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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 145 records · Page 8
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Materials property changes in ETU-10 graphite due to neutron irradiation at elevated temperatures

Graphite grade ETU-10, from IBIDEN Co., Ltd. Has been irradiated in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). The irradiation program was developed to provide a preliminary study the irradiation-induced property changes to the dimensions/volume, elastic properties, strength, electrical resistivity, coefficient of thermal expansion, and thermal diffusivity/conductivity over a range of temperatures and neutron exposures that may be relevant for future nuclear reactors. The irradiation envelope covers a range of irradiation temperatures (300°C–900 °C) and fluences (up to 40 × 10 25 n/m 2 [E > 0.1 MeV] or ~30 dpa) that would be relevant for advanced nuclear reactors. Further, the dimensional change was observed to be anisotropic for an isotropic graphite, the specimen dimensions, volume, Young's modulus, shear modulus, and strength all displayed a parabolic fluence dependence, the electrical resistivity had a rapid rise followed by a decrease and a later increase, at high fluence the mean coefficient of thermal expansions was similar for all irradiation temperatures, and thermal conductivity rapidly decreased followed by a continued loss.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Investigation of ASIC-based signal readout electronics for LEGEND-1000

LEGEND, the Large Enriched Germanium Experiment for Neutrinoless ββ Decay, is a ton-scale experimental program to search for neutrinoless double beta (0νββ) decay in the isotope 76 Ge with an unprecedented sensitivity. Building on the success of the low-background 76 Ge-based GERDA and Majorana Demonstrator experiments, the LEGEND collaboration is targeting a signal discovery sensitivity beyond 10 28 yr on the decay half-life with approximately 10 t∙yr of exposure. Signal readout electronics in close proximity to the detectors plays a major role in maximizing the experiment's discovery sensitivity by reducing electronic noise and improving pulse shape analysis capabilities for the rejection of backgrounds. However, the proximity also poses unique challenges for the radiopurity of the electronics. Application-specific integrated circuit (ASIC) technology allows the implementation of the entire charge sensitive amplifier (CSA) into a single low-mass chip while improving the electronic noise and reducing the power consumption. In this work, we investigated the properties and electronic performance of a commercially available ASIC CSA, the XGLab CUBE preamplifier, together with a p-type point contact high-purity germanium detector. We show that low noise levels and excellent energy resolutions can be obtained with this readout. Moreover, we demonstrate the viability of pulse shape discrimination techniques for reducing background events.

47 OTHER INSTRUMENTATION↗
DOE OSTI2017

A Summary of Actinide Enrichment Technologies and Capability Gaps

The evaluation performed in this study indicates that a new program is needed to efficiently provide a national actinide radioisotope enrichment capability to produce milligram-to-gram quantities of unique materials for user communities. This program should leverage past actinide enrichment, the recent advances in stable isotope enrichment, and assessments of the future requirements to cost effectively develop this capability while establishing an experience base for a new generation of researchers in this vital area. Preliminary evaluations indicate that an electromagnetic isotope separation (EMIS) device would have the capability to meet the future needs of the user community for enriched actinides. The EMIS technology could be potentially coupled with other enrichment technologies, such as irradiation, as pre-enrichment and/or post-enrichment systems to increase the throughput, reduce losses of material, and/or reduce operational costs of the base EMIS system. Past actinide enrichment experience and advances in the EMIS technology applied in stable isotope separations should be leveraged with this new evaluation information to assist in the establishment of a domestic actinide radioisotope enrichment capability.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2019

Initial Assessment for a Generic Horizontal Fissile Solution Vessel with an Electron Beam-based Neutron Generator

