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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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All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Developing Multiphysics, Integrated, High-Fidelity, Massively Parallel Computational Capabilities for Fusion Applications Using MOOSE

As the need for fusion as a clean, sustainable, and abundant energy source grows internationally, so does the need for multiphysics, computational tools to model, study, and predict the complex interactions between plasma, materials, and engineering processes. These tools have a crucial role to play in solving scientific and engineering challenges and accelerating fusion energy deployment. To address these needs, modeling capabilities should enable massively parallel, multiphysics, fully integrated high-fidelity simulations of fusion systems. Additional attributes, such as being open source and modular while maintaining high software quality assurance standards will maximize impact by ensuring accessibility for all and wide acceptance, rapid expansion and development, as well as reliability, efficiency, and robustness. In this paper, we describe how the Multiphysics Object-Oriented Simulation Environment (MOOSE) framework, which has a track record of success in the fission space thanks to the attributes listed above, can be leveraged in the fusion energy field. We highlight key successes of the MOOSE application in the fission space and describe how MOOSE has been and is being applied to fusion applications in the United States---e.g., Tritium Migration Analysis Program, version 8 (TMAP8), MOOSE Fusion Module, Fusion ENergy Integrated multiphys-X (FENIX)---and the United Kingdom---e.g., AURORA, Achlys, Apollo. These efforts aim to establish a suite of tools that can be further extended to accelerate fusion energy deployment.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2026

Evaluating Lithium-Lead Mixtures for Increased Tritium Breeding in Fusion Energy Blanket Systems: Exploring the Impact of Composition and Temperature

This project investigates hypo- and hyper-eutectic liquid lithium-lead (PbLi) mixtures as a coolant in fusion blanket systems, with a focus on testing the compatibility of compositions with better tritium breeding ratios (TBR). The 12-month study concentrates on static corrosion experiments, comparing silicon carbide (SiC) corrosion rates in Li-rich and eutectic Li-Pb mixtures, and examining interactions between Mo and PbLi. The research seeks to determine the maximum operating temperatures for this material combination, acknowledging the balance between Li-Pb melting temperature and achievable TBR to unlock new opportunities for blanket design. This work lays the groundwork for future collaboration between Kyoto Fusioneering and Oak Ridge National Laboratory (ORNL). Findings could enhance the TBR of fusion blanket designs, and impact fusion energy development by confirming material compatibility between different Li-Pb mixtures with SiC and Mo at higher temperatures, including at temperatures relevant for very high-temperature blankets (1,000 °C), which could enable higher electricity conversion efficiencies and commercial applications using process heat.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2015

Passive neutron coincidence counting with plastic scintillators for the characterization of radioactive waste drums

The quantification of radioactive material is essential in the fields of safeguards, criticality control of nuclear processes, dismantling of nuclear facilities and components, or radioactive waste characterization. The Nuclear Measurement Laboratory (LMN) of CEA is involved in the development of time-correlated neutron detection techniques using plastic scintillators. Usually, 3He proportional counters are used for passive neutron coincidence counting owing to their high thermal neutron capture efficiency and gamma insensitivity. However, the global {sup 3}He shortage in the past few years has made these detectors extremely expensive. In addition, contrary to {sup 3}He counters for which a few tens of microseconds are needed to thermalize fast neutrons, in view to maximize the {sup 3}He(n,p){sup 3}H capture cross section, plastic scintillators are based on elastic scattering and therefore the light signal is formed within a few nanoseconds, correlated pulses being detected within a few dozen- or hundred nanoseconds. This time span reflects fission particles time of flight, which allows reducing accordingly the duration of the coincidence gate and thus the rate of random coincidences, which may totally blind fission coincidences when using {sup 3}He counters in case of a high (α,n) reaction rate. However, plastic scintillators are very sensitive to gamma rays, requiring the use of a thick metallic shield to reduce the corresponding background. Cross talk between detectors is also a major issue, which consists on the detection of one particle by several detectors due to elastic or inelastic scattering, leading to true but undesired coincidences. Data analysis algorithms are tested to minimize cross-talk in simultaneously activated detectors. The distinction between useful fission coincidences and the correlated background due to cross-talk, (α,n) and induced (n,2n) or (n,n'γ) reactions, is achieved by measuring 3-fold coincidences. The performances of a passive neutron coincidence counting system for radioactive waste drums using plastic scintillators have been studied using the Monte Carlo radiation transport code MCNPX-PoliMi v2.0 coupled to data processing algorithms developed with ROOT data analysis software. In addition to the correlated background, accidental coincidences are taken into account in the simulation by randomly merging pulses from different calculations with fission and (α,n) sources. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2015

