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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 1,549 records · Page 86
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Self-Navigating Robotic Assistant for Long-term Wide Area Floor Contamination Monitoring - 18213

For Routine Contamination Testing (RCT) of an operational nuclear facility or one undergoing Decontamination and Decommissioning (D and D), workers utilize specified travel corridors (enclosed hallways or areas designated with tape or rope in open spaces) to minimize the risk of spreading contamination. Radiation health technicians must continually monitor these areas for contamination to minimize worker dose and prevent its spread to uncontrolled areas. For many sites, it is impractical for the technician to adequately survey such areas. The Nuclear Robotics Group (NRG) at UT Austin has developed a self-navigating robotic platform equipped with radiation detectors to augment the technician to improve both survey coverage and frequency. While maintaining a controlled velocity and floor-to-sensor distance, the system takes radiation measurements to identify and/or localize contamination sources. Readings above background levels are reported to a remote workstation to allow technicians to then respond accordingly. The completion of these mundane tasks by a robot allows the trained technicians to focus instead on operational tasks, D and D, or responding to alarms. The robot generates a map using a Lidar range sensor with established Simultaneous Localization and Mapping (SLAM) techniques. Range data is then compared to previously generated maps for robot localization and autonomous navigation. To maximize coverage, the robot executes a Complete Coverage Path Planning (CCPP) algorithm using a static map that is discretized into cells to assure comprehensive yet efficient coverage. As the system executes the generated plan, it builds a radiation map of the area for the technician to validate and verify the results. If contamination is detected, the robot will minimize the spread of contamination by stopping, sounding an alarm, and alerting the remote technician. The proposed system allows for long-term deployment with minimal human intervention. It is robust to both highly dynamic obstacles (i.e. personnel) and low dynamic changes (i.e. temporally stored cart) in the environment. This is accomplished through continuous spatiotemporal updating of the static map, and the utilization of a supervisory controller that allows the system to reliably respond to unexpected events or states (i.e. low battery, closed door, etc.). While the supervisory controller can respond to temporary complications, the system also recognizes the environment can change slowly over time necessitating that the CCPP automatically be updated to account for observed 'permanent' changes to the environment. To do this, the robot uses intermediate probabilistic data to make intelligent planning decisions offline while still maintaining local reactive obstacle avoidance and safety features. The self-updating of the map allows the robot to run longer without the need to remake the map, leading to more consistent monitoring data and less maintenance from the human technician. This paper will review the current system hardware, the CCPP algorithm, the supervisory controller, the probabilistic algorithm for identifying permanent changes, and recent experiments at UT Austin to experimentally validate the reported sensor coverage. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Commercialization of High-Density High Assay Low Enriched Uranium Fuel Systems

The Office of Reactor Conversion and Uranium Supply (NA 231) at the National Nuclear Security Administration leads the conversion effort for the United States High Performance Research Reactors (USHPRR). These reactors are the final civilian reactors in the US to transition from High Enriched Uranium (HEU) to high assay low enriched uranium (HALEU). Each of these reactors represents unique capabilities and no currently available fuel system meets their needs for conversion. The Fuel Fabrication (FF) Pillar of the USHPRR project is responsible for the fabrication of experimental elements, conversion elements, and establishing a commercial economical production capability. FF is also responsible to share with the domestic and international community the theoretical knowledge gained. Other pillars within the USHPRR project provide the experimental and conversion fuel designs, assist the reactors with licensing activities, and ensure the entire fuel cycle is evaluated. Over the last decade, FF has worked with the production partners at Y-12 National Security Complex (Y-12) and BWXT Nuclear Operations Group, Research and Test Reactors (BWXT). Y-12 has begun processing the alloy feedstock for the conversion elements with a qualified process. BWXT has started the final fabrication of the experimental elements. Once the experimental elements are complete, BWXT will begin conversion element fabrication. The FF Pillar resides at Pacific Northwest National Laboratory (PNNL) and uses PNNL, universities, commercial vendors, and the DOE national laboratory system to evaluate process development activities to improve the process steps. FF supports the fabrication of two high density fuel systems, monolithic U-10Mo (Figure 1) and Uranium Silicide (Figure 2). The U-10Mo fuel system is further along the development process. FF assists in long term planning with the production partners. This includes ramping production of the elements from experimental quantities to annual steady state needs. As part of the ramp up, opportunities to improve yield and product quality are identified to ensure the fuel systems are cost effective.

