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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 811 records · Page 45
All sourcesNASA NTRSDOE OSTI
DOE OSTI2011

Final Report: Towards Optimal Petascale Simulations (TOPS), ER25785

Multiscale, multirate scientific and engineering applications in the SciDAC portfolio possess resolution requirements that are practically inexhaustible and demand execution on the highest-capability computers available, which will soon reach the petascale. While the variety of applications is enormous, their needs for mathematical software infrastructure are surprisingly coincident; moreover the chief bottleneck is often the solver. At their current scalability limits, many applications spend a vast majority of their operations in solvers, due to solver algorithmic complexity that is superlinear in the problem size, whereas other phases scale linearly. Furthermore, the solver may be the phase of the simulation with the poorest parallel scalability, due to intrinsic global dependencies. This project brings together the providers of some of the world's most widely distributed, freely available, scalable solver software and focuses them on relieving this bottleneck for many specific applications within SciDAC, which are representative of many others outside. Solver software directly supported under TOPS includes: hypre, PETSc, SUNDIALS, SuperLU, TAO, and Trilinos. Transparent access is also provided to other solver software through the TOPS interface. The primary goals of TOPS are the development, testing, and dissemination of solver software, especially for systems governed by PDEs. Upon discretization, these systems possess mathematical structure that must be exploited for optimal scalability; therefore, application-targeted algorithmic research is included. TOPS software development includes attention to high performance as well as interoperability among the solver components. Support for integration of TOPS solvers into SciDAC applications is also directly supported by this proposal. The role of the UCSD PI in this overall CET, is one of direct interaction between the TOPS software partners and various DOE applications scientists' specifically toward magnetohydrodynamics (MHD) simulations with the Center for Extended Magnetohydrodynamic Modeling (CEMM) SciDAC and Applied Partial Differential Equations Center (APDEC) SciDAC, and toward core-collapse supernova simulations with the previous Terascale Supernova Initiative (TSI) SciDAC and in continued work on INCITE projects headed by Doug Swesty, SUNY Stony Brook. In addition to these DOE applications scientists, the UCSD PI works to bring leading-edge DOE solver technology to applications scientists in cosmology and large-scale galactic structure formation. Unfortunately, the funding for this grant ended after only two years of its five-year duration, in August 2008, due to difficulties at DOE in transferring the grant to the PI's new faculty position at Southern Methodist University. Therefore, this report only describes two years' worth of effort.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2018

Online Weld Quality Inspection System with Infrared Thermography

Today’s passenger vehicle on average contains 4,000 to 6,000 resistance spot welds (RSW). All US automakers perform destructive teardown evaluation to ensure the quality of the spot welds due to their criticality in crash resistance and performance of vehicles. The destructive teardown evaluation is labor intensive and costly. Sponsored by DOE VT Office, ORNL recently developed a prototype system (hardware and software) to non-destructively evaluate the quality of RSW based on infrared (IR) thermography. Highly cost competitive, ORNL’s system has been tested extensively in the laboratory with initial successful production environment validation. Ford, Honda, GM, Chrysler and major Tier One suppliers all have shown keen interest in this new technology. ORNL’s Technology Transfer Office licensed this technology to a small business (APLAIR Manufacturing Solutions) in TN. The licensee planned to develop a new hardware system based on ORNL’s IR NDE technology for real-time online weld inspection. In this CRADA project, ORNL was tasked to assist the licensee to transfer the technology and develop the new product suitable for automotive body-in-white production environment, making the technology ready for ‘real-world’ auto-body assembly line production environment.

47 OTHER INSTRUMENTATION↗
DOE OSTI2017

Resource management based on physical authentication and authorization

Systems and methods are disclosed that provide for management of resources of one or more systems included in an access-controlled area of a distributed site of an electric power delivery system. In certain embodiments, one or more users entering and access-controlled area may be identified via physical access control credentials provided to an associated access control system. A determination may be made as to whether the users have access rights to one or more hardware and/or software resources of systems included in the access-controlled area. Based on the determination, control signals may be generated by the access control system to enable and/or disable associated resources.

