiency. Methods: Five cervical carcinoma cases were planned with the Corvus, Monaco, Pinnacle and Xio TPSs by experienced planners using appropriate optimization parameters and dose constraints to meet the clinical acceptance criteria. Plans were normalized for at least 95% of PTV to receive the prescription dose (Dp). Dose-volume histograms and isodose distributions were compared. Other quantities such as Dmin(the minimum dose received by 99% of GTV/PTV), Dmax(the maximum dose received by 1% of GTV/PTV), D100, D95, D90, V110%, V105%, V100% (the volume of GTV/PTV receiving 110%, 105%, 100% of Dp), conformity index(CI), homogeneity index (HI), the volume of receiving 40Gy and 50 Gy to rectum (V40,V50) ; the volume of receiving 30Gy and 50 Gy to bladder (V30,V50) were evaluated. Total segments and MUs were also compared. Results: While all plans meet target dose specifications and normal tissue constraints, the maximum GTVCI of Pinnacle plans was up to 0.74 and the minimum of Corvus plans was only 0.21, these four TPSs PTVCI had significant difference. The GTVHI and PTVHI of Pinnacle plans are all very low and show a very good dose distribution. Corvus plans received the higer dose of normal tissue. The Monaco plans require significantly less segments and MUs to deliver than the other plans. Conclusion: To deliver on a Varian linear-accelerator, the Pinnacle plans show a very good dose distribution. Corvus plans received the higer dose of normal tissue. The Monaco plans have faster beam delivery.
Mandelli, Diego; Prescott, Steven; Smith, Curtis; Alfonsi, Andrea; Rabiti, Cristian; Cogliati, Joshua; Kinoshita, Robert
In this paper we evaluate the impact of a power uprate on a pressurized water reactor (PWR) for a tsunami-induced flooding test case. This analysis is performed using the RISMC toolkit: the RELAP-7 and RAVEN codes. RELAP-7 is the new generation of system analysis codes that is responsible for simulating the thermal-hydraulic dynamics of PWR and boiling water reactor systems. RAVEN has two capabilities: to act as a controller of the RELAP-7 simulation (e.g., component/system activation) and to perform statistical analyses. In our case, the simulation of the flooding is performed by using an advanced smooth particle hydrodynamics code called NEUTRINO. The obtained results allow the user to investigate and quantify the impact of timing and sequencing of events on system safety. The impact of power uprate is determined in terms of both core damage probability and safety margins.
Mandelli, Diego; Prescott, Steven; Smith, Curtis; Alfonsi, Andrea; Rabiti, Cristian; Cogliati, Joshua; Kinoshita, Robert
In this paper we evaluate the impact of a power uprate on a pressurized water reactor (PWR) for a tsunami-induced flooding test case. This analysis is performed using the RISMC toolkit: the RELAP-7 and RAVEN codes. RELAP-7 is the new generation of system analysis codes that is responsible for simulating the thermal-hydraulic dynamics of PWR and boiling water reactor systems. RAVEN has two capabilities: to act as a controller of the RELAP-7 simulation (e.g., component/system activation) and to perform statistical analyses. In our case, the simulation of the flooding is performed by using an advanced smooth particle hydrodynamics code calledmoreÂ Â» NEUTRINO. The obtained results allow the user to investigate and quantify the impact of timing and sequencing of events on system safety. The impact of power uprate is determined in terms of both core damage probability and safety margins.Â«Â less
Cao, Y; Li, R; Chi, Z
Purpose: To compare the performances of four commercial treatment planning systems (TPS) used for the intensity-modulated radiotherapy (IMRT). Methods: Ten patients of nasopharyngeal (4 cases), esophageal (3 cases) and cervical (3 cases) cancer were randomly selected from a 3-month IMRT plan pool at one radiotherapy center. For each patient, four IMRT plans were newly generated by using four commercial TPS (Corvus, Monaco, Pinnacle and Xio), and then verified with Matrixx (two-dimensional array/IBA Company) on Varian23EX accelerator. A pass rate (PR) calculated from the Gamma index by OminiPro IMRT 1.5 software was evaluated at four plan verification standards (1%/1mm, 2%/2mm, 3%/3mm, 4%/4mm and 5%/5mm) for each treatment plan. Overall and multiple pairwise comparisons of PRs were statistically conducted by analysis of covariance (ANOVA) F and LSD tests among four TPSs. Results: Overall significant (p>0.05) differences of PRs were found among four TPSs with F test values of 3.8 (p=0.02), 21.1(>0.01), 14.0 (>0.01), 8.3(>0.01) at standards of 1%/1mm to 4%/4mm respectively, except at 5%/5mm standard with 2.6 (p=0.06). All means (standard deviation) of PRs at 3%/3mm of 94.3 ± 3.3 (Corvus), 98.8 ± 0.8 (Monaco), 97.5± 1.7 (Pinnacle), 98.4 ± 1.0 (Xio) were above 90% and met clinical requirement. Multiple pairwise comparisons had not demonstrated a consistent low or high pattern on either TPS. Conclusion: Matrixx dose verification results show that the validation pass rates of Monaco and Xio plans are relatively higher than those of the other two; Pinnacle plan shows slight higher pass rate than Corvus plan; lowest pass rate was achieved by the Corvus plan among these four kinds of TPS.