Separation of tritiated water using graphene oxide membrane

In future nuclear fuel reprocessing plants and possibly for nuclear power plants, the cleanup of tritiated water will be needed for hundreds of thousands of gallons of water with low activities of tritium. This cleanup concept utilizes graphene oxide laminar membranes (GOx) for the separation of low-concentration (10 -3 -10 µCi/g) tritiated water to create water that can be released to the environment and a much smaller waste stream with higher tritium concentrations. Graphene oxide membranes consist of hierarchically stacked, overlapping molecular layers and represent a new class of materials. A permeation rate test was performed with a 2-µm-thick cast Asbury membrane using mixed gas permeability testing with zero air (highly purified atmosphere) and with air humidified with either H 2 O or D 2 O to a nominal 50% relative humidity. The membrane permeability for both H 2 O and D 2 O was high with N 2 and O 2 at the system measurement limit. The membrane water permeation rate was compared to a Nafion® membrane and the GOx permeation was approximately twice as high at room temperature. The H 2 O vapor permeation rate was 5.9 × 10 2 cc/m 2 /min (1.2 × 10 -6 g/min-cm 2 ), which is typical for graphene oxide membranes. To demonstrate the feasibility of such isotopic water separation through GOX laminar membranes, an experimental setup was constructed to use pressure-driven separation by heating the isotopic water mixture at one side of the membrane to create steam while cooling the other side. Several membranes were tested and were prepared using different starting materials and by different pretreatment methods. The average separation result was 0.8 for deuterium and 0.6 for tritium. Higher or lower temperatures may also improve separation efficiency but neither has been tested yet. A rough estimate of cost compared to current technology was also included as an indication of potential viability of the process. The relative process costs were based on the rough size of facility to accommodate the large surface area of the membranes and the energy needed to evaporate the water and pass through the membranes as compared to the currently used combined electrolysis and catalytic exchange process.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Tritium Distribution and Capture Requirements for UNF Reprocessing

Four radionuclides have been identified as being sufficiently volatile in reprocessing of nuclear fuel that their gaseous release needs to be controlled to meet U.S. regulatory requirements. These radionuclides are {sup 3}H, {sup 14}C, {sup 85}Kr, and {sup 129}I. Previous studies have discussed the need for treating off-gas streams to control {sup 3}H, {sup 14}C, {sup 85}Kr, and {sup 129}I from the reprocessing of used nuclear fuel (UNF) and the regulations that set the limits for releases from such a facility. These regulations ultimately prescribe the decontamination factor (DF) that must be met in the facility. Jubin et al. examined the role of off-gas pathways since the gaseous radionuclides can partition to different reprocessing off-gas streams. Based on this report and subsequent analysis, the required DF for {sup 129}I could be 1000 to 3000. The required DF for {sup 3}H could be as high as 200, depending on the age of the fuel processed. The DF for {sup 85}Kr could be up to ∼35, also depending on fuel age. The required DF for {sup 14}C is 1 in many cases (no treatment required) but could be higher in specific cases. While iodine has been the focus of a number of studies due to the high DF requirements on multiple escape pathways, it is clear from the studies noted above that the control of tritium will also be a significant issue for future reprocessing plants, especially if processing relatively short-cooled fuels (<10 yr post reactor discharge). Tritium contained in a UNF matrix will be released into the liquid and off-gas streams during fuel reprocessing. The amount released into the process streams is a function of the type and burnup of the fuel processed, and the chemical processes employed within the reprocessing plant. Previous studies have examined the tritium capture requirements for the reprocessing facility assuming that all of the tritium is volatilized in the head-end operations, i.e., tritium pretreatment is employed and tritium is completely released from the cladding. This is clearly a worst case scenario from the standpoint of the amount of tritium that could be released, but the scenario tends to limit the release pathway to the head-end off-gas stream. In this analysis, the volume of HTO and the HTO to H{sub 2}O ratio for each of the three plant-wide tritium DFs were determined. The overall amount of HTO that had to be captured from the off-gas streams was virtually the same for the three DF cases evaluated; 98%, 99% and 99.5% of the tritium in the fuel entering the plant needed to be recovered for overall plant DFs of 50, 100 and 200, respectively. The use of the 'tritium barrier' to limit the spread of HTO from the head-end systems to the balance of the plant results in an increase in tritiated waste concentration in the head-end recycle acid systems which results in a decrease in the total amount of HTO + H{sub 2}O that must recovered from the gaseous effluent from the plant. The use of the 'tritium barrier' decreases the waste volume due to the higher HTO:H{sub 2}O ratio in the treated off-gas stream. The results also show that increasing the acid recycle percentage increases the concentration of HTO within the UNF reprocessing plant. This is then manifested in less HTO + H{sub 2}O that must be recovered from the gaseous effluents. Finally, the application of tritium recovery on the recycle acid reduces the HTO concentration in the recycle acid stream and generates a separate HTO + H{sub 2}O stream that must be managed. However, the resulting volume of tritium-containing waste generated plant-wide is a function of the selectivity of the tritium recovery system and the ability of the tritium recovery system to limit the accumulation of HTO in the recycle acid such that the HTO + H{sub 2}O recovered from gaseous effluents does not contribute significantly to the overall tritium waste volume. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Radiological mapping of facilities under dismantling process by digital autoradiography - 15230

