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At least 1,603 records · Page 89

All sources NASA NTRS DOE OSTI

DOE OSTI2026

Activating magnetite ores for aqueous ironmaking at high current densities

Low-temperature electrochemical cells reducing iron oxides to metal in alkaline electrolytes can support fully electrified steelmaking processes. Previous studies on these cells have primarily focused on high-surface-area hematite, Fe 2 O 3 , reactants whereas attempts to reduce suspensions of magnetite, Fe 3 O 4 —one of the two feedstocks for existing ironmaking reactors—have generally been limited to low rates of reaction (<30 mA cm −2 ). Here, in this study, we control the crystalline domain size of Fe 2 O 3 and Fe 3 O 4 particles in 10 M NaOH electrolytes to study how the nanoscale morphology of oxides controls the rate of electrochemical ironmaking. Rotating-ring disk electrode measurements of Fe 2+ , in situ Raman spectroscopy of the electrode surface, and ex situ electron microscopy were consistent with a hypothesized passivation process at Fe 3 O 4 surfaces that may prevent the continuous formation of soluble intermediates. Sufficiently small (<100 nm diameter) oxide particles yielded Fe partial current densities >160 mA cm −2 , a fivefold increase relative to previously reported rates for Fe 3 O 4 suspensions and comparable to active Fe 2 O 3 . Electron microscopy revealed that electrodeposited films were composed of micron-scale crystalline Fe domains with a porous film of Fe 3 O 4 nanoparticles and supports a model where Fe is grown primarily from soluble Fe 2+ intermediates. Based on these insights, inactive blast-furnace-grade iron-oxides were transformed into high surface area nanoparticles (1.6 to 229.2 m 2 g −1 ) via reprecipitation, leading to a ninefold enhancement in faradaic efficiency and an Fe partial current density of 120 mA cm −2 . When integrated with chlor-iron cells producing reagents for reprecipitation, this approach could lead to a cost-competitive process for electrochemical ironmaking from industrially relevant feedstocks.

TEM↗

DOE OSTI2025

Disassembly and Initial Post-Irradiation Examination of the Mini99 Experiment

The Miniature Fuel Molybdenum-99 (Mini99) experiment is a first-of-a-kind experiment in the High Flux Isotope Reactor. Mini99 is the first MiniFuel irradiation experiment to use liquid metal inside the MiniFuel subcapsule for additional temperature control. In this work, the initial post-irradiation examination is reported. As-built simulations were conducted and compared with designs, and physical post-irradiation examinations were conducted to determine actual test temperatures via SiC thermometry samples, to evaluate the extent of fission gas release from the samples, and to remove the fuel from the subcapsules. As-built simulations compared well with experiment designs and were in reasonable agreement with the measured temperatures inferred from the passive SiC thermometers. Additionally, fission gas release values were relatively small (on the order of 3%); one sample reached approximately 8%. Although some of this behavior is attributable to recoil, most of the gas release was likely caused by the physics of the fuel behavior during the test. Finally, disassembly showed that at least four of the fuel samples degraded into small components and granules at some point during the test, with the state of the other two samples unproven. Additional work is planned for further examination of these samples via various microscopy techniques.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

NASA NTRS1970

Efficient/reliable dc-to-dc inverter circuit

Feedback loop, which contains an inductor in series with a saturable reactor, is added to a standard inverter circuit to permit the inverter power transistors to be switched in a controlled and efficient manner. This inverter is applicable where the power source has either high or low impedance properties.

Pasciutti, E. R.↗

NASA NTRS1982

Preparation of large-particle-size monodisperse polystyrene latexes in microgravity

Three large-particle-size monodisperse latexes (3.44, 4.08, and 4.98 micron diameter) were prepared in an automated four-reactor apparatus on the third orbital mission of the 'Columbia' begun on March 22. Comparison with ground-based controls showed that the 4.98 micron-size flight sample was the more uniform; the uniformity at the other two sizes was about the same. The rates of polymerization in microgravity and on earth were the same within experimental error, demonstrating that radical-initiated vinyl addition polymerizations are unaffected by the weightless environment.

Vanderhoff, J. W.↗

NASA NTRS1988

Neutron effects on the electrical and switching characteristics of NPN bipolar power transistors

The use of nuclear reactors to generate electrical power for future space missions will require the electrical components used in the power conditioning, control, and transmission subsystem to operate in the associated radiation environments. An initial assessment of neutron irradiation on the electrical and switching characteristics of commercial high power NPN bipolar transistors was investigated. The results clearly show the detrimental effects caused by neutron irradiation on the electrical and switching characteristics of the NPN bipolar power transistor.