The LANL NA-23 program office has provided an extensive design study for a medical isotope (i.e. Molybdenum-99) production facility based on a fissile solution vessel with various accelerator technologies. These concept designs and associated R&D have provided affordable pathways for system optimization for industry counter partner such as SHINE and Global Medial Isotope Supply (GMIS). A horizontal fissile solution vessel design with an electron beam (E-beam) accelerator is proposed and studied with modeling capabilities developed for previous concepts (i.e. a generic vertical fissile solution vessel with a deuterium-tritium (DT) accelerator). An initial system assessment is performed by estimating theoretical Molybdenum-99 production yield with MCNP6.2. A baseline system configuration is constructed based on early design discussions between LANL and GMIS. Since only an initial design concept is available at this stage, all associated system design parameters can be adjusted to fit the best system performance characteristics. In this work, unlike previous vertical system studies, the cooling configuration of the horizontal fissile solution vessel is not considered, and only fission reactions are used to quantify the theoretical product yield. Five parametric design studies are proposed, preliminary results are briefly reported, and extended design studies beyond what is proposed here are also discussed. It is concluded that a comprehensive system analysis via multiphysics coupled calculation developed by LANL should be further conducted for the final design.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Effects of Experimental Warming and Elevated CO 2 on CO 2 and CH 4 Fluxes in an Ombrotrophic Black-Spruce. Final Report

High latitude peatlands and permafrost soils contain nearly half of the soil carbon pool on Earth. The potential for large soil C stocks in peatlands and permafrost to become a positive feedback to climate change by increasing net emissions of CO₂ and CH₄ to the atmosphere is a subject of grave concern. This research was conducted at the Spruce and Peatland Responses Under Changing Environments (herein SPRUCE) experimental facility, located in the USDA’s Marcell Experimental Forest in northern Minnesota. The site is dominated by Picea mariana [herein black spruce], ericaceous shrubs and Sphagnum spp. Ten 12m-diameter open top chambers expose portions of the 8.1-ha S1 bog to warming up to +9°C in 2.25°C increments at ambient and elevated CO₂ (eCO₂, 900ppm). The Finzi lab group began manual measurements of CO₂ and CH₄ fluxes and their C-isotopic compositions in 2014. With funding from the DOE-TES program automated measurements of these species began in 2015. With a one-year no cost extension we were able to make measurements through the 2018 growing season. Across the study period we find that experimental warming of the peatland significantly increases the flux of CO₂ and CH₄ to the atmosphere. The highest temperature treatments had the highest fluxes. As peatlands have complex topography and water table dynamics, many of these results were contingent upon the topographic location in which the measurements were made. In the raised portions of the peatland divorced from the water table, fluxes of CO₂ were high and fluxes of CH₄ were low. By contrast, in the lower portions of the peatland surface, which is at or very near the water table, fluxes of CH₄ were substantially higher. A significant fraction of the total annual CH₄ flux was derived from episodic ebullitive fluxes. The occurrence and size of these ebullitive fluxes also increased with temperature. Experimental fumigation of the black spruce peatland with elevated concentrations of atmospheric CO₂ [herein eCO₂] also had significant effects on emissions of greenhouse gasses. In general these effects were far more subtle than the effects of rising temperature. This indicates that temperature change more than atmospheric chemistry changes are affect the belowground cycle of C in the peatland. Overall, this study shows that boreal peatlands are highly sensitive to changes in temperature. The warmer it becomes the more C will be lost from the peatland to the atmosphere. This appears to create a positive feedback loop that is likely to enhance the concentration of greenhouse gases in the atmosphere.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2016

Poster - 40: Treatment Verification of a 3D-printed Eye Phantom for Proton Therapy