Catalan, Michael A. [BATTELLE (PACIFIC NW LAB)]↗
DOE OSTI2024

Initial Development of Fusion Magnet Simulation Capabilities for Performance and Safety Evaluation Using the MOOSE Framework

Fusion energy holds the promise of being a transformative technology as a carbon-neutral, sustainable source of energy. Whole device modeling and the development of fusion digital twins will be increasingly important for emerging fusion device concepts at both national laboratories and within the commercial fusion industry. However, meeting the challenge of whole device modeling of fusion energy devices requires robust, multiphysics, multiscale modeling and simulation technologies capable of running on large-scale supercomputers. Detailed analysis of individual systems at-scale is also required to ensure safe and efficient operation as well as provide the safety basis for future device designs and licensing activities. In a tokamak, toroidal and poloidal magnets confine and shape the fusion plasma to promote the fusion reaction. High plasma temperatures and high magnetic field requirements in modern design concepts (leading to high amounts of energy stored within each magnet) impose electrical, thermal, and mechanical loads on the magnet components, which in turn impacts the safety considerations of the magnet and their supporting systems. Idaho National Laboratory (INL) has a history of working in this space, including development and benchmarking of the Magnetic System Circuitry Analysis Program (MSCAP) and Magnet Arcing (MAGARC) codes to study magnet quench events; notably, MAGARC was used to study quenching during the ITER Engineering Design Activity. However, these legacy codes and capabilities are not parallel and scalable, and new tools are required for future advances in this area, which leads to the INL-developed Multiphysics Object-Oriented Simulation Environment (MOOSE) framework. Developed originally for fission reactor systems under United States Department of Energy, Office of Nuclear Energy modeling and simulation programs, the MOOSE framework has been well-suited to multiscale, multiphysics modeling and simulation needs for nuclear systems. The framework is open-source, well-tested, under continuous development and deployment, and developed to a Nuclear Quality Assurance, Level 1 software quality standard. MOOSE has also been used in the fusion space previously in several projects: INL’s Tritium Migration Analysis Program, Version 8 (TMAP8) for tritium migration, UK Atomic Energy Authority’s A Unified Resource for OpenMC (fusion) Reactor Applications (AURORA) code for fusion thermo-mechanical and neutronics analysis, and Argonne National Laboratory’s Cardinal for high-fidelity computational fluid dynamics and neutronics. However, to model superconducting magnets, several MOOSE enhancements are required: additions to the current MOOSE electromagnetic capabilities, new material libraries for superconductors of interest (such as YBCO), as well as fusion-specific models for thermo-mechanics. This talk will discuss initial development activities to build these capabilities in MOOSE, focusing on initial validation and benchmarking activities. Proposed coupling workflows and future work to support the simulation of fusion magnets and magnet structural assemblies for performance and safety evaluation in MOOSE will also be discussed.

70 - PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2026

Comparisons of triple-differential cross sections for quasielastic-like $ν_μ$-hydrocarbon interactions using $\langle E_ν\rangle \sim$ 3~GeV versus $\sim$ 6~GeV beams in MINERvA

Neutrino charged-current quasielastic-like scattering, a reaction category extensively used in neutrino oscillation measurements, receives contributions from single nucleon knockout processes, multinucleon processes, and inelastic scattering with subsequent rescattering or absorption in the nucleus to produce only nucleons in the final state. In this article, comparisons are presented of the same measurement in two different wideband neutrino beams: one beam peaks near 3 GeV with few neutrinos above 6 GeV; the other peaks near 6 GeV with few neutrinos above 10 GeV. Comparisons of differential cross sections in muon and proton kinematics for these two exposures probe deviations from free-neutron scattering that arise from the processes involving the nuclear medium, and provide a test of neutrino interaction models used to infer neutrino energies in oscillation experiments. Discrepancies are observed between the data and predictions that point to overestimates of the final state interactions of both protons and charged pions in quasielastic-like events.

Ruterbories, D. [Rochester U.]↗
DOE OSTI2018

Field Test of the Methodology for Succession Planning for Technical Experts

This report complements A Methodology for Succession Planning for Technical Experts (Ron Cain, Shaheen Dewji, Carla Agreda, Bernadette Kirk, July 2017), which describes a methodology for identifying and evaluating the loss of key technical skills at nuclear operations facilities. This report targets the methodology for identifying critical skills, hereafter referred to as “core competencies”. The methodology has been field tested by interviewing selected retiring subject matter experts (SMEs).