Frasca, Albert J.↗

NASA NTRS1989

Conceptual arcjet system design considerations for the SP-100 mission

Conceptual spacecraft design considerations for using 30-kW class arcjet engines as an active load for a 100-kWe SP-100 space nuclear reactor power system flight experiment are discussed. These design considerations involve configuration trades; interfaces with power, attitude and velocity control, thermal management, orbit insertion and reaction control propulsion systems; imposed environments concerned with electromagnetic noise, survivability, and ammonia exhaust plumes; and associated diagnostic instrumentation.

Zafran, Sidney↗

DOE OSTI2016

Progress In Developing Laser Based Post Irradiation Examination Infrastructure

To be able to understand the performance of reactor fuels and materials, irradiated materials must be characterized effectively and efficiently in a high rad environment. The characterization work must be performed remotely and in an environment hostile to instrumentation. Laser based characterization techniques provide the ability to be remote and robust in a hot-cell environment. Laser based instrumentation also can provide high spatial resolution suitable for scanning and imaging large areas. The INL is currently developing three laser based Post Irradiation Examination (PIE) stations for the Hot Fuel Examination Facility at the INL. These laser based systems will characterize irradiated materials and fuels. The characterization systems are the following: Laser Shock Laser based ultrasonic C-scan system Gas Assay, Sample, and Recharge system (GASR, up-grade to an existing system). The laser shock technique will characterize material properties and failure loads/mechanisms in various materials such as LWR fuel, plate fuel, and next generation fuel forms, for PIE in high radiation areas. The laser shock-technique induces large amplitude shock waves to mechanically characterize interfaces such as the fuel-clad bond. The shock wave travels as a compression wave through the material to the free (unconfined) back surface and reflects back through the material under test as a rarefaction (tensile) wave. This rarefaction wave is the physical mechanism that produces internal de-lamination failure. As part of the laser shock system, a laser-based ultrasonic C-scan system will be used to detect and characterize debonding caused by the laser shock technique. The laser ultrasonic system will be fully capable of performing classical non-destructive evaluation testing and imaging functions such as microstructure characterization, flaw detection and dimensional metrology in complex components. The purpose of the GASR is to measure the pressure/volume of the plenum of an irradiated fuel element and obtain fission gas samples for analysis. The study of pressure and volume in the plenum of an irradiated fuel element and the analysis of fission gases released from the fuel is important to understanding the performance of reactor fuels and materials. This system may also be used to measure the pressure/volume of other components (such as control blades) and obtain gas samples from these components for analysis. The main function of the laser in this application is to puncture the fuel element to allow the fission gas to escape and if necessary to weld the spot close. The GASR station will have the inherent capability to perform cutting welding and joining functions within a hot-cell.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2017

How Useful Are Imaging and Remote Sensing Technologies in Decontamination and Decommissioning? - 17363

The examples of application of imaging and remote sensing technologies during decontamination and decommissioning works in nuclear environment at Kurchatov Institute site are presented. The systems which realized these technologies are: radiometric system installed on mechanical arm of Brokk remote controlled vehicle, collimated scanning spectrometric system and portable gamma cameras of different types. Objects of application: facilities of research reactors at characterization and under dismantling, temporal radwaste storages during discharge and remediation. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2025

Development of a microwave-assisted downdraft moving-bed gasifier for continuous processing of lignite and biomass chars

This research illustrates a microwave-assisted downdraft moving-bed gasifier for the first time. Such design enables continuous solid gasification process. An adjustable auger was applied to control the solid removal rate and the gas-solid interaction time. Both lignite and biomass chars were investigated to determine the capability of the current system for low-tar feedstocks with different densities. Here, the presented reactor design was able to operate continuously for 3 hours and 20 minutes under 700 ℃ and atmospheric pressure, with air as the gasifying agent. For yellow pine char, the processing rate could reach 34.1 grams per hour with decent syngas production. The downdraft moving-bed design shows better cold gas and syngas production efficiencies compared to the common fixed-bed design, due to controllable residence time and more homogeneous microwave heating. The limitations of the current design and the direction of novel microwave-assisted chemical reactor design were discussed. This novel reactor design provides a way to improve the efficiency of microwave-assisted gasification process and shows its potential to be incorporated into other established chemical reaction processes as a modular add-on.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