4 18.104.22.168 Additional Wildlife Observations In September, a dead crow (Corvus brachyrhynchos) was observed dangling from the top of a spruce tree at the Site. Closer inspection revealed there was fishing line wrapped around the foot of the bird. Evidently the crow picked up the fishing line on its foot at one of the nearby reservoirs (since fishing is not allowed at the Site). The line was probably dangling from its foot when it landed on the branch. When it attempted to take off the line
10 image Outside view of the T=4 subunit arrangement. Â» Links Scientific Highlight Johnson Lab Â» Share this Article Laboratree Ologeez SciLink LabSpaces Following the pH-dependent Conformational Changes of a Maturing Viral Capsid summary written by Raven Hanna The capsid that surrounds viruses is formed from subunit proteins that interact in specific ways to form a tight shell. The processes of coming together and forming interactions are multistep and complex and are fundamental events to
February 25, 2010 seafloor_biofilms Image of pillow basalts from inside the Pisces Submersible. Â» Links Scientific Highlight Templeton Lab EMSL News Imaging at SSRL Â» Share this Article Laboratree Ologeez SciLink LabSpaces Researchers Discover an Unexpected Source of Energy for Deep-sea Microbial Communities summary written by Raven Hanna New rock formed by deep undersea volcanoes does not stay bare long. Microbes quickly move onto these basalts to form communities in the form of biofilms. As
Hasan Research Princeton News Release Â» Share this Article Laboratree Ologeez SciLink LabSpaces Macroscopic Quantum Insulator State Observed summary written by Raven Hanna One of the strangest consequences of quantum mechanics is the seemingly instantaneous communication of subatomic particles over long distances. Known as quantum entanglement, pairs or groups of particles can become linked so that any changes made to one will cause the others to respond quicker than the time it takes for light
of covered, raven-proof trash receptacles; disposal of edible trash in trash receptacles following the end of each work day, and disposal of trash in a designated sanitary landfill at the end of each work week. Material placed in a sanitary landfill will be covered daily when the landfill is open, as per NNSAMSO standard operating procedures. h. NNSAMSO submitted a habitat reclamation plan to the Service that describes the methods of stabilizing and revegetating sites (DOE 2004). The goal of
Provides a tool with which create and characterize a very large set of matrix-based visual analogy problems that have properties that are similar to RavenÂ™s Progressive Matrices (RPMs). The software uses the same underlying patterns found in RPMs to generate large numbers of unique matrix problems using parameters chosen by the researcher. Specifically, the software is designed so that researchers can choose the type, direction, and number of relations in a problem and then createmoreÂ Â» any number of unique matrices that share the same underlying structure (e.g. changes in numerosity in a diagonal pattern) but have different surface features (e.g. shapes, colors).RavenÂ™s Progressive Matrices (RPMs) are a widely-used test for assessing intelligence and reasoning ability. Since the test is non-verbal, it can be applied to many different populations and has been used all over the world. However, there are relatively few matrices in the sets developed by Raven, which limits their use in experiments requiring large numbers of stimuli. This tool creates a matrix set in a systematic way that allows researchers to have a great deal of control over the underlying structure, surface features, and difficulty of the matrix problems while providing a large set of novel matrices with which to conduct experiments.Â«Â less