Gammel, Dennis↗
DOE OSTI2015

Next-generation system analysis model for studying alternative Waste Management Systems - 15232

The U.S. Department of Energy's Office of Nuclear Energy (DOE-NE) is conducting planning activities within the Nuclear Fuels Storage and Transportation Planning Project (NFST). The plans involve (1) laying the groundwork for implementing interim storage (including dealing with any associated transportation that is needed) per the Administration's Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste of January 2013 and (2) developing a foundation for a new nuclear waste management organization. An overarching goal is to develop options for designing an integrated system that will be considered by decision-makers. Based on the dynamic system analysis and software specification requirements that must be met, Argonne National Laboratory (Argonne) is leading the development of a consolidated waste management system tool known as the Next-Generation System Analysis Model (NGSAM). The development of NGSAM leverages the approaches and methods used previously to develop the Transportation Storage Logistics (TSL) and Total System Model (TSM) system analysis tools, but the new tool takes advantage of advanced computational capabilities. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

SU-E-T-341: DVH-Based Comparison of True 3D Measurements to a Delta4 System

Purpose: Delta4 dosimetric software can be used to calculate DVH-based metrics for patient-specific quality assurance from measurements made by a Delta4 QA device. This study investigates the effectiveness of a novel transform method that transposes measurements made with a full-density 3D dosimeter onto patient anatomy, enabling the calculation of DVHs. This allows for DVH comparisons from the transformed dose distribution, which are based on true 3D measurements, to those from the Delta4 system, which are based on semi-3D measurements and interpolation. Methods: A double-arc VMAT treatment for a head-and-neck case was delivered to a 1kg PRESAGE 3D dosimeter inserted into a polyurethane head phantom. The dosimeter was readout using an in-house optical-CT scanner to gather full-density 3D dosimetric data. The transform method is achieved by multiplication of the measured doses with a “transformation matrix” which accounts for heterogeneities and differences in geometry between the patient and the phantom. The transformation matrix is a voxel-by-voxel division of the patient planned dose by the phantom planned dose, both calculated in the treatment planning system (Eclipse). The transformed distribution was then overlaid on the patient CT data, enabling the calculation of DVHs. The same VMAT treatment was delivered to the Delta4 phantom and DVH data was calculated using its associated software. Results: The transformed dose distribution showed good agreement with calculated patient values, determined by similarity in dose profiles between the two distributions and a 3D gamma index passing rate of 94.87% for 3%/3mm criteria. For every structure contained within the dosimeter volume, the transformed DVHs demonstrated better agreement than the Delta4 DVHs, when compared to the values calculated in the treatment planning system. Conclusion: The coupled technique of full-density 3D dose measurements and the presented transform method enables clinical patient-specific quality assurance data that is more accurate than the semi-3D Delta4 system. This work was supported by NIH R01CA100835.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2016

Akatosh: Automated Cyber Incident Verification and Impact Analysis

Akatosh is a highly configurable system based on the integration of the capabilities of one or more Intrusion Detection Systems (IDS) and automated forensic analysis. Akatosh reduces the false positive rates of IDSs and alleviates costs of incident response by pointing forensic personnel to the root cause of an incident on affected endpoint devices. Akatosh is able to analyze a computer system in near real-time and provide operations and forensic analyst personnel with continuous feedback on the impact of malware and software on deployed systems. Additionally, Akatosh provides the ability to look back into any prior state in the history of the computer system along with the ability to compare one or more prior system states with any other prior state.

Smith, Jared M↗
DOE OSTI2021

Cryogenic control system operational experience at SNS

Abstract The helium cryogenic system at Spallation Neutron Source (SNS) provides cooling to 81 superconducting radio frequency cavities. To support the operation of the cryogenic facility, a highly reliable control system consisting of software, hardware, and Human Machine Interface (HMI) has been developed and improved during the first fifteen years of operation. Integrating the cryogenic control system with other subsystems of the SNS complex is an important aspect to the success of the operation. The operating experience, lessons learned and recommendations to consider for future facilities will be detailed in this paper.