NASA NTRS1994

An Integrated Fault Tolerant Robotic Controller System for High Reliability and Safety

This paper describes the concepts and features of a fault-tolerant intelligent robotic control system being developed for applications that require high dependability (reliability, availability, and safety). The system consists of two major elements: a fault-tolerant controller and an operator workstation. The fault-tolerant controller uses a strategy which allows for detection and recovery of hardware, operating system, and application software failures.The fault-tolerant controller can be used by itself in a wide variety of applications in industry, process control, and communications. The controller in combination with the operator workstation can be applied to robotic applications such as spaceborne extravehicular activities, hazardous materials handling, inspection and maintenance of high value items (e.g., space vehicles, reactor internals, or aircraft), medicine, and other tasks where a robot system failure poses a significant risk to life or property.

robotics fault-tolerant control systems↗

DOE OSTI2019

Integrated operation of steady-state long-pulse H-mode in Experimental Advanced Superconducting Tokamak

Recent Experimental Advanced Superconducting Tokamak (EAST) experiments have successfully demonstrated a long-pulse steady-state scenario with improved plasma performance through integrated operation since the last IAEA FEC in 2016. A discharge with a duration over 100 s using pure radio frequency (RF) power heating and current drive has been obtained with the required characteristics for future long-pulse tokamak reactors such as good energy confinement quality (H 98y2 ~ 1.1) with electron internal transport barrier inside ρ < 0.4, small ELMs (frequency ~100–200 Hz), and good control of impurity and heat exhaust with the tungsten divertor. The optimization of X-point, plasma shape, the outer gap and local gas puffing near the low hybrid wave (LHW) antenna were integrated with global parameters of BT and line-averaged electron density for higher current drive efficiency of LHW and on-axis deposition of electron cyclotron heating in the long-pulse operation. More recently, a high βP RF-only discharge (β P ~ 1.9 and β N ~ 1.5, $\langle$n e $\rangle$/n GW ~ 0.80, f bs ~ 45% at q 95 ~ 6.8) was successfully maintained over 24 s with improved hardware capabilities, demonstrating performance levels needed for the China Fusion Engineering Test Reactor steady-state operation. In this work, a higher energy confinement is observed at higher β P and with favorable toroidal field direction. Towards the next goal (≥400 s long-pulse H-mode operations with ~50% bootstrap current fraction) on EAST, an integrated control of the current density profile, pressure profile and radiated divertor will be addressed in the near future.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2017

Thermal Evaluations for Hypothetically Drain-Down Spent Fuel Storage Facility at SRS

The L-Area Spent Nuclear Fuel (SNF) underwater storage facility at Savannah River Site (SRS), referred to as L-Basin SNF facility, has an array of Vertical Tube Storage (VTS) racks and a group of High Flux Irradiation Reactor (HFIR) fuel cores. The underwater facility is integrally designed as a wet SNF storage system with a chemically-controlled water pool. The SNF storage racks are seismically qualified. The L-Basin pool holds about 3.4 million gallons of water with concrete wall thickness of about 5 ft and water depths of 17 ft to 30 ft. The facility has mainly two different elevation levels of 30 ft and 17 ft below the ground level. At the level 30 ft below the ground, a maximum of 110 racks of VTS and 5 racks of Dry Cave can be stored. At the level of 17 ft below the ground, 120 HFIR cores can be stored, considering that each rack contains two HFIR cores. In case of a hypothetical accident, all cooling water in the storage pool may be drained away, resulting in dry SNF storage condition. The primary objective of the work is to estimate the maximum fuel temperature in a conservative manner when the L-Basin facility is completely dry due to the hypothetical drain-down of the pool water. A three-dimensional Computational Fluid Dynamics (CFD) modeling approach was taken for the thermal evaluations of the water-drained SNF storage facility. The modeling calculations were performed, assuming that all of the storage racks in L-Basin are fully loaded with SNF assemblies containing decay heat sources. In the case of full loading, the L-Basin facility stores 14,600 SNF assemblies in VTS racks and 120 cores in HFIR racks. The thermal analysis results of the L-Basin model yielded maximum temperatures of VTS and HFIR fuel assemblies for different fuel loading scenarios and boundary conditions of the SNF storage racks.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2025

Progress on the MARVEL Cybersecurity by Design Model-Based Systems Engineering Project