RAVEN has 3 major functionalities: 1. Provides a Graphical User Interface for the pre- and post-processing of the RELAP-7 input and output. 2. Provides the capability to model nuclear power plants control logic for the RELAP-7 code and dynamic control of the accident scenario evolution. This capability is based on a software structure that realizes a direct connection between the RELAP-7 solver engine (MOOSE) and a python environment where the variables describing the plant statusmoreÂ Â» are accessible in a scripting environment. RAVEN support the generation of the probabilistic scenario control by supplying a wide range of probability and cumulative distribution functions and their inverse functions. 3. Provides a general environment to perform probability risk analysis for RELAP-7, RELAP-5 and any generic MOOSE based applications. The probabilistic analysis is performed by sampling the input space of the coupled code parameters and it is enhanced by using modern artificial intelligence algorithms that accelerate the identification of the areas of major risk (in the input parameter space). This environment also provides a graphical visualization capability to analyze the outcomes. Among other approaches, the classical Monte Carlo and Latin Hypercube sampling algorithms are available. For the acceleration of the convergence of the sampling methodologies, Support Vector Machines, Bayesian regression, and collocation stochastic polynomials chaos are implemented. The same methodologies here described could be used to solve optimization and uncertainties propagation problems using the RAVEN framework.Â«Â less
Hg Methylation Â» Links Scientific Highlight BL2-3 BL6-2C BL9-3 BL10-2 Â» Share this Article Laboratree Ologeez SciLink LabSpaces Imaging Mercury in the Rhizosphere of Wetland Plants summary written by Raven Hanna High levels of mercury in our diets can have adverse effects on our health, and fish are a major source of dietary mercury. Because of a process called biomagnification, mercury levels in fish can build up to be at a much higher concentration than in the surrounding water. The mercury
BE THE FACE OF CLIMATE CHANGE W elcom ing Rem arks Adam Cohen, Chief Operating Officer Native Am erican Dance To Honor the Earth Emelie Jeffries and Company 2012 PPPL Green M achine Aw ards Presentation Adam Cohen, COO Native Dance To Honor The Earth. PRINCETON PLASMA PHYSICS LABORATORY Be A Face of Climate Change Presents Earth Week 2013 at RavensWing Productions presents The Four Directions Native American Dancers To dance is to pray, to pray is to heal, to heal is to give, to give is to live,
Abrasion Testing of Critical Components of Hydrokinetic Devices 10/17/2012 University of Alaska Anchorage 2 Project Team o Ocean Renewable Power Company (ORPC) o Jarlath McEntee o Monty Worthington o University of Alaska Anchorage (UAA) o Faculty o Thomas Ravens o Todd Petersen o Muhammad Ali o Research Assistants o Tim Kirk o Jacob Clark o Angus Bromaghin 10/17/2012 University of Alaska Anchorage 3 ORPC Technology o TideGen Power System (TGU) o Designed to generate electricity at water depths
Maryland" ,"Plant","Primary energy source","Operating company","Net summer capacity (MW)" 1,"Chalk Point LLC","Petroleum","NRG Chalk Point LLC",2248 2,"Calvert Cliffs Nuclear Power Plant","Nuclear","Calvert Cliffs Nuclear PP LLC",1716 3,"Morgantown Generating Plant","Coal","GenOn Mid-Atlantic LLC",1423 4,"Brandon Shores","Coal","Raven
Mandelli, Diego; Prescott, Steven R; Smith, Curtis L; Alfonsi, Andrea; Rabiti, Cristian; Cogliati, Joshua J; Kinoshita, Robert A
In the Risk Informed Safety Margin Characterization (RISMC) approach we want to understand not just the frequency of an event like core damage, but how close we are (or are not) to key safety-related events and how might we increase our safety margins. The RISMC Pathway uses the probabilistic margin approach to quantify impacts to reliability and safety by coupling both probabilistic (via stochastic simulation) and mechanistic (via physics models) approaches. This coupling takes place through the interchange of physical parameters and operational or accident scenarios. In this paper we apply the RISMC approach to evaluate the impact of a power uprate on a pressurized water reactor (PWR) for a tsunami-induced flooding test case. This analysis is performed using the RISMC toolkit: RELAP-7 and RAVEN codes. RELAP-7 is the new generation of system analysis codes that is responsible for simulating the thermal-hydraulic dynamics of PWR and boiling water reactor systems. RAVEN has two capabilities: to act as a controller of the RELAP-7 simulation (e.g., system activation) and to perform statistical analyses (e.g., run multiple RELAP-7 simulations where sequencing/timing of events have been changed according to a set of stochastic distributions). By using the RISMC toolkit, we can evaluate how power uprate affects the system recovery measures needed to avoid core damage after the PWR lost all available AC power by a tsunami induced flooding. The simulation of the actual flooding is performed by using a smooth particle hydrodynamics code: NEUTRINO.