Howell, M. (ORCID:0000000310360707)↗
DOE OSTI2021

Displacement driven balancing of a diamond turning machine flycutter

The spindle balancing operation is critical to the performance of precision lathes and should be conducted anytime the configuration of the spindle or its tooling is modified such that an unbalanced condition may occur. Within the machine tool community, accelerometers are typically used as the feedback device during the balancing operation. At Lawrence Livermore National Laboratory (LLNL), we have nearly 40 years of history using displacement sensors, such as capacitance gauges, in place of accelerometers when balancing precision spindles. The process of balancing a precision machine spindle using displacement sensors is presented in the context of two work examples. In this case, the details of the one-dimensional balancing operation performed after replacing a flycutter on the spindles of two unique large optic diamond turning machines is discussed. The pros and cons of accelerometer driven balancing and displacement driven balancing are also presented. In the accelerometer context, the level of balance achieved for our work example is less than 100 micro-g or a total indicator reading (TIR) of 1.6 µm. The process of balancing has been enhanced over the years as advances have been made in software, data acquisition systems, and instrumentation. The balancing system is described including details about instrumentation, set-up, how phase and spindle structure stiffness is calculated, and how the LLNL developed software that guides the user through the process by suggesting corrective action.

42 ENGINEERING↗
DOE OSTI2017

Application of a bayesian belief network model to reliability assessment of nuclear safety-related software - 207

As the instrumentation and control (I and C) systems in nuclear power plants (NPPs) have been replaced with digital-based systems, the need to incorporate software failures into NPP probabilistic risk assessments has arisen. In order to assess the probability of software failure on demand, a Bayesian belief network (BBN) model was developed which estimates the number of defects and the resulting probability of software failure on demand in nuclear safety-related software. To assess the feasibility of the BBN framework, the BBN model was applied to the prototype Integrated Digital Protection System-Reactor Protection System (IDiPS-RPS) to estimate the number of remaining faults and the software failure probability of a target software. The developmental- and V and V-activities carried out during the IDiPS-RPS development process were evaluated based on the well-defined checklist derived by the V and V team and were estimated based on expert elicitation. In addition, the attribute evaluations and the number of FPs of the target software is provided as the inputs for the BBN model. The application results showed the feasibility of using BBNs for quantifying software failure probabilities and several insights were gained from the applications of the BBN model. The proposed BBN framework can be applied to estimate the software failure probability for other safety-related NPP software and provide an insight on modeling the software development process that involves iterations between different development phases. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

The effectiveness of pretreatment physics plan review for detecting errors in radiation therapy

Purpose: The pretreatment physics plan review is a standard tool for ensuring treatment quality. Studies have shown that the majority of errors in radiation oncology originate in treatment planning, which underscores the importance of the pretreatment physics plan review. This quality assurance measure is fundamentally important and central to the safety of patients and the quality of care that they receive. However, little is known about its effectiveness. The purpose of this study was to analyze reported incidents to quantify the effectiveness of the pretreatment physics plan review with the goal of improving it. Methods: This study analyzed 522 potentially severe or critical near-miss events within an institutional incident learning system collected over a three-year period. Of these 522 events, 356 originated at a workflow point that was prior to the pretreatment physics plan review. The remaining 166 events originated after the pretreatment physics plan review and were not considered in the study. The applicable 356 events were classified into one of the three categories: (1) events detected by the pretreatment physics plan review, (2) events not detected but “potentially detectable” by the physics review, and (3) events “not detectable” by the physics review. Potentially detectable events were further classified by which specific checks performed during the pretreatment physics plan review detected or could have detected the event. For these events, the associated specific check was also evaluated as to the possibility of automating that check given current data structures. For comparison, a similar analysis was carried out on 81 events from the international SAFRON radiation oncology incident learning system. Results: Of the 356 applicable events from the institutional database, 180/356 (51%) were detected or could have been detected by the pretreatment physics plan review. Of these events, 125 actually passed through the physics review; however, only 38% (47/125) were actually detected at the review. Of the 81 events from the SAFRON database, 66/81 (81%) were potentially detectable by the pretreatment physics plan review. From the institutional database, three specific physics checks were particularly effective at detecting events (combined effectiveness of 38%): verifying the isocenter (39/180), verifying DRRs (17/180), and verifying that the plan matched the prescription (12/180). The most effective checks from the SAFRON database were verifying that the plan matched the prescription (13/66) and verifying the field parameters in the record and verify system against those in the plan (23/66). Software-based plan checking systems, if available, would have potential effectiveness of 29% and 64% at detecting events from the institutional and SAFRON databases, respectively. Conclusions: Pretreatment physics plan review is a key safety measure and can detect a high percentage of errors. However, the majority of errors that potentially could have been detected were not detected in this study, indicating the need to improve the pretreatment physics plan review performance. Suggestions for improvement include the automation of specific physics checks performed during the pretreatment physics plan review and the standardization of the review process.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2024