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2019

Modeling the IFR-1 Experiment: A BISON Metallic Fuel Benchmark

As the 1970s nuclear reactor fleet slowly begins the decommission process, the US Department of Energy Office of Nuclear Energy (DOE-NE) aims to stimulate and support the construction and implementation of new nuclear reactors across the country in hopes of continuing the generation of clean, reliable, resilient energy. As a result, various efforts have been introduced to accelerate this process, including the Versatile Test Reactor project (VTR) and the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

LANL Test Engineering’s Structural Dynamics Research and Development Efforts [Slides]

LANL's priority roles are serving as a nuclear weapons design agency and a nuclear weapons production agency; addressing nuclear threats; and performing national security science, technology, and engineering. LANL’s Test Engineering organization serves the national mission of maintaining a strategic nuclear deterrent by providing high-quality, empirical evidence through the execution and assessment of weapons system and component testing. Shock and vibration environmental testing provides one key piece of evidence for evaluation and qualification of these systems and components in the service environments incurred during a lifetime in the stockpile. LANL’s Test Engineering group continuously pursues research and development in structural dynamics to improve shock and vibration testing capabilities, processes, and analysis techniques to increase the fidelity of service environment recreations in laboratory testing settings. This talk will highlight several ongoing projects to represent these R&D efforts.

42 ENGINEERING↗
NASA NTRS1990

A new method for using Cf-252 in SEU testing

A system using Cf-252 and associated nuclear instrumentation has determined the single-event upset (SEU) cross section versus linear energy transfer (LET) curve for several 2K x 8 static random access memories (SRAMs). The Cf-252 fission fragments pass through a thin-film organic scintillator detector (TFD) on the way to the device under test (DUT). The TFD provides energy information for each transiting fragment. Data analysis provides the energy of the individual ion responsible for each SEU; thus, separate upset cross sections can be developed for different energy and mass regions of the californium spectrum. This californium-based device is quite small and fits onto a bench top. It provides a convenient and inexpensive supplement or alternative to accelerator and high-altitude/space SEU testing.

Costantine, A.↗
DOE OSTI2018

Verification and Validation of Optimization of Replacement Portal Monitoring Replay Tools

The Los Alamos National Laboratory and the Johns Hopkins Applied Physics Laboratory (APL) are both developing optimization software to fulfill the objective of Phase III of the Replacement Portal Monitoring (RPM) Program for the Department of Homeland Security. The objective of the Phase III of the RPM Program is to conduct an evaluation of three RPM systems designed to enhance mission effectiveness, gain operational efficiencies, and to address emerging mission needs against the Countering Weapons of Mass Destruction (CWMD ) nuclear threat basis and special nuclear material (SNM) used in performance testing. As such this task involves the use of global optimization algorithms.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2025

Operational Guidelines for Use of Radiation Portal Monitors in Nuclear Facilities

This article provides guidelines (not requirements) for the testing, operation, and maintenance of radiation portal monitors used to scan pedestrian and vehicular traffic entering or exiting nuclear facilities to prevent unauthorized removal of nuclear material. The intended audience is facility managers, supervisors, and operators who are responsible for establishing and maintaining radiation detection systems, who are likely well acquainted with physical security procedures, but may not have a familiarity with the operation of radiation detection.

Enders, Alexander [Oak Ridge National Laboratory (↗
NASA NTRS2018

Electrically Heated Testing of the Kilowatt Reactor Using Stirling TechnologY (KRUSTY) Experiment Using a Depleted Uranium Core

The Kilopower project aims to develop and demonstrate scalable fission-based power technology for systems capable of delivering 1 to 10 kW of electric power with a specific power ranging from 2.5 to 6.5 W/kg. This technology could enable high-power science missions or could be used to provide surface power for manned missions to the Moon or Mars. NASA has partnered with the U.S. Department of Energy's National Nuclear Security Administration, Los Alamos National Laboratory, Nevada National Security Site (NNSS), and Y−12 National Security Complex to develop and test a prototypic reactor and power system using existing facilities and infrastructure. This technology demonstration, referred to as the "Kilowatt Reactor Using Stirling TechnologY (KRUSTY)," will undergo nuclear ground testing by the end of calendar year (CY) 2017 at the NNSS. The 1-kWe variation of the Kilopower system was chosen for the KRUSTY demonstration. The concept for the 1-kWe flight system consists of a 4 kWt highly enriched uranium-molybdenum reactor operating at 800 degC coupled to sodium heat pipes. The heat pipes deliver heat to the hot ends of eight 125-W Stirling convertors producing a net electrical output of 1 kW. Waste heat is rejected using titanium-water heat pipes coupled to carbon composite radiator panels. The KRUSTY test, based on this design, uses a prototypic highly enriched uranium-molybdenum core coupled to prototypic sodium heat pipes. The heat pipes transfer heat to two Advanced Stirling Convertors (ASC−E2s) and six thermal simulators, which simulate the thermal draw of full-scale power conversion units. Thermal simulators and Stirling engines are gas cooled. The most recent project milestone was the completion of nonnuclear system-level testing using an electrically heated depleted uranium (DU) (nonfissioning) reactor core simulator at the NASA Glenn Research Center. System-level testing has validated performance predictions and has demonstrated system-level operation and control in a test configuration that replicates the one to be used at the Device Assembly Facility (DAF) at the NNSS. Fabrication, assembly, and testing of the DU core has allowed for higher fidelity system-level testing at Glenn, and has validated the fabrication methods to be used on the highly enriched uranium core that will supply heat for the DAF KRUSTY demonstration.