because safety standards have changed. In that context, radioactive waste management is a challenging task faced by existing nuclear power countries. Therefore, characterization is essential to classify wastes and to reduce costs. In many facilities, after the removal of glove boxes, fume cupboards and so on, the remaining materials on the floor, walls, and ceiling must be characterized. Gamma emitters can be detected remotely by commercial devices. However, analytical techniques to detect alpha and beta emitters in situ are generally not well suited for large area investigations. The Digital Autoradiography (DA) technique, initially dedicated to research in biology, was re-routed for final state characterization before facility demolition. It is based on the exposure of radio-sensitive phosphor screens on the materials to characterize. Phosphor screens, after a scanning process, provide a picture of the radioactivity, and are reusable several tens of times. This analytical technique is sensitive to all types of radioactivity, requires neither operators nor electricity during acquisition process, and produces only very few wastes. DA has already been tested in different facilities to obtain semi-quantitative values of C-14 and H-3 activities. Large amount of data gathered to cover the whole surface to be characterized. A Geographic Information System (GIS) associated with geostatistical method has been used to provide Radiological Mapping. This method to interpret Digital Autoradiography results has been proved very efficient to prepare remediation project. (authors)

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2020

Solid Cherenkov detector for studying nucleosynthesis in inertial confinement fusion

Measuring gamma rays emitted from nuclear reactions gives insight into their nuclear structure. Notably, there are several nuclear reactions that produce gamma rays at ~1 MeV–3 MeV energies such as T( 4 He, γ) 7 Li, 4 He( 3 He, γ) 7 Be, and 12 C(p, γ) 13 N, which may solve questions lingering about big-bang nucleosynthesis and stellar nucleosynthesis. To observe 1 MeV–3 MeV gamma rays in an inertial confinement fusion system, a new style of the Cherenkov detector was developed using aerogel and fused silica as a Cherenkov medium. Utilizing the OMEGA laser facility, both aerogel and fused silica media were compared with the existing gas-medium Cherenkov detector to validate the concept. Furthermore, gamma ray measurements from high yield inertial confinement fusion implosions (deuterium–tritium and deuterium– 3 He) demonstrated that aerogel and fused silica were viable Cherenkov media, paving the way for a potential optimized detector to make these cross section measurements on OMEGA or the National Ignition Facility.