CommBench: Micro-Benchmarking Hierarchical Networks with Multi-GPU, Multi-NIC Nodes

Modern high-performance computing systems have multiple GPUs and network interface cards (NICs) per node. The resulting network architectures have multilevel hierarchies of subnetworks with different interconnect and software technologies. These systems offer multiple vendor-provided communication capabilities and library implementations (IPC, MPI, NCCL, RCCL, OneCCL) with APIs providing varying levels of performance across the different levels. Understanding this performance is currently difficult because of the wide range of architectures and programming models (CUDA, HIP, OneAPI). We present CommBench, a library with cross-system portability and a high-level API that enables developers to easily build microbenchmarks relevant to their use cases and gain insight into the performance (bandwidth & latency) of multiple implementation libraries on different networks. We demonstrate CommBench with three sets of microbenchmarks that profile the performance of six systems. Our experimental results reveal the effect of multiple NICs on optimizing the bandwidth across nodes and also present the performance characteristics of four available communication libraries within and across nodes of NVIDIA, AMD, and Intel GPU networks.

Hidayetoglu, Mert↗
DOE OSTI2025

The SAS4A/SASSYS-1 Version 5.8 Safety Analysis Code System

SAS4A/SASSYS-1 is a software simulation tool used to perform deterministic analysis of anticipated events as well as design basis and beyond design basis accidents for advanced nuclear reactors. Detailed, mechanistic models of steady-state and transient thermal, hydraulic, kinetic, and mechanical phenomena are employed to describe the response of the reactor core, the reactor primary and secondary coolant loops, the reactor control and protection systems, and the balance-of-plant to accidents caused by changes in coolant flow, loss of heat rejection, or reactivity insertion. The consequences of single and double-fault accidents can be modeled, including fuel and coolant heating, fuel and cladding mechanical behavior, core reactivity feedbacks, coolant loop performance including natural circulation, and decay heat removal. Analyses are typically terminated upon demonstration of reactor and plant shutdown to permanently coolable conditions, or upon violation of design basis margins. The objective of the analysis is to quantify accident consequences as measured by the transient behavior of system performance parameters, such as fuel and cladding temperatures, reactivity, and cladding strain. Originally developed for analysis of sodium cooled reactors with oxide fuel clad by stainless steel, the models were subsequently extended and specialized to metallic fuel clad with advanced alloys and to several other coolant options, including lead, LBE, and water.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

