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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 127 records · Page 7
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Sandwiching of MOF nanoparticles between graphene oxide nanosheets among ice grains

Current strategies to tailor the formation of nanoparticle clusters require specificity and directionality built into the surface functionalization of the nanoparticles by involved chemistries that can alter their properties. Here, we describe a non-disruptive approach to place nanomaterials of different shapes between nanosheets, i.e., nano-sandwiches, absent any pre-modification of the components. We demonstrate this with metal-organic frameworks (MOFs) and silicon oxide (SiO 2 ) nanoparticles sandwiched between graphene oxide (GO) nanosheets, MOF-GO and SiO 2 -GO, respectively. For the MOF-GO, the MOF shows significantly enhanced conductivity and retains its original crystallinity, even after one-year exposure to aqueous acid/base solutions, where the GO effectively encapsulates the MOF, shielding it from polar molecules and ions. The MOF-GOs are shown to effectively capture CO 2 from a high-humidity flue gas while fully maintaining their crystallinities and porosities. Similar behavior is found for other MOFs, including water-sensitive HKUST-1 and MOF-5, promoting the use of MOFs in practical applications. The nanoparticle sandwich strategy provides opportunities for materials science in the design of nanoparticle clusters consisting of different materials and shapes with predetermined spatial arrangements.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2022

Non-premixed counterflow methane flames in DC/AC/NS electric fields

The response of counterflow diffusion flames to sub-breakdown DC and AC electric fields, as well as their superposition with ns pulse discharge waveforms, is studied in the plane-to-plane electrode geometry. Sub-breakdown DC and low-frequency AC electric fields cause the flame displacement toward the grounded electrode, in the direction of the applied field, indicating that the body force on the positive ions exceeds that on the electrons and negative ions. As the AC frequency increases, the flame response becomes less pronounced, due to the reduction of the electrohydrodynamic (EHD) body force impulse over the AC half-period. The electric field in the electrode gap is determined by ps Electric Field Induced Second Harmonic (E-FISH) generation, with absolute calibration using sub-breakdown ns pulses overlapped with the measured electric field waveform. The results show that the electric field distribution across the flame in the current saturation regime follows the Laplacian field. This indicates that the space charge density in the gap is too low to distort the applied DC or AC field, consistent with the kinetic modeling predictions. Combining a nanosecond pulse discharge with a sub-breakdown DC field generates a diffuse plasma across the entire gap. Time-resolved and spatially resolved measurements of the electric field in the discharge indicate the ionization wave propagation between the electrodes. The present results do not exhibit a detectable flame displacement enhancement by ns discharge pulses combined with a sub-breakdown field, observed previously. Kinetic modeling calculations show that the absence of this effect in the plane-to-plane geometry is due to the rapid plasma self-shielding. This indicates that alternative electrode geometries limiting the self-shielding would be more effective for the plasma / electric field enhanced flameholding and flame stabilization applications.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2023

Shielding Analysis of a Preclinical Bremsstrahlung X-ray FLASH Radiotherapy System within a Clinical Radiation Therapy Vault

A preclinical radiotherapy system producing FLASH dose rates with 12 MV bremsstrahlung x rays is being developed at Stanford University and SLAC National Accelerator Laboratory. Because of the high expected workload of 6,800 Gy w –1 at the isocenter, an efficient shielding methodology is needed to protect operators and the public while the preclinical system is operated in a radiation therapy vault designed for 6 MV x rays. Here, in this study, an analysis is performed to assess the shielding of the local treatment head and radiation vault using the Monte Carlo code FLUKA and the empirical methodology given in the National Council on Radiation Protection and Measurements Report 151. Two different treatment head shielding designs were created to compare single-layer and multilayer shielding methodologies using high-Z and low-Z materials. The multilayered shielding methodology produced designs with a 17% reduction in neutron fluence leaking from the treatment head compared to the single layered design of the same size, resulting in a decreased effective dose to operators and the public. The conservative assumptions used in the empirical methods can lead to over-shielding when treatment heads use polyethylene or multilayered shielding. High-Z/Low-Z multilayered shielding optimized via Monte Carlo is shown to be effective in the case of treatment head shielding and provide more effective shielding design for external beam radiotherapy systems that use 12 MV bremsstrahlung photons. Modifications to empirical methods used in the assessment of MV radiotherapy systems may be warranted to capture the effects of polyethylene in treatment head shielding.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2020