Briggs, Maxwell H.↗
DOE OSTI2019

Performance Evaluation of AG-1 FC Filters Under Upset Conditions and the Effects of Aging on Filters - 19501

High efficiency particulate air (HEPA) filters are routinely used in nearly every operating Department of Energy (DOE) and National Nuclear Safety Administration (NNSA) nuclear facility. These nuclear grade HEPA filters installed in DOE and NNSA confinement ventilation systems are credited as the final barrier to protect the environment, the public, and the worker from accidental release of virtually all airborne radioactive materials, and must meet the American Society of Mechanical Engineers' (ASME) Code on Nuclear Air and Gas Treatment (CONAGT) AG-1 standard. This testing activity evaluates ASME AG-1 Section FC axial flow HEPA filters, both separator and separator-less, under upset conditions. Additionally, testing efforts have analyzed the effects of filter degradation due to aging. In 1997, the Defense Nuclear Facilities Safety Board (DNFSB) raised concerns of the potential vulnerability of HEPA filters in vital safety systems, both from a design stand point and the potential for degradation of HEPA filter performance over time. The DOE Office of Environmental Management (DOE-EM) has a Cooperative Agreement with Mississippi State University (MSU) Institute for Clean Energy Technology (ICET) to conduct testing of HEPA filters. DOE-EM has established a Technical Working Group (TWG) of over thirty subject matter experts from the nuclear industry to provide input and oversight of ICET projects and activities. This has included establishment/approval of testing parameters for evaluating these filters under ASME NQA-1 compliant Test Plans. The current scope of work involves evaluation of the performance envelope for new and aged separator and separator-less designs of AG-1 Section FC axial flow HEPA filters. A range of elevated test conditions (T and RH) are used to determine the operating envelope within which DOE nuclear safety experts can credit installed HEPA filter performance, thereby establishing a risk-informed DOE service life. This paper will present preliminary test results of new and aged axial flow filters tested at ICET. Results in this paper include data collected from new separator and separator-less type FC HEPA filters and equivalent test data for a limited group of aged separator and separator-less type FC HEPA filters. Testing for both scopes of work evaluate initial filter performance under ambient conditions of temperature (60 deg. F - 80 deg. F (∼16 - ∼27 deg. C)) and relative humidity (40-60%), loading of each filter with Aluminum Trihydroxide, Al(OH)3, to a specified differential pressure, and then exposing the filter to specified elevated temperature and relative humidity conditions. In addition to the upset conditions of elevated temperature and relative humidity, the performance of HEPA filters under fire scenarios is not well characterized. ICET is currently designing a burn box and testing apparatus to study the effects of fire from burning liquids and a wide range of blends of combustible solids. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Operational Experience and Development of a Reliability Model for the ATR Demineralizer System (Slides)

The Advanced Test Reactor (ATR) is a light water reactor with aluminum clad driver fuel. Strict limits are implemented on pH, conductivity, and filterable solids to assure the performance of the driver fuel limit corrosion of the primary coolant system (PCS) pressure boundary components. A bypass demineralizer system is used to maintain PCS coolant within allowable bands for each of the aforementioned variables. Unlike many other reactors, the ATR does not have a filtration system. As such, failure of ion exchange resin in the bypass demineralizer can defeat the purpose of the system and result in high filterable solids in the PCS. Power operations with high filterable solids is prohibited by a technical safety requirement. This paper discusses recent failures of the ATR bypass demineralizer system. The testing and analysis done to deduce the cause of the failures is covered, as well as short term mitigative steps that were taken. To prevent the likelihood of future issues, historical data from the ATR and industry is used to develop a reliability model of the ATR bypass demineralizer system. Evidence of radiolytic decomposition of ion exchange resin was found in the cause investigation. As such, special attention is given to ion exchange resin failure due to high radiation environments. The reliability model is used to develop the ATR testing and maintenance program. Specifically, limits on resin service life are discussed.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
NASA NTRS1991