47 OTHER INSTRUMENTATION↗
DOE OSTI2013

Normal and outlying populations of the Milky Way stellar halo at [Fe/H] <–2

From detailed abundance analysis of >100 Hamburg/ESO candidate extremely metal-poor (EMP) stars we find 45 with [Fe/H] < –3.0 dex. We identify a heretofore unidentified group: Ca-deficient stars with sub-solar [Ca/Fe] ratios and the lowest neutron-capture abundances; the Ca-deficient group comprises ∼10% of the sample, excluding Carbon stars. Our radial velocity distribution shows that the carbon-enhanced stars with no s-process enhancements, CEMP-no, and which do not show C{sub 2} bands are not preferentially binary systems. Ignoring Carbon stars, approximately 15% of our sample are strong (≥5σ) outliers in one or more elements between Mg and Ni; this rises to ∼19% if very strong (≥10σ) outliers for Sr and Ba are included. Examples include: HE0305–0554 with the lowest [Ba/H] known; HE1012–1540 and HE2323–0256, two (non-velocity variable) C-rich stars with very strong [Mg,Al/Fe] enhancements; and HE1226–1149, an extremely r-process rich star.

79 ASTRONOMY AND ASTROPHYSICS↗
DOE OSTI2024

Bayesian batch optimization for molybdenum versus tungsten inertial confinement fusion double shell target design

Access to reliable, clean energy sources is a major concern for national security. Much research is focused on the “grand challenge” of producing energy via controlled fusion reactions in a laboratory setting. For fusion experiments, specifically inertial confinement fusion (ICF), to produce sufficient energy, the fusion reactions in the ICF fuel need to become self-sustaining and burn deuterium-tritium (DT) fuel efficiently. The recent record-breaking NIF ignition shot was able to achieve this goal as well as produce more energy than used to drive the experiment. This achievement brings self-sustaining fusion-based power systems closer than ever before, capable of providing humans with access to secure, renewable energy. In order to further progress toward the actualization of such power systems, more ICF experiments need to be conducted at large laser facilities such as the United States's National Ignition Facility (NIF) or France's Laser Mega-Joule. The high cost per shot and limited number of shots that are possible per year make it prohibitive to perform large numbers of experiments. As such, experimental design relies heavily on complex predictive physics simulations for high-fidelity “preshot” analysis. These multidimensional, multi-physics, high-fidelity simulations have to account for a variety of input parameters as well as modeling the extreme conditions (pressures and densities) present at ignition. Such simulations (especially in 3D) can become computationally prohibitive to turn around for each ICF experiment. In this work, we explore using Bayesian optimization with Gaussian processes (GPs) to find optimal designs for ICF double shell targets, while keeping computational costs to manageable levels. These double shell targets have an inner shell that grades from beryllium on the outer surface to the higher Z material molybdenum, as opposed to the nominally used tungsten, on the inside in order to trade off between the high performance associated with high density inner shells and capsule stability. We describe our results for “capsule-only” xRAGE simulations to study the physics between different capsule designs, inner shell materials, and potential for future experiments.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
NASA NTRS2018

Pulsed Fission-Fusion (PuFF)

In September 2013 the NASA Innovative Advanced Concept (NIAC) organization awarded a phase I contract to the PuFF team. Our phase 1 proposal discussed a pulsed fission-fusion propulsion system that injected gaseous deuterium (D) and tritium (T) as a mixture in a column, surrounded concentrically by gaseous uranium fluoride (UF6) and then an outer shell of liquid lithium. A high power current would flow down the liquid lithium and the resulting Lorentz force would compress the column by roughly a factor of 10. The compressed column would reach criticality and a combination of fission and fusion reactions would occur. The fission reactions would further energize the fusion center, and the fusion reactions would generate neutrons that promote more complete burnup of the fission fuel. The lithium liner provides some help as a neutron reflector but also acts as a propulsive medium, being converted to plasma which is then expanded against a magnetic nozzle for thrust. The expansion of the (primarily) lithium plasma against the nozzle's magnetic field inducts a current that is used to charge the system for the next pulse. Our concept also included secondary injection of a Field Reversed Configuration (FRC) plasmoid that would provide a secondary compression direction, axially against the column, and push the column away from the injection manifold, increasing the manifold's survivability.Our phase 1 proposal included modeling the above process first under steady state assumptions and second under a time variant integration. We proposed including these results into a Mars concept vehicle and finally proposing promising conditions to be evaluated experimentally in Phase II. In phase I we quickly realized that we needed to modify our approach. Our steady state work was completed as proposed, and the results indicated that one, a two stage compression system was not needed and two, that we wanted to move away from UF6. The steady state model shows much more margin than expected, to the point that we may well reach breakeven with the Charger – 1 facility, a 572 kJ Marx bank currently under refurbishment at UAH. Additionally we found that using gaseous D-T and UF6, provided a relatively simple prospect of using a pulsed injector, made reaching criticality more difficult. The introduction of large amounts of fluorine meant a radiative sink, sapping power from the fusion plasma and was harder to handle. Therefore we moved to a solid uranium target that held D-T under pressure. In so doing we could move our target closer to criticality and remove any material that did not sustain the reaction.