Formal model-based systems engineering (MBSE) combines a model, systems thinking, and systems engineering to visually depict the boundaries, context, and behavior of interconnected systems, facilitating effective design, development, and utilization of engineered systems throughout the systems engineering lifecycle. Although nuclear reactor vendors employ these tools to integrate functionality, performance, and safety, they are not yet addressing digital risk concerns introduced by use of operational technology, such as digital instrumentation and control systems. To accomplish this objective, the Microreactor Applications Research Validation and EvaLuation (MARVEL) microreactor was used as an MBSE case study. This real-world application provides a first-of-a-kind opportunity to demonstrate the benefits of integrating digital risk and cybersecurity into the MBSE design process of a nuclear reactor. This paper provides an update of the ongoing MARVEL Cyber MBSE project as it specifically relates to the integration of digital risk management and cybersecurity by design.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2020

Contaminant Migration Modeling to Support the In Situ Decommissioning of Hardened Nuclear Facilities - 20215

The decommissioning of hardened nuclear facilities provides a unique challenge in balancing current worker dose with exposure to future receptors and media. In situ decommissioning (ISD) is a cost-effective and safe option for the closure of such facilities but requires a detailed understanding of the threat that radioactive and hazardous constituents remaining in the facilities pose to groundwater and surrounding surface water. The Savannah River Site (SRS) extensively employs contaminant migration models within the decommissioning process to develop site-specific removal, grouting, and monitoring strategies in support of safe and effective final end states of nuclear facilities. This work discusses the use of contaminant migration modeling in the ISD process at SRS, including specific examples from the closure of P-Reactor and R-Reactor, Building 235-F, F-Canyon and F/H Laboratory Complex. Migration models developed during the closure of P-Reactor and R-Reactor were divided into four source areas: the reactor vessel, the process area, the disassembly basin, and the purification wing. Each source area was assigned a specific inventory and migration pathway and was then subject to varying hypothetical removal, capping, and grouting schemes. Modeling indicated that groundwater Maximum Concentration Limits (MCLs) might be exceeded if no action was taken for eleven and ten constituents of concern (COCs) at P-Reactor and R-Reactor, respectively. These exceedances could occur in as few as 200 to 500 years. Alternatively, the migration modeling demonstrated that the selected ISD actions reduced contaminant mobility which allowed for significant radioactive decay and resulted in fewer predicted exceedances of groundwater MCLs (five COCs for P-Reactor; eight COCs for R-Reactor). In response to the modeling results for P-Reactor and R-Reactor, effectiveness monitoring programs were developed to target contaminants, specific to each reactor, that may migrate to groundwater. Contaminant migration modeling also revealed that roof collapse was a large factor in the release of COCs to the environment, giving rise to roof improvements, and an inspection and vegetation control program to ensure roof stability over time. Contaminant migration modeling is also aiding in the closure planning for hardened facilities in F Area. Building 235-F housed the Actinide Billet Line, which produced Np-237 billets for irradiation in SRS reactors, and the Plutonium Fuel Form (PuFF) facility that produced Pu-238 heat sources for the space program. As a result of these missions, areas within Building 235-F contain considerable residual amounts of both Pu-238 and Np-237. Contaminant migration modeling of Building 235-F was originally performed in 2012 to identify the feasibility of ISD and the amount of radioactive material removal required to prevent the exceedance of groundwater MCLs. The original model indicated that a 60% reduction in the PuFF facility Pu-238 inventory could keep groundwater concentrations below standards, while Np-237 did not pose a threat to groundwater. However, updates to the model with an emphasis on source impact pathways revealed that, due to the orientation of the source areas relative to groundwater flow, no amount of reasonable removal of Pu-238 would keep groundwater concentrations below MCLs and that Np-237 could be a large contributor to localized MCL exceedances. With this insight, the refined 2019 model is being used to assist in the development of grouting plans specific to each facility source area, where bentonite may be utilized to slow the migration of Pu-238 and its daughter products from the PuFF facility and a reducing grout may decrease Np-237 transport by ensuring the nuclide remains in the less mobile +IV oxidation state. The beginning phases of contaminant migration modeling are underway for F-Canyon and associated facilities using lessons learned from the Reactors and Building 235-F. A contaminant migration pathway, similar to the pathway used for the reactor vessels, is being developed for the hot and warm canyons. The F-Canyon inventories are being spatially refined to identify specific and localized source zones, comparable to Building 235-F, that may impact groundwater. Contaminant migration modeling provides vital input within the ISD process, from facility investigation to post closure effectiveness monitoring, making it a valuable tool in the development of safe and effective end states for hardened nuclear facilities at SRS. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2017