Zhegang Ma; Diego Mandelli; Curtis Smith
A previous study used RELAP and RAVEN to conduct a boiling water reactor station black-out (SBO) case study in a simulation based environment to show the capabilities of the risk-informed safety margin characterization methodology. This report compares the RELAP/RAVEN simulation results with traditional PRA model results. The RELAP/RAVEN simulation run results were reviewed for their input parameters and output results. The input parameters for each simulation run include various timing information such as diesel generator or offsite power recovery time, Safety Relief Valve stuck open time, High Pressure Core Injection or Reactor Core Isolation Cooling fail to run time, extended core cooling operation time, depressurization delay time, and firewater injection time. The output results include the maximum fuel clad temperature, the outcome, and the simulation end time. A traditional SBO PRA model in this report contains four event trees that are linked together with the transferring feature in SAPHIRE software. Unlike the usual Level 1 PRA quantification process in which only core damage sequences are quantified, this report quantifies all SBO sequences, whether they are core damage sequences or success (i.e., non core damage) sequences, in order to provide a full comparison with the simulation results. Three different approaches were used to solve event tree top events and quantify the SBO sequences: “W” process flag, default process flag without proper adjustment, and default process flag with adjustment to account for the success branch probabilities. Without post-processing, the first two approaches yield incorrect results with a total conditional probability greater than 1.0. The last approach accounts for the success branch probabilities and provides correct conditional sequence probabilities that are to be used for comparison. To better compare the results from the PRA model and the simulation runs, a simplified SBO event tree was developed with only four top events and eighteen SBO sequences (versus fifty-four SBO sequences in the original SBO model). The estimated SBO sequence conditional probabilities from the original SBO model were integrated to the corresponding sequences in the simplified SBO event tree. These results were then compared with the simulation run results.
Skripnikov, L. V.; Petrov, A. N.; Titov, A. V.; Mawhorter, R. J.; Baum, A. L.; Sears, T. J.; Grabow, J. -U.