Improvements to the Modeling of the TREAT Reactor and Experiments

This paper summarizes the latest improvements and lessons learned from the modeling and simulation of the transient test reactor at Idaho National Laboratory using the MAM- MOTH reactor physics application. MAMMOTH is a MOOSE-based, Finite Element Method application that specializes in the analysis of the spatial dynamics behavior of nuclear reactors. Since early 2018 several transient tests have been conducted at TREAT, thus providing the opportunity to apply and benchmark modern modeling and simulation tools. MAMMOTH was used to provide predictions of the power coupling factor between the core and the experiment for various experiments. Even though the power coupling factor predictions agree very well with the experimental data, within the bounds of the experimental uncertainty, one shortcoming was the underprediction of the total energy deposited in the core and in the experiment. Determination of the sources for this discrepancy are ongoing, but several key problems have been identified and resolved, thus providing valuable insights for future research. This paper discusses several of these lessons learned. First, the heat capacity data for the TREAT fuel has some significant problems due to limitations of the measurement techniques used circa 1960s. The sensitivity of the peak power and the total energy deposition to various representations of the heat capacity is approximately 5%. Second, the effects of the biological shield and thermal column on the modeling of the core are non-negligible, since they affect the mean generation time and the effective reflection of neutrons back into the core, which is suspected to be important during the core heat up. Matching the reactor period resolves the fact that the reduced spatial domain used in the MAMMOTH model underpredicts the mean generation time. The neutron reflection from these regions is marginally improved with the use of an albedo boundary condition. Third, modeling of the control rod movement with a multi-scheme method is introduced and its current limitations are exposed. Fourth, we explore the effects of using a homogenized model with Superhomogenization equivalence and how that differs from fully heterogeneous simulations. Finally, the energy condensation effects for this graphite core are significant. Solutions with 10 and 26 energy groups show the benefits of using a finer coarse group structure.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2026

Surrogate modeling of Monte Carlo radiation transport with convolutional neural networks for shielding optimization

Here, we present a machine learning (ML)-based surrogate model using convolutional neural networks (CNN) designed to emulate the attenuation of neutron fields as they pass through various shielding materials. This model can compute the outgoing neutron flux almost instantaneously and achieves reasonable accuracy compared to traditional Monte Carlo (MC)-based codes, which are computationally intensive. This emulator alleviates the complexity of neutron radiation transport through shielding materials by reducing the dimensionality and enables shielding optimization for a known radiation environment. This optimization process, which would have taken an unrealistic timeline due to several complex radiation transport simulations, can now be achieved in minutes, thus increasing computational capabilities in radiation shielding assessment. We demonstrate the applications of this emulator in computing effective dose rates and optimizing shielding solutions for a heavy-ion accelerator facility, such as the Facility for Rare Isotope Beams, where secondary neutrons produced via beam interactions dominate the radiation environment.

accelerator shielding↗
DOE OSTI2022

Particulate Fuel Modeling of MC 2 -3 using Iterative Local Spatial Self-shielding Method

We report a new spatial self-shielding method for particulate fuels has been developed based on disadvantage factors and implemented in the MC 2 -3 code. This method named the iterative local spatial self-shielding (ILSS) method considers the shadowing effect of randomly distributed particles on spatial self-shielding in particles through a homogenized composition region added outside the particle of interest at the center. The self-shielded cross sections of the central particle are determined iteratively since they are used in determining the cross sections of the homogenized composition region. The ILSS method was verified for infinite stochastic medium problems of single and multiple types of particles, VHTR unit cell problems, and HTTR assembly problems. The verification test results show that the ILSS method accurately predicts the stochastic particle shadowing effect and reaction rates in particles, whereas the regular array model and the stochastic collision probability method underpredict the particle shadowing effect and overestimate reaction rates in particles.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

TEAMER: DAISY Flow-noise Testing (Abstract)

Evaluating impacts of marine energy devices is generally difficult given the dynamic environment where these devices need to be placed. The University of Washington’s (UW) DAISY (The Drifting Acoustic Instrumentation SYstem) is designed to measure radiated noise around marine energy converters operating in energetic waves and currents. In currents, a primary limitation for measurement fidelity at low frequencies (< 100 Hz) is the potential for non-propagating “flow-noise” to mask propagating sound and inflate estimates of the radiated noise from marine energy converters at frequencies that overlap with hearing sensitivities of fish and some marine mammals. While free-drifting measurements help to minimize the relative velocity that produces flow-noise, significant levels were still observed during initial DAISY tests. This motivated the development of a fabric “flow shield” around the hydrophone that disrupts both flow-noise generation mechanisms proposed by the initial tests. First, the flow shield is a source of substantial drag which keeps the hydrophone package moving with the approximate velocity of the surrounding water, compensating for differential wind or current forcing on the surface expression. By minimizing relative velocity around the hydrophone, turbulence shed by the hydrophone is also minimized. Second, the flow shield creates a largely quiescent pocket around the hydrophone, minimizing advection of free stream turbulence over the hydrophone element. Field data collected in these experiments will test effectiveness of these flow shields and provide quantitative data use and deployment.