The SAS4A/SASSYS-1 Safety Analysis Code System

SAS4A/SASSYS-1 is a software simulation tool used to perform deterministic safety analysis of anticipated events as well as design basis and beyond design basis accidents for advanced nuclear reactors. This software can be used to assess the safety of a prescribed reactor design, but it cannot be used to configure a design to meet targeted performance objectives. Detailed, mechanistic models of steady-state and transient thermal,hydraulic, kinetic, and mechanical phenomena are employed to describe the response of the reactor core, the reactor primary and secondary coolant loops, the reactor control and protection systems and the balance-of-plant to accidents caused by changes in coolantflow, oss of heat rejection, or reactivity insertion. The consequences of single and double-fault accidents can be modeled, including fuel and coolant heating, fuel and cladding mechanical behavior, core reactivity feedbacks, coolant loop performance including natural circulation, and decay heat removal. Analyses are typically terminated upon demonstration of reactor and plant shutdown to permanently coolable conditions or upon violation of design basis margins. The objective of the analysis is to quantify accident consequences as measured by the transient behavior of system performance parameters, such as fuel and cladding temperatures, reactivity, and cladding strain. Originally developed for analysis of sodium cooled reactors with oxide fuel clad by stainless steel, the models In SAS4A/SASSYS-1 were subsequently extended and specialized to metallic fuel clad with advanced alloys and to several other coolant options including lead and LBE. METHOD OF SOLUTION: Each subassembly in the reactor core is represented with single or multiple-pin models. One channel represents one or more fuel pins and many channels are employed for a whole-core representation. Heat transfer in each pin is modeled with a two-dimensional (r/z) heat conduction equation. Single and two-phase coolant thermal-hydraulics are simulated with a unique, one-dimensional (axial) multiple-bubble liquid metal boiling model. The transient fuel and cladding mechanical oehavior models are integrated with fission product production,release, and transport models. UNIQUE FEATURES: The physical models in SAS4AJSASSYS-1 are highly detailed 1umerical representations of reactor accident conditions based on extensive laboratory and test reactor results. The models are specialized to liquid metal (sodium) cooled fast reactors with oxide or metallic fuel clad with stainless steel.

Source record↗
DOE OSTI2017

SAS4A/SASSYS-1 VERSION 5.2 SAFETY ANALYSIS CODE SYSTEM

SAS4A/SASSYS-1 is a software simulation tool used to perform deterministic safety analysis of anticipated events as well as design basis and beyond design basis accidents for advanced nuclear reactors. This software can be used to assess the safety of a prescribed reactor design, but it cannot be used to configure a design to meet targeted performance objectives. Detailed, mechanistic models of steady-state and transient thermal, hydraulic, kinetic, and mechanical phenomena are employed to describe the response of the reactor core, the reactor primary and secondary coolant loops, the reactor control and protection systems and the balance-of-plant to accidents caused by changes in coolant flow, oss of heat rejection, or reactivity insertion. The consequences of single and double-fault accidents can be modeled, including fuel and coolant heating, fuel and cladding mechanical behavior, core reactivity feedbacks, coolant loop performance including natural circulation, and decay heat removal. Analyses are typically terminated upon demonstration of reactor and plant shutdown to permanently coolable conditions or upon violation of design basis margins. The objective of the analysis is to quantify accident consequences as measured by the transient behavior of system performance parameters, such as fuel and cladding temperatures, reactivity, and cladding strain. Originally developed for analysis of sodium cooled reactors with oxide fuel clad by stainless steel, the models In SAS4A/SASSYS-1 were subsequently extended and specialized to metallic fuel clad with advanced alloys and to several other coolant options including lead and LBE. METHOD OF SOLUTION: Each subassembly in the reactor core is represented with single or multiple-pin models. One channel represents one or more fuel pins and many channels are employed for a whole-core representation. Heat transfer in each pin is modeled with a two-dimensional (r/z) heat conduction equation. Single and two-phase coolant thermal-hydraulics are simulated with a unique, one-dimensional (axial) multiple-bubble liquid metal boiling model. The transient fuel and cladding mechanical behavior models are integrated with fission product production,release, and transport models. UNIQUE FEATURES: The physical models in SAS4AJSASSYS-1 are highly detailed 1umerical representations of reactor accident conditions based on extensive laboratory and test reactor results. The models are specialized to liquid metal (sodium) cooled fast reactors with oxide or metallic fuel clad with stainless steel.«