We report the results of our theoretical study and analysis of earlier experimental data for the g-factor tensor components of the ground 2II1/2 state of the free PbF radical. These values obtained both within the relativistic coupled-cluster method combined with the generalized relativistic effective core potential approach and with our fit of the experimental data from [R. J. Mawhorter, B. S. Murphy, A. L. Baum, T. J. Sears, T. Yang, P. M. Rupasinghe, C. P. McRaven, N. E. Shafer-Ray, L. D. Alphei, and J.-U. Grabow, Phys. Rev. A 84, 022508 (2011); A. L. Baum, B.A. thesis, Pomona College, 2011]. ThemoreÂ Â» obtained results agree very well with each other but contradict the previous fit performed in the cited works. Our final prediction for g factors is Gâˆ¥=0.081(5),GâŠ¥=â€“0.27(1).Â«Â less
Smith, Curtis L; Ma, Zhegang; Tom Riley; Mandelli, Diego; Nielsen, Joseph W; Alfonsi, Andrea; Rabiti, Cristian
This report summarizes the research activity developed during the Fiscal year 2014 within the Risk Informed Safety Margin and Characterization (RISMC) pathway within the Light Water Reactor Sustainability (LWRS) campaign. This research activity is complementary to the one presented in the INL/EXT-??? report which shows advances Probabilistic Risk Assessment Analysis using RAVEN and RELAP-7 in conjunction to novel flooding simulation tools. Here we present several analyses that prove the values of the RISMC approach in order to assess risk associated to nuclear power plants (NPPs). We focus on simulation based PRA which, in contrast to classical PRA, heavily employs system simulator codes. Firstly we compare, these two types of analyses, classical and RISMC, for a Boiling water reactor (BWR) station black out (SBO) initiating event. Secondly we present an extended BWR SBO analysis using RAVEN and RELAP-5 which address the comments and suggestions received about he original analysis presented in INL/EXT-???. This time we focus more on the stochastic analysis such probability of core damage and on the determination of the most risk-relevant factors. We also show some preliminary results regarding the comparison between RELAP5-3D and the new code RELAP-7 for a simplified Pressurized Water Reactors system. Lastly we present some conceptual ideas regarding the possibility to extended the RISMC capabilities from an off-line tool (i.e., as PRA analysis tool) to an online-tool. In this new configuration, RISMC capabilities can be used to assist and inform reactor operator during real accident scenarios.
Skripnikov, L. V.; Petrov, A. N.; Titov, A. V.; Mawhorter, R. J.; Baum, A. L.; Sears, T. J.; Grabow, J. -U.
We report the results of our theoretical study and analysis of earlier experimental data for the g-factor tensor components of the ground 2II1/2 state of the free PbF radical. These values obtained both within the relativistic coupled-cluster method combined with the generalized relativistic effective core potential approach and with our fit of the experimental data from [R. J. Mawhorter, B. S. Murphy, A. L. Baum, T. J. Sears, T. Yang, P. M. Rupasinghe, C. P. McRaven, N. E. Shafer-Ray, L. D. Alphei, and J.-U. Grabow, Phys. Rev. A 84, 022508 (2011); A. L. Baum, B.A. thesis, Pomona College, 2011]. The obtained results agree very well with each other but contradict the previous fit performed in the cited works. Our final prediction for g factors is G?=0.081(5),G?=–0.27(1).
Mandelli, Diego; Smith, Curtis; Prescott, Steven; Alfonsi, Andrea; Rabiti, Cristian; Cogliati, Joshua; Kinoshita, Robert
The existing fleet of nuclear power plants is in the process of extending its lifetime and increasing the power generated from these plants via power uprates. In order to evaluate the impacts of these two factors on the safety of the plant, the Risk Informed Safety Margin Characterization project aims to provide insights to decision makers through a series of simulations of the plant dynamics for different initial conditions (e.g., probabilistic analysis and uncertainty quantification). This paper focuses on the impacts of power uprate on the safety margin of a boiling water reactor for a flooding induced station black-out event. Analysis is performed by using a combination of thermal-hydraulic codes and a stochastic analysis tool currently under development at the Idaho National Laboratory, i.e. RAVEN. We employed both classical statistical tools, i.e. Monte-Carlo, and more advanced machine learning based algorithms to perform uncertainty quantification in order to quantify changes in system performance and limitations as a consequence of power uprate. Results obtained give a detailed investigation of the issues associated with a plant power uprate including the effects of station black-out accident scenarios. We were able to quantify how the timing of specific events was impacted by a higher nominal reactor core power. Such safety insights can provide useful information to the decision makers to perform risk informed margins management.