16 TIDAL AND WAVE POWER↗
DOE OSTI2020

Best Practices for Shielding Analyses of Activated Metals and Spent Resins from Reactor Operation

This report investigated best practices for performing shielding evaluations of Type B waste packages, as defined in 10 CFR Part 71 regulations on packaging and transportation of radioactive material, or packages for which the contents are not defined before loading and may include a broad range of nuclides, geometries, and non-fuel materials. The following non-fuel waste streams were analyzed: Activated metals from decommissioned commercial power reactors, including Type 304 stainless steel, reactor vessel steel, and Inconel, Control blades from boiling water reactors, Neutron-activated corrosion products on surfaces of activated metals, and Spent resins from power plant operations. Measured elemental compositions, including major constituents and impurities, for steel and Inconel samples from commercial power reactors were used in activation calculations to determine radionuclide inventories in activated metals. For a simplified cask model, 60 Co contribution to the total external package dose rate at 30 days after shutdown varied from approximately 60% to 95%, depending on the activated metal, initial cobalt impurity concentration in the metal, and the thickness of the overpack gamma shield. Its maximum contribution to the total external package dose rate of approximately 100% was reached within the time interval of 2 to 5 years after shutdown and was maintained for up to 45 to 60 years after shutdown, depending on material, initial cobalt impurity concentration, and shield thickness. Thereafter, the 60 Co contribution to external package dose rate decreased with increasing decay time. Cobalt-60 is primarily produced by neutron reactions with the cobalt impurity in steel and Inconel. Other important radionuclides in activated metals contributing to package external dose rate are radionuclides with relatively short decay times, including 51 Cr, 59 Fe, 58 Co, and 54 Mn. These radionuclides may be represented as an equivalent 60 Co activity/source because 60 Co gamma ray emissions are bounding in terms of source strength and energy to other important radionuclides identified in the analyzed activated metals. Approaches for modeling the neutron-activated corrosion products that may be attached to activated reactor components were analyzed in this report. It was demonstrated that a surface source is more conservative than a uniform volumetric source for the treatment of neutron-activated corrosion products with respect to external package dose rates. An analysis of the maximum radionuclide loadings reported on spent resins identified the radionuclides 137 Cs, 60 Co, 134 Cs, 65 Zn, and 58 Co as the primary contributors to external package dose rate. For a resin cooled for 3.08 years, the external package dose rate was entirely produced by the reported 137 Cs and 60 Co inventory. The neutron sources from actinides found on spent resins or activated metals produced negligible dose rates and may be ignored in dose rate analyses. Effects of idealized waste material, source geometry, and spatial material/source distributions on external package dose rates were determined based on dose rate results for a simplified cask model under normal conditions of transport. Type 304 stainless steel, zirconium, and aluminum with adjusted mass densities based on a maximum content weight were analyzed for material modeling effects on external package dose rate. These materials produced identical external package dose rates within the statistical uncertainties of the dose rate estimates. Among four different source geometry configurations with homogeneous material of different mass densities, uniform volumetric source distribution, and the same source strength, the geometry configuration with lowest mass density (i.e., minimum self-shielding effect among the four cases) was most conservative. Spatial source distributions that better represent localized peak 60 Co activity values were more conservative than a uniform volumetric source distribution, assuming the same weight and total source strength per package. The increase in external dose rate caused by localized source peaks can be as much as the ratio between source peak activity density to average activity density, depending on the location of the activated metal with peak activity density. Therefore, the shielding analysis may be simplified if localized peak activities and the average activity per package can be measured/determined and documented at the time of cask loading. For simplicity, external package dose rates may be determined based on average source activity and a uniform volumetric source distribution. The dose rate results from that calculation model multiplied by the ratio of peak activity density to average activity density will produce maximum dose rate values for conservative estimates.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2026

Redesign of Nuclear Material Storage Container with Additive Manufacturing

The goal of this project is to reduce worker dose and improve criticality safety by improving nuclear material storage containers. The current method of medium to long term (up to 15 years) storage of small amounts of nuclear material is to store it in a stainless-steel flat tube with a copper crush gasket. These containers use 10 bolts to fasten the lid, and the crush gasket must be replaced after every use. Stainless-steel acts as a gamma shield, but there is not any material to account for neutron dose. A redesign of this container should prioritize reducing radiation doses to handlers of the container by reducing the amount of time handlers are exposed to the container and introducing material with effective neutron and gamma shielding mechanics.