Fanning, Thomas↗
DOE OSTI2024

Modernizing the Legacy Fission Wire Measurement System for the Advanced Test Reactor-Critical Facility

Operational lifetime extensions of existing research reactors have emphasized the need for refurbishment, replacements, and upgrades to supporting equipment and instrumentation. The Advanced Test Reactor (ATR) at Idaho National Laboratory (INL), which entered service in 1967, has recently completed the sixth core internals change-out and has scheduled operations until at least 2040. Reactor maintenance and operational risk management is critically important in the research reactor community, however supporting measurement systems sometimes get overlooked when maintenance is planned. The Fission Wire Measurement System (FWMS) is a custom measurement system designed in the 1960s to measure the beta-particle activity of irradiated uranium-aluminum fission wires. This measurement is conducted to determine the fission rate profile of the Advanced Reactor Test Critical (ATR-C) facility. The ATR-C is an open-pool, low-power test reactor that was purpose driven to resemble ATR and is used to qualify experiment configurations and verify core models prior to full-power experiment irradiations in ATR. A power distribution measurement in ATR-C uses uranium-aluminum wires that are distributed throughout the ATR-C core to validate simulation and modeling results. These measurements require 340 to 1500 wires to be irradiated and measured within a 12-hour window. The activity of the wires is measured in the required time with the FWMS, which was put into service in 1965 at the Radiation Measurements Laboratory (RML). The system consists of 4 measurement channels and one reference channel, each with a 2-pi proportional gas flow detector and the measurement channels each have an automated sample changer. This legacy system is crucial to the continued operations of ATR and has undergone some minor hardware upgrades since 1965, however the system presently relies on custom control boards, custom gas ion chambers, analog amplifiers/discriminators, and a user interface (UI) for the system written in outdated code. Much of the equipment and software is custom with no commercial replacements or support and limited documentation. The existing control software requires an operating system that is no longer supported, creating more vulnerabilities to continued operations. A project is underway with a third-party vendor to design, build, and document a new control and data acquisition system (CDAS) for the FWMS. The new upgrade will replace the control system, computer, UI, sample changer motors, and main power supply while maintaining the interface with existing detector hardware. The upgraded system will be operated in parallel with the current hardware and software to conduct validation testing. This equipment upgrade demonstrates the commitment at ATR to ensuring successful operations and potential future research reactors at INL.

46 - INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AN↗
DOE OSTI2019

Technical Equivalency Evaluation of Lynx Electronics and Apex-InVivo software for the LLNL Whole Body Scanning Bed System

Lawrence Livermore National Laboratory (LLNL) Whole Body Counter (WBC) is replacing the following components for its scanning bed counting system: NIM electronics with modern Lynx electronics, The Alpha Workstation running VMS with a PC running Windows , and Abacos-Plus in-vivo counting software with the new Apex-InVivo software. These new components provide a compact design, extend the lifecycle of the system as a whole, reduce maintenance and user time, an improve overall ease of use. The technical evaluations and results described in this document demonstrate that the total system, incorporating these new components, provides results that comply with Internal Dosimetry requirements.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2021

Predicting Deformation of Parts from HP Printers due to Sintering with Modeling and Simulation (CRADA NFE-18-07363)

This work proposes to leverage Oak Ridge National Laboratory’s (ORNL) sintering simulation software developed at the Manufacturing Demonstration Facility to 1) investigate the deformation of sintered parts and 2) to develop fast surrogate-based optimization methods for HP Inc. This includes the development of ORNL’s software on HP systems, integration and testing with HP part geometries and material properties, the development of surrogate models based on the results of the simulation, and modifications to the software based on feedback from HP. The expected impacts of this work are a drastic reduction in print costs, an increase in printer throughput, and the identification of novel, possibly exotic materials for 3D printing. This work will also decrease the time to market for HP’s print technology by decreasing the amount of resources and type dedicated to physical testing.

42 ENGINEERING↗