Upgraded NERVA systems: Enabler nuclear system

The NERVA/Rover Enabler technology enables to go on a low risk, short-term program to meet the requirements of the Mars mission and maybe some lunar missions. The following subject areas are covered: NERVA technology - the foundation for tomorrow's space missions; NERVA/Rover reactor system test sequence; NERVA engine development program; nuclear thermal reactor capability based on many related Westinghouse technology programs; investment in Rover/Nerva technology; synergistic applications of NERVA technology; flow schematic of the NDR engine; the NERVA nuclear subsystem; and technology evolution.

Farbman, Gerry↗
DOE OSTI2019

FY19 Transient Prescriptions in TREAT Full-Slotted Core

The transient reactor test facility (TREAT) is a graphite-based transient test reactor at the Idaho National Laboratory. TREAT started operation in 1959 and performed thousands of transient irradiations for the primary purpose of nuclear fuel safety research. TREAT operations were suspended in 1994 and then, several years later, resumed reactor operations in 2017 to support development of advanced nuclear fuels. During TREAT’s first era of operation the transient rod drive system and eventual Automatic Reactor Control System (ARCS) were upgraded and improved to create a transient reactor capable of unparalleled power shaping capabilities. TREAT’s first year of modern experiment work was performed in 2018 and saw the demonstration of new kinds of transients focused on modern experiment needs; especially focused on light water reactor (LWR) category transients including reactivity-initiated accidents (RIA) and loss of coolant accidents (LOCA). These transients were performed using a non-fueled neutronic mockup of a historical sodium loop which was not a perfect representation of future core configurations, but was adequate for demonstrating capabilities, successfully troubleshooting transient programming parameters, and calibrating nuclear models of the core. These physics tests were termed “transient prescription tests”. Later in 2018 the core was reconfigured to a modern fueled experiment arrangement and TREAT began performing irradiations on LWR rodlets where crucial fuel specimen nuclear heating data were obtained, and several supporting systems were commissioned. This successful effort gave way to additional fueled experiments which were performed throughout 2019.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Innovative Implementation of Consolidated Dry Fuel Storage at the INL - 19655

An innovative nuclear-fuel storage bucket design provides increased fuel storage capacity at the Irradiated Fuel Storage Facility (IFSF) on the Idaho National Laboratory (INL) Site. The IFSF is a dry storage facility that is at approximately 93% of its fuel storage capacity. This innovative fuel storage bucket provides for higher density packaging, thereby allowing for the receipt of more fuel at the IFSF and subsequent completion of the wet-to-dry fuel transfer scope at the INL's Idaho Nuclear Technology and Engineering Center (INTEC). The subject fuel originates from the Advanced Test Reactor (ATR) at the INL Site. All spent nuclear fuel (SNF) at INTEC is required by the Idaho Settlement Agreement (ISA) to be out of wet storage by December 31, 2023. The ISA is a tri-party agreement between the U.S. Department of Energy, the U.S. Navy, and the State of Idaho. Currently, the INTEC CPP-666 pools contain all of the INL's SNF remaining in wet storage, and INTEC's Spent Fuel Project is on schedule to meet the ISA deadline. Interestingly, the INL Site's environmental cleanup contract awarded to Fluor Idaho in 2016 also initially required the Spent Fuel Project to continue SNF receipts from the ATR into wet storage at the CPP-666 pools. The continued receipts of ATR SNF into the CPP-666 basin were able to be redirected to the IFSF because of the new storage package design, enabling the Spent Fuel Project to meet the ISA requirement to have all SNF out of wet storage by 2023. Complicating the need to redirect ATR receipts from wet storage is the lack of interim storage capacity at the ATR site. This lack of ATR storage capacity threatened to terminate reactor operations if the SNF could not be received at INTEC. The contractual requirement for continued ATR SNF receipts to prevent termination of reactor operations is resolved by the implementation of the innovative consolidated dry storage configuration in the IFSF. Consolidated fuel storage at the IFSF provides increased fuel storage capacity to facilitate meeting the ISA and the Fluor Idaho contract requirements. It provides for the completion of the ATR wet-to-dry fuel transfer scope and supports continued operation of the ATR by allowing the receipt of additional ATR SNF directly into IFSF dry storage. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