Fusion↗
NASA NTRS1994

High efficiency GaP power conversion for Betavoltaic applications

AstroPower is developing a gallium phosphide (GaP) based energy converter optimized for radio luminescent light-based power supplies. A 'two-step' or 'indirect' process is used where a phosphor is excited by radioactive decay products to produce light that is then converted to electricity by a photovoltaic energy converter. This indirect conversion of beta-radiation to electrical energy can be realized by applying recent developments in tritium based radio luminescent (RL) light sources in combination with the high conversion efficiencies that can be achieved under low illumination with low leakage, gallium phosphide based devices. This tritium to light approach is inherently safer than battery designs that incorporate high activity radionuclides because the beta particles emitted by tritium are of low average energy and are easily stopped by a thin layer of glass. GaP layers were grown by liquid phase epitaxy and p/n junction devices were fabricated and characterized for low light intensity power conversion. AstroPower has demonstrated the feasibility of the GaP based energy converter with the following key results: 23.54 percent conversion efficiency under 968 muW/sq cm 440 nm blue light, 14.59 percent conversion efficiency for 2.85 muW/sq cm 440 nm blue light, and fabrication of working 5 V array. We have also determined that at least 20 muW/sq cm optical power is available for betavoltaic power systems. Successful developments of this device is an enabling technology for low volume, safe, high voltage, milliwatt power supplies with service lifetimes in excess of 12 years.

Sims, Paul E.↗
DOE OSTI2017

Post-Closure Monitoring Report for Corrective Action Unit 98, Frenchman Flat, Underground Test Area, Nevada National Security Site, Nevada for Calendar Year 2016 (January 2016–December 2016), Revision 1

Corrective Action Unit (CAU) 98: Frenchman Flat on the Nevada National Security Site was the location of 10 underground nuclear tests. CAU 98 underwent a series of investigations and actions in accordance with the Federal Facility Agreement and Consent Order to assess contamination of groundwater by radionuclides from the tests. A Closure Report completed that process in 2016 and called for long-term monitoring, use restrictions (URs), and institutional controls to protect the public and environment from potential exposure to contaminated groundwater. Three types of monitoring are performed for CAU 98: water quality, water level, and institutional control. These are monitored to determine whether the UR boundaries remain protective of human health and the environment, and to ensure that the regulatory boundary objectives are being met. Monitoring data will be used in the future, once multiple years of data are available, to evaluate consistency with the groundwater flow and contaminant transport models because the contaminant boundaries calculated with the models are the primary basis of the UR boundaries. Six wells were sampled for water-quality monitoring in 2016. Contaminants of concern were detected only in the two source/plume wells already known to contain contamination as a result of a radionuclide migration experiment. The tritium concentration in both wells is slowly decreasing. The water-level monitoring network includes 16 wells. Depth to water measured in 2016 in all wells is consistent with recent measurements. Institutional control monitoring confirmed the URs are recorded in U.S. Department of Energy and U.S. Air Force land management systems, and that no activities within Frenchman Flat basin are occurring that could potentially affect the contaminant boundaries. Survey of groundwater resources in basins surrounding Frenchman Flat similarly identify no current or pending development that would indicate the need to increase monitoring activities or would otherwise cause concern for the closure decision. The URs continue to prevent exposure of the public, workers, and the environment to contaminants of concern by preventing use of potentially contaminated groundwater.