42 ENGINEERING↗
DOE OSTI2025

Challenges and Opportunities Specific to Microreactors: Compact and Light-Weight Shielding During Operation and Transportation

The problem of designing a shield for a mobile microreactor has been addressed. The principal challenges encountered in this endeavor are the dual goals of increasing radiation protection and decreasing weight. It is recognized that these two goals are contradictory. Increasing shielding naturally increases weight. Conversely, decreasing weight naturally calls for decreasing material inventory, including possibly decreasing shield mass. The operational and regulatory drivers for shielding and transportation constraints are identified and summarized. The main constraints pertain to volume, mass, radiation dose, material feasibility and availability, and decommissioning expectations. The latter are limits on residual contamination and dose rates after the reactor is removed. Progress toward solving the challenge of shielding a transportable microreactor is made by considering a tungsten tetraboride (WB4) shell surrounding the reactor vessel. The shield and a generic microreactor were modeled using the MCNP code. It was determined that the presence of a 16-cm-thick WB4 shield reduces the dose equivalent rate during operations to 3.6 mrem/h just outside of a concrete vault and dome and soil berm surrounding the reactor, which is very near the occupational dose rate limits. Without the WB4 layer, the dose rate at the same location outside the berm would be 120 mrem/h. Most importantly, the presence of the WB4 shield substantially protects the surrounding concrete and soil from activation, thus decreasing the duration of administrative control for the site following removal of the reactor from 16 to 3.4 years. At that point, the site may be decommissioned and returned to unrestricted public access. A limitation on the incorporation of a WB4 shield as an intrinsic component of a microreactor is that it results in a design that is transportable only by using the largest road conveyances and that the load would require special permits for transit through the U.S. highway system. New directions for further improvements and possibly more effective solutions to the shielding problem of a mobile microreactor are discussed in the conclusion section.

Microreactor↗
DOE OSTI2023

Mitigation of Cosmic Rays-Induced Errors in Superconducting Quantum Processors

Environmental radioactivity and cosmic-rays have recently been identified as a source of decoherence in super-conducting quantum bits (qubits). In particular, the absorption of cosmic-ray muons and gamma rays emitted by naturally occurring radioactive isotopes in the qubit substrate leads to correlated errors in superconducting quantum processors, posing significant challenges to quantum error correction. To enable quantum computing to scale, it is therefore necessary the devel-opment of mitigation strategies to prevent, or keep under control, error bursts due to particle impacts in the chip. While most environmental radioactive sources can be effectively suppressed using dedicated shielding, cosmic-ray muons, with their high penetration capability, can only be mitigated by moving the entire facility in a deep underground laboratory. This work explores the potential for developing a novel class of quantum processors equipped with an active veto system to protect superconducting-based quantum computers from the detrimental effects of atmospheric muons. Such a device would enable the identification of an atmospheric muon interaction within the processor and veto all operations performed during the occurrence of such an interaction. By demonstrating high detection efficiency and negligible dead time, we aim to establish that the future of quantum processors can be envisioned in above-around facilities.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2020

Electromagnetic Shielding Design for 200 kW Stationary Wireless Charging of Light-Duty EV

Wireless power transfer (WPT) is a developing technology with the advantage of convenience and flexible charging. SAE recommended practice J2954 defines typical size and geometry with aluminum or ferrite plate shielding to limit leakage electromagnetic (EM) fields for WPT with power levels lower than 22 kVA from the input side. However, as the WPT power goes up to 100- or 200-kW level, EM safety surrounding the WPT becomes a critical concern. To address this oncoming safety challenge, a novel ferrite shielding design is proposed in this paper. Different misalignment scenarios in accordance with definitions in SAE J2954 are also taken into consideration to ensure EM safety under various operation scenarios. Simulation results, which are preliminarily verified by magnetic field measurements at 1.1 m from the center of the vehicle side coil under 100 kW operation, indicate that the magnetic field leakage can be maintained below the limits defined in SAE J2954 for 200 kW operation. A 3.3 kW scale-down test was also carried out and compared to parallel scale-down simulations. 26.8% field emission reduction is observed from the scale-down test, which supports the effectiveness of the proposed shielding design.