The effect of support functions in a computerized procedure system on interface management tasks - 235

By adapting digital and computer technology, the design of main control rooms (MCR) in nuclear power plants has been changed. When operators operate a nuclear power plant, the tasks using soft controls in advanced MCRs include primary tasks and secondary tasks. Primary tasks are to control nuclear power plant devices and secondary tasks are to find appropriate screens or devices by screen navigations and selections before performing the primary task. The secondary task including interface management tasks is newly introduced in advanced MCRs. In this paper, the tasks are analyzed based on the simulation data of APR1400 which is one of the digitalized nuclear power plants. For analyzing tasks, the unit tasks are divided into three types: navigation, manipulation, and acknowledgement. The primary task represents the manipulation of devices and the secondary task indicates the navigating screens, opening control panels, closing control panels, and confirming the acknowledging alarms. To analyze the effect of support functions of a computerized procedure system (CPS), one of the functions which is expected to have a significant effect on the secondary tasks was considered, that is 'related screen link button.' Two factors, 'ratio of screen alteration' and 'ratio of using screen link button', were used to see the effect of 'related screen link button'. From the task analysis result, it was observed that the secondary tasks which is one of the unique features in advanced MCRs have a quite much portion than that of the primary tasks. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2019

Addressing Nuclear I&C Modernization Through Application of Techniques Employed in Other Industries

This research focuses on presenting advancements that have been made by non-nuclear I&C OT vendors and potential savings that could be realized by their optimized nuclear industry application. To promote understanding and to bound the initial scope for this effort, a pilot vendor and an associated pilot implementation were chosen as an integrated, presentation example for illustration. This integrated example is unique in that a utility in the United States (Duke Energy) engaged a traditional non-nuclear vendor (Honeywell Process Solutions) to install a non-safety, Distributed Control System (DCS) in four of its nuclear units. Duke Energy is currently exploring performing a “technology refresh” of these installed systems as part of a program to address digital obsolescence. This research presents techniques developed by distributed control system vendors to enable such upgrades and potential methods to employ them in a nuclear industry context. The objectives of such an effort are to minimize the utility cost both to refresh the systems and to put the systems into a repeatable refresh cycle for future obsolescence management. This will enable the utility to maximize the benefits of expanded use of their DCS as part of a larger plant strategy to eliminate obsolete I&C systems, reduce Operating and Maintenance (O&M) costs, and improve operational performance through digitalization.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2025

Dynamic Modeling of a Fast Spectrum Molten Salt Reactor Integrated Energy System

Liquid-fueled molten salt fast reactors and nuclear-powered integrated energy systems (IESs) have the potential to play a pivotal role in the green energy transition. However, these systems have little to no operating experience. There is therefore increased interest and value in modeling and simulating these systems. The IES dynamic model developed in this work utilizes a lumped-parameter control volume methodology to investigate the behavior of the IES in a variety of accident scenarios. The results provide initial evidence for the potential inherent safety of the advanced reactor because of temperature-dependent reactivity feedback and the efficiency of hydrogen and electricity production at the high temperatures provided by the advanced reactor.

MCFR↗

DOE OSTI2016

NECP-Cypress: Development and Benchmarking of a 3D PWR Core-Analysis Code

As one of the key components of nuclear reactor design and safety analysis software for Pressurized Water Reactor (PWR), the advance of the core-analysis code indicates the nuclear power development of the corresponding country. Hence, Nuclear Engineering Computational Physics Lab. (NECP) developed a PWR core-analysis code NECP-Cypress. Currently, it contains several modules including a generalized cross section module, a diffusion module using the Variational Nodal Method (VNM), a depletion-calculation using the traditional macroscopic module, a thermal-hydraulics module solving the single channel heat transfer equation and 1D heat conduction equation. With the VNM employed, NECP-Cypress can eliminate the control rod cusping effect and provide a better pin power reconstruction. In this paper, we firstly introduce the modules of the code NECP-Cypress. Secondly, two large-scale PWR core benchmarks were employed to demonstrate the ability of the code by comparing with SIMULATE-3, including the BEAVRS problem and a pseudo CAP1400 reactor core problem. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