Purpose: Ocular melanoma is a form of eye cancer which is often treated using proton therapy. The benefit of the steep proton dose gradient can only be leveraged for accurate patient eye alignment. A treatment-planning program was written to plan on a 3D-printed anatomical eye-phantom, which was then irradiated to demonstrate the feasibility of verifying in vivo dosimetry for proton therapy using PET imaging. Methods: A 3D CAD eye model with critical organs was designed and voxelized into the Monte-Carlo transport code FLUKA. Proton dose and PET isotope production were simulated for a treatment plan of a test tumour, generated by a 2D treatment-planning program developed using NumPy and proton range tables. Next, a plastic eye-phantom was 3D-printed from the CAD model, irradiated at the TRIUMF Proton Therapy facility, and imaged using a PET scanner. Results: The treatment-planning program prediction of the range setting and modulator wheel was verified in FLUKA to treat the tumour with at least 90% dose coverage for both tissue and plastic. An axial isotope distribution of the PET isotopes was simulated in FLUKA and converted to PET scan counts. Meanwhile, the 3D-printed eye-phantom successfully yielded a PET signal. Conclusions: The 2D treatment-planning program can predict required parameters to sufficiently treat an eye tumour, which was experimentally verified using commercial 3D-printing hardware to manufacture eye-phantoms. Comparison between the simulated and measured PET isotope distribution could provide a more realistic test of eye alignment, and a variation of the method using radiographic film is being developed.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2023

Isotope effect in formaldehyde steam reforming on Pt/m-ZrO 2 : Insight into chemical promotion by alkalis

Infrared spectroscopy, temperature programmed reaction mass spectrometry (TP-reaction/MS), and catalyst testing were used to investigate alkali promotion of dehydrogenation selectivity during formaldehyde steam reforming (FSR) on Pt/m-ZrO 2 . In a preferred pathway, formaldehyde reacts with water forming hydrogen and formate, followed by forward formate decomposition to CO 2 and H 2 . Alkali-doping of 2 wt% Pt/m-ZrO 2 increases catalyst basicity, which weakens the formate C-H bond promoting formate dehydrogenation / decarboxylation. Promotion by alkali in FSR was observed through a formate nu(CH) band shift to lower wavenumbers in infrared spectroscopy, and through a decrease in the normal isotope effect in switching from H-to D-labeled formalde-hyde in TP-reaction/MS.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2020

In situ liquid SIMS analysis of uranium oxide

Trace analysis of nuclear materials in solid particles collected in the environment or particles in liquid slurry generated in nuclear material manufacturing processes can pinpoint elemental, organic, and isotopic signatures of nuclear fuel cycle activities and processes. Such information can support nuclear safeguards programs by increasing our ability to detect undeclared nuclear materials, routine activities for safeguarding at declared facilities, and illicit activities. However, trace radioactive material analysis in liquids and slurries is challenging using bulk approaches. For example, one drawback of sensitive analysis such as inductively coupled plasma mass spectrometry (ICP-MS) is that sample is consumed or destroyed as a result of the technical approach. We developed a vacuum compatible microfluidic interface to enable surface analysis of liquids and liquid-solid interactions using time-of-flight secondary ion mass spectrometry (ToF-SIMS). In this work, we illustrate initial results of liquid standards containing uranium using in situ liquid SIMS. Specifically a series of diluted uranium standards used for ICPMS analysis were studied and the limit of detection was estimated. Since the liquid SIMS analysis is almost nondestructive, the same sample can be analyzed by other more destructive approaches afterwards or saved for future references. Consequently, multimodal analysis is possible. Our results demonstrate that in situ liquid SIMS can be used as a new approach to analyze radioactive materials in liquid and slurry forms of relevance to the IAEA safeguards missions.

In situ liquid ToF-SIMS, uranium, limit of detecti↗
DOE OSTI2023

From Count Rates to Quantifying Isotopic Activities – Field Analysis of Radiation Monitoring Data