D. Mandelli; C. Smith; T. Riley; J. Nielsen; J. Schroeder; C. Rabiti; A. Alfonsi; Cogliati; R. Kinoshita; V. Pasucci; B. Wang; D. Maljovec
Dynamic Probabilistic Risk Assessment (DPRA) methodologies couple system simulator codes (e.g., RELAP, MELCOR) with simulation controller codes (e.g., RAVEN, ADAPT). While system simulator codes accurately model system dynamics deterministically, simulation controller codes introduce both deterministic (e.g., system control logic, operating procedures) and stochastic (e.g., component failures, parameter uncertainties) elements into the simulation. Typically, a DPRA is performed by: 1) sampling values of a set of parameters from the uncertainty space of interest (using the simulation controller codes), and 2) simulating the system behavior for that specific set of parameter values (using the system simulator codes). For complex systems, one of the major challenges in using DPRA methodologies is to analyze the large amount of information (i.e., large number of scenarios ) generated, where clustering techniques are typically employed to allow users to better organize and interpret the data. In this paper, we focus on the analysis of a nuclear simulation dataset that is part of the Risk Informed Safety Margin Characterization (RISMC) Boiling Water Reactor (BWR) station blackout (SBO) case study. We apply a software tool that provides the domain experts with an interactive analysis and visualization environment for understanding the structures of such high-dimensional nuclear simulation datasets. Our tool encodes traditional and topology-based clustering techniques, where the latter partitions the data points into clusters based on their uniform gradient flow behavior. We demonstrate through our case study that both types of clustering techniques complement each other in bringing enhanced structural understanding of the data.
Matzen, Laura E.; Trumbo, Michael Christopher Stefan
Training a person in a new knowledge base or skill set is extremely time consuming and costly, particularly in highly specialized domains such as the military and the intelligence community. Recent research in cognitive neuroscience has suggested that a technique called transcranial direct current stimulation (tDCS) has the potential to revolutionize training by enabling learners to acquire new skills faster, more efficiently, and more robustly (Bullard et al., 2011). In this project, we tested the effects of tDCS on two types of memory performance that are critical for learning new skills: associative memory and working memory. Associative memory is memory for the relationship between two items or events. It forms the foundation of all episodic memories, so enhancing associative memory could provide substantial benefits to the speed and robustness of learning new information. We tested the effects of tDCS on associative memory, using a real-world associative memory task: remembering the links between faces and names. Working memory refers to the amount of information that can be held in mind and processed at one time, and it forms the basis for all higher-level cognitive processing. We investigated the degree of transfer between various working memory tasks (the N-back task as a measure of verbal working memory, the rotation-span task as a measure of visuospatial working memory, and Raven's progressive matrices as a measure of fluid intelligence) in order to determine if tDCS-induced facilitation of performance is task-specific or general.
Windsor, Bradford T.; Woodring, Mitchell L.; Myjak, Mitchell J.
1.0 Overview The TSA systems VM-250AGN portal monitor is a set of two pillars made to detect nuclear material in a vehicle. Each pillar contains two polyvinyl toluene (PVT) plastic gamma ray detectors and four 3He neutron detectors, as well as a power supply and electronics to process the output from these detectors. Pacific Northwest National Laboratory has designed and built a continuous high-resolution PVT readout (CHRPR) for the TSA portal to allow spectral readout from the gamma and neutron detectors. The CHRPR helps differentiate between different types of radioactive material through increased spectroscopic capability and associated developments. The TSA VM-250AGN continually monitors the natural neutron and gamma ray background which occurs around the pillars. When the system is installed, the two pillars are placed on either side of a roadway, and a vehicle presence sensor records the passage of cars between them. When radiation measurements exceed a preset alarm threshold, the system alarms to let the user know that a radioactive material is present. Time-stamped measurements are continually sent to a computer, where they can be recorded via a Windows terminal or the TSA RAVEN software. For each pillar in the original TSA model, output from each detector is amplified and shaped by a single channel analyzer, the SCA-775. Information from both SCA-775’s are passed to the SC-770 in the master pillar. This is the detector interface module and main data processor. It counts electrical pulses and uses program software to output total readings to the computer, as well as trigger any appropriate alarms. The CHRPR allows a parallel approach to recording radiation readings from the TSA system. After installing the CHRPR system, all TSA power and signal connections are unchanged. The CHRPR captures electrical pulses containing detector and occupancy sensor information from the SCA-775 on either side. These pulses are converted to a signal with a time width proportional to the amplitude, via voltage to pulse width converters (VPW). These time widths are then digitized by a field programmable gate array (FPGA) and transmitted over Ethernet to a data acquisition computer. The CHRPR records the magnitude of each pulse to a continuous event mode file on or each detector and occupancy sensor This manual begins with CHRPR installation instructions, then a section on CHRPR software. Afterward is a brief overview of how the TSA system works, then an explanation of the CHRPR. This manual is meant as a supplement to the TSA VM-250AGN manual, which can be found at http://tsasystems.com/library/manuals/pm700agn-vm250agn_manual.pdf . That manual is the manufacturer’s guide for the installation, programming, and maintenance of the portal system.