Zhang, Bo↗
DOE OSTI2020

Electromagnetic Shielding Design for 200 kW Stationary Wireless Charging of Light-Duty EV

Wireless power transfer (WPT) is a developing technology with the advantage of convenience and flexible charging. SAE recommended practice J2954 defines typical size and geometry with aluminum or ferrite plate shielding to limit leakage electromagnetic (EM) fields for WPT with power levels lower than 22 kVA from the input side. However, as the WPT power goes up to 100- or 200-kW level, EM safety surrounding the WPT becomes a critical concern. To address this oncoming safety challenge, a novel ferrite shielding design is proposed in this paper. Different misalignment scenarios in accordance with definitions in SAE J2954 are also taken into consideration to ensure EM safety under various operation scenarios. Simulation results, which are preliminarily verified by magnetic field measurements at 1.1 m from the center of the vehicle side coil under 100 kW operation, indicate that the magnetic field leakage can be maintained below the limits defined in SAE J2954 for 200 kW operation. A 3.3 kW scale-down test was also carried out and compared to parallel scale-down simulations. 26.8% field emission reduction is observed from the scale-down test, which supports the effectiveness of the proposed shielding design.

25 ENERGY STORAGE↗
DOE OSTI2026

A Unified Theory for the Global Thunderstorm Distribution and Land–Sea Contrast

This article evaluates Entraining CAPE (ECAPE) as a thunderstorm proxy in climate studies using Global Precipitation Measurement satellite observations. ECAPE modifies traditional CAPE to account for the dependence of entrainment on the vertical wind shear, the lifted condensation level (LCL) height, and the properties of a cloud's surrounding atmosphere. ECAPE shows stronger pattern correlations with global regions of intense thunderstorms than previous metrics for updraft speed. In these regions, large CAPE, large shear, and high LCLs conspire to produce wide updrafts that are shielded from the negative effects of dry-air entrainment. ECAPE more skillfully discriminates intense thunderstorms from their less intense counterparts than other metrics commonly used in climatology and climate change studies of thunderstorms. We provide evidence that the well-known land-sea contrast in thunderstorm intensity is a consequence of larger CAPE and higher LCL heights over land than over the ocean.

entraining CAPE↗
DOE OSTI2023

The Association of Radiation Exposure with Stable Chromosome Aberrations in Atomic Bomb Survivors Based on DS02R1 Dosimetry and FISH Methods

The frequency of stable chromosome aberrations (sCA) in lymphocytes is a recognized radiation biological dosimeter. Its analysis can provide insights into factors that affect individual susceptibility as well as into the adequacy of radiation dose estimates used in studies of atomic bomb survivors. We analyzed the relationship between atomic bomb radiation exposure using the most recent DS02R1 dose estimates and the frequency of sCA as determined by FISH in 1,868 atomic bomb survivors. We investigated factors that may affect the background sCA rate and the shape and magnitude of the dose response. As in previous analyses of sCA in atomic bomb survivors that were based on Giemsa staining methods and used older DS86 dose estimates, the relationship between radiation dose and sCA rate was significant (P < 0.0001) with a linear-quadratic relationship at lower doses that did not persist at higher doses. As before, age at the time of the bombing and type of radiation shielding were significant dose-effect modifiers (P < 0.0001), but in contrast the difference in dose response by city was not so pronounced (P = 0.026) with a city effect not evident at doses below 1.25Gy. Background sCA rate increased with age at the time of examination (P < 0.0001), but neither sex, city, nor smoking was significantly associated with background rate. Based on FISH methods and recent dosimetry, the relationship between radiation dose and sCA frequency is largely consistent with previous findings, although the lesser importance of city as an effect modifier may reflect better dosimetry as well as more reproducible scoring of sCA. The persisting difference in sCA dose response by shielding category points to remaining problems with the accuracy or precision of radiation dose estimates in some A-bomb survivors.

62 RADIOLOGY AND NUCLEAR MEDICINE↗
DOE OSTI2024

Radiation Modeling of Z1: Validation of a Novel Radioisotope System

This paper presents the development of a radiation model for a novel strontium-90 (Sr-90) radioisotope heat source developed by Zeno Power Systems (Zeno), which demonstrates a groundbreaking fuel and shielding design that significantly reduces weight compared to traditional concepts. A Monte Carlo N-Particle (MCNP) model has been created to assess the effectiveness of this fuel and shielding design, however validation of the MCNP model is desired. Zeno has developed a prototype device (Z1) to aid in this model validation effort. This project is a collaborative effort between Zeno, the University of Dayton Research Institute (UDRI) and Pacific Northwest National Laboratory (PNNL), where the Z1 prototype was constructed and characterized.

RTG↗