The Nevada National Security Site (NNSS) provides a comprehensive bicoastal radiological and nuclear emergency response to United States Department of Energy/National Nuclear Security Administration. A major part of the support is to provide systematic radiological search for lost or stolen sources, Radiological search is a core competency of the NNSS with its origin dating back to nuclear weapons test era. Search operations from multiple platforms is the common thread among the various NNSS assets, which include Aerial Measuring System (AMS), Maritime Support Team (MST), National Capitol Response (NCR), National Search Team (NST) and Radiological Assistance Program (RAP). Information collected and analyzed during search operations add to the actionable intelligence for the law enforcement agencies and provide valuable guidance for the tactical resolution of a nuclear or radiological crisis. Search is an intelligence and situational awareness driven operation and most often called upon during a radiological emergency, however it can be brought into play to thwart a potential threat by providing monitoring and surveillance support. The Office of Nuclear Incident Response (NA-84) serves as the technical leader in responding to and resolving nuclear and radiological threats worldwide and integrates its efforts with other NNSA stakeholders (e.g., NNSA office of Defense Nuclear Non-proliferation NA-22). The response includes expertise in the areas of radiological search, render safe, and consequence management. This article will discuss the methodologies, tools, procedures, and techniques to extract maximum radiological characterization information (isotopic composition, activities for individual isotopes, threat assessment etc.) from field monitoring or Search operation data.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2023

Ground Water Age Predictor

Machine Learning script to predict groundwater ages based on auxiliary features in a publicly available dataset, based on publicly available software libraries. Code applied to data from the Groundwater Ambient Monitoring and Assessment (GAMA) program in California, including well location and construction information, chemical constituents and isotopic tracers, and land use metrics.

Chakraborty, Indrasis↗
DOE OSTI2023

Report on the Analysis of Plutonium-238 Oxide, Campaign #P5PO

Oak Ridge National Laboratory (ORNL) is currently producing Plutonium-238 oxide for NASA space programs. Through the irradiation of NpO 2 pellets in the High Flux Isotope Reactor (HFIR), the production of Pu-238 oxide will be used for electrical power and heat on NASA spacecraft such as the Curiosity Mars Rover and the Voyager 1 and 2. The Transuranium Analytical Laboratory (TAL) leads the analytical effort for this project, which includes the dissolution of plutonium product oxide (PUP) and the analysis of the dissolved product solution using both radiological and inductively coupled plasma mass spectrometry (ICP-MS) techniques that are required to meet product specifications. The Transuranic Analytical Laboratory received three samples of Pu-238 oxide for analysis in January of 2022. This report documents the analytical protocols performed by the TAL along with the results produced from said analytical protocols. This report also includes a comparison with the general-purpose heat source (GPHS) specifications.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2016

STATUS OF TRISO FUEL IRRADIATIONS IN THE ADVANCED TEST REACTOR SUPPORTING HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGNS

The United States Department of Energy’s Advanced Reactor Technologies (ART) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is irradiating up to seven low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The experiments will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of several independent capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2) started irradiation in June 2010 and completed in October 2013. The third and fourth experiments have been combined into a single experiment designated (AGR-3/4), which started its irradiation in December 2011 and completed in April 2014. Since the purpose of this experiment was to provide data on fission product migration and retention in the NGNP reactor, the design of this experiment was significantly different from the first two experiments, though the control and monitoring systems are very similar. The final experiment, AGR-5/6/7, is scheduled to begin irradiation in early summer 2017.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Combining RHF and HFIR Disposition Campaigns – Analysis, Opportunity, and Lessons Learned

Savannah River Nuclear Solutions has tasked its H-Canyon facility to prepare for disposition of used Oak Ridge National Laboratory High Flux Isotope Reactor (HFIR) cores. HFIR cores consist of an inner and outer assembly of highly enriched uranium fuel plates. The two elements are annular, nest inside each other, and the fuel form is a curved plate held between two collars. H-Canyon has a dissolver insert specifically designed for HFIR cores and has performed this type of campaign before between 1976 and 1988. At the H-Canyon facility, a dissolver is a large stainless steel pot loaded with a nitric acid solution into which uranium-alloy fuel is dissolved. The dissolver uses a device called an insert to feed the fuels to the acid solution in a controlled manner. The mission will be to dissolve a number of HFIR cores, purify the uranium, and down blend it to low enriched uranium suitable for commercial power reactors.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Processing and Disposition of Special Actinide Target Materials - 17071