Franzese, Oscar; Lascurain, Mary Beth; Capps, Gary J
Since the early part of the 20th century, the US trucking industry has provided a safe and economical means of moving commodities across the country. At the present time, nearly 80% of the US domestic freight movement involves the use of trucks. The US Department of Energy (DOE) is spearheading a number of research efforts to improve heavy vehicle fuel efficiencies. This includes research in engine technologies (including hybrid and fuel cell technologies), lightweight materials, advanced fuels, and parasitic loss reductions. In addition, DOE is developing advanced tools and models to support heavy vehicle truck research, and is leading the 21st Century Truck Partnership whose stretch goals involve a reduction by 50% of the fuel consumption of heavy vehicles on a ton-mile basis. This Medium Truck Duty Cycle (MTDC) Project is a critical element in DOE s vision for improved heavy vehicle energy efficiency and is unique in that there is no other national database of characteristic duty cycles for medium trucks. It involves the collection of real-world data for various situational characteristics (rural/urban, freeway/arterial, congested/free-flowing, good/bad weather, etc.) and looks at the unique nature of medium trucks drive cycles (stop-and-go delivery, power takeoff, idle time, short-radius trips), to provide a rich source of data that can contribute to the development of new tools for fuel efficiency and modeling, provide DOE a sound basis upon which to make technology investment decisions, and provide a national archive of real-world-based medium-truck operational data to support heavy vehicle energy efficiency research. The MTDC project involves a two-part field operational test (FOT). For the Part-1 FOT, three vehicles, each from two vocations (urban transit and dry-box delivery) were instrumented for one year of data collection. The Part-2 FOT will involve the towing/recovery and utility vocations. The vehicles participating in the MTDC project are doing so through gratis partnerships in return for early access to the results of this study. Partnerships such as these are critical to FOTs in which real-world data is being collected. In Part 1 of the project, Oak Ridge National Laboratory(ORNL) established partnerships with the H.T. Hackney Company, one of the largest wholesale distributors in the country, distributing products to 21 states; and with the Knoxville Area Transit (KAT), the City of Knoxville s transit system, operating services across the city of Knoxville and parts of Knox co. These partnerships and agreements provided ORNL access to three Class-7 2005/2007 International day-cab tractors, model 8600, which regularly haul 28 ft pup trailers (H.T. Hackney Co) and three Class-7 2005 Optima LF-34 buses (KAT), for collection of duty cycle data. In addition, ORNL has collaborated with the Federal Motor Carrier Safety Administration (FMCSA) to determine if there were possible synergies between this duty cycle data collection effort and FMCSA s need to learn more about the operation and duty cycles of the second-largest fuel consuming commercial vehicle category in the US. FMCSA s primary interest was in collecting safety data relative to the driver, carrier, and vehicle. In order to collect the duty cycle and safety-related data, ORNL developed a data acquisition and wireless communication system that was placed on each test vehicle. Each signal recorded in this FOT was collected by means of one of the instruments incorporated into each data acquisition system (DAS). Native signals were obtained directly from the vehicle s J1939 and J1708 data buses. A VBOX II Lite collected Global Positioning System related information including speed, acceleration, and spatial location information at a rate of 5 Hz, and communicated this data via the CAN (J1939) protocol. The Air-Weigh LoadMaxx, a self-weighing system which determines the vehicle s gross weight by means of pressure transducers and posts the weight to the vehicle s J1939 data bus, was used to collect vehicle payload information. A cellular modem, the Raven X