The DOE manages an inventory of materials that contain a range of long-lived radioisotopes that were produced from the 1960's through the 1980's by irradiating targets in high-flux reactors at the SRS to produce special heavy isotopes for DOE programmatic use, scientific research, and industrial and medical applications. Among the products were Cf-249, -250, -251, and -252; heavy curium (=50% Cm-246); and Pu-242 and -244. The irradiation of the Mark-18A (Mk-18A) targets at extremely high fluxes produced the world's supply of heavy curium and Pu-244. Approximately one fourth of the Mk-18A targets were processed in the 1970's to recover and enrich the material that has been used as the global source of Pu-244 and heavy curium by industry and researchers. These isotopes are still in demand, and the available supply of purified material is dwindling. A program has been initiated to recover heavy curium and Pu-244 from the remaining stored unprocessed Mk-18A targets. Improved flowsheets to process the materials to recover Pu-244 and heavy curium are being developed. Evaluations are also underway to determine whether additional materials present in the targets in much smaller quantities, such as long-lived fission product (LLFP) isotopes, can be recovered for research and development applications. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

AFRRI TRIGA Reactor Neutron and Gamma Dose Characterization Preliminary Results

Integral Experiment Request (IER) 484 is part of a series of dose characterization and nuclear accident dosimeter (NAD) exercises performed under the Department of Energy (DOE) Nuclear Criticality Safety Program (NCSP). The Armed Forces Radiobilogy Research Institute (AFRRI) 1.1 MW Training, Research, Isotopes, General Atomics (TRIGA) Mark-F reactor is the third facility to undergo a dose study to establish reference dose values. A preliminary measurement was performed earlier this year, where a variety of foils were irradiated and activation gamma rays were measured using a high-purity germanium (HPGe) detector. The data was taken and analyzed to calculate the neutron spectrum in the irradiation facility. This paper will present preliminary results of the neutron dose measured. The characterization of the AFRRI dose field is planned in August.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2017

FY16 Progress in Resonance Evaluations of Gadolinium for the NCSP

A reevaluation of the resonance regions of the five most abundant isotopes of gadolinium is under way. The task of the reevaluation is being done as a collaboration between the nuclear data group at Oak Ridge National Laboratory (ORNL) and at the Institute for Radiological Protection and Nuclear Safety (IRSN) under a memorandum of understanding between the US Department of Energy (DOE) Nuclear Criticality Safety Program (NCSP) and IRSN.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2023

Molybdenum-99 from Molten Salt Reactor as a Source of Technetium-99m for Nuclear Medicine: Past, Current, and Future of Molybdenum-99

Technitium-99m ( 99m Tc), a widely used radioisotope, is used in tens of millions of medical diagnostic procedures annually. However, it is hard to store and must be immediately used upon production due to its short half-life (i.e., 6 h); thus, it is currently produced from 99 Mo, which itself is a result of 235 U fission. The majority of 99 Mo supplies to U.S. patients are currently provided by foreign producers and produced using highly enriched uranium (HEU). In order to minimize the proliferation risks of HEU-based medical isotope production, the U.S. Department of Energy’s National Nuclear Security Administration has funded a program to accelerate the development of technologies to produce 99 Mo without the use of HEU. Today, the global supply of 99 Mo depends on a limited number of nuclear reactors, and production has been interrupted unexpectedly since 2009 due to the fleet’s advanced age. Herein, alternative options for 99 Mo production are discussed, and one potential option is to obtain 99m Tc from molten salt reactors (MSRs). A MSR is a nuclear fission reactor that can operate at or close to atmospheric pressure with liquid fuel, which allows for producing isotopes in a timely manner. In this paper, the past and current production of 99 Mo via nuclear reactors is described, and the future of 99 Mo production by MSRs is discussed. The behavior and chemical properties of molybdenum in fluoride salts in MSRs and the possible extraction methods are also examined in addition to the limitation of current studies.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