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

High Compression in Palladium Alloy Metal Foil Pumps

Metal foil pumps (MFPs) are key components in the direct internal recycling inner fuel cycle loop for the recovery of hydrogen isotopes from deuterium-tritium fusion exhaust. Operating under vacuum conditions, they utilize superthermal hydrogen as the feed gas in a process called superpermeation. A notable feature of MFPs is their ability to pump against a pressure gradient. This study examines the compression capabilities of PdAg and PdCu MFPs at low temperatures with a constant feed pressure of 10 Pa. At 75°C, compression ratios exceeding 200 were readily achieved, with downstream pressures exceeding 4500 Pa using PdCu. For both alloys, net fluxes decreased by only ~15% at downstream pressures of 1000 Pa, which offers potential simplifications for the downstream pump train. Performance declined markedly when the temperature was elevated to 200°C. Pump curves were constructed and advocated as the most appropriate manner to assess MFP performance. Separate pressure-driven-permeation experiments at relevant conditions were conducted, providing a direct measurement of the hydrogen dissociation constant $k$ d , which was found to be in good agreement with the previous literature. Further, these measurements were used to predict pump curves and maximum compression ratios by balancing superpermeation with pressure-driven permeation, achieving excellent agreement with experiment. Last, experiments using asymmetric MFPs revealed the detrimental impact that surface impurities have on performance in this system.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2018

Development of LWR fuels with enhanced accident tolerance. Final report

After a tsunami triggered plant black out at the Fukushima Daiichi site, followed by hydrogen explosions, the US Department of Energy partnered with fuel vendors to study safer alternatives to the current UO2 – zirconium alloy fuel system. This accident tolerant fuel (ATF) alternative should better tolerate a loss of cooling in the core for a considerably longer time while maintaining or improving the fuel performance during normal operation conditions. General Electric, Oak Ridge National Laboratory, and their partners are proposing to replace zirconium alloy cladding in current commercial light water power reactors with an iron-chromium-aluminum (FeCrAl) cladding such as APMT, C26M or FeCrAl ODS, which has been named IronClad by GE. Extensive testing and evaluation has been conducted to determine the suitability of FeCrAl for the entire fuel cycle from rod fabrication to used fuel disposition. Most of the environmental characterization was conducted to evaluate the behavior of IronClad under normal operation conditions and under severe accident conditions. Results show that FeCrAl has excellent corrosion resistance under normal operation conditions and FeCrAl is several orders of magnitude more resistant than zirconium alloys to degradation by superheated steam under accident conditions, generating less heat of oxidation and lower amount of combustible hydrogen gas. Higher parasitic neutron absorption and potential tritium release into the coolant effects can be minimized by design and operational measures. IronClad material can be manufactured using traditional melting and forging (C26M) or powder metallurgy processes (APMT & FeCrAl ODS). Tubes can be made to full length of 5 m with a wall thickness of 0.4 mm by pilgering or drawing. The fuel rods can be readily hermetically sealed in an industrial production setting by welding end caps to the thin wall tubing using either TIG or resistance welding. The implementation of FeCrAl IronClad cladding is a near term solution to enhance the safety of the current fleet of commercial light water power reactors.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

EVALUATION OF AN UPWARD TREND IN BACKGROUND COUNT RATES FROM A STACK PARTICULATE CONTINUOUS AIR MONITOR

The Pacific Northwest National Laboratory (PNNL) operates the Radiochemical Processing Laboratory (RPL), which is a multi-purpose, non-reactor nuclear research facility. Regulations require both continuous sampling and monitoring of radioactive particulates and tritium gas in the exhaust from the main stack. Releases of other radioactive gases, including planned releases of radon, are tracked separately in a database and reported. During the 2015 calibration of the RPL stack continuous air monitor (CAM), measured alpha and beta background count rates were much higher than expected, especially when compared to count rates from previous calibrations. The source of the higher background count rates was examined by trending of historical CAM measurements and a comparison to the sampler data. The analysis revealed that the sample results showed no increase in emissions, whereas the CAM showed a steady increase in count rates. Ultimately, the CAM filter was analyzed and found to contain higher-than-normal background levels of 220Rn progeny. Assessments were performed to determine the cause of the increased background values, including reviews of building research activities, radioactive material usage and storage, adequacy of procedures, and the potential for internal CAM contamination. Project reviews determined that a research activity involving 228Th was left in a semi-stored state, resulting in 220Rn being released from a hot cell into the exhaust system. The 228Th source material was subsequently repackaged and contained resulting in a decrease of CAM background count rates. An estimate of the 220Rn release was made and the contribution to the annual offsite dose from the facility was calculated. The released activity and reported dose results were well below the permit limits for the facility.

radon, 220Rn, environmental assessment, air monito↗
DOE OSTI2017

Terrace Geochemistry at the Shiprock, New Mexico, Disposal Site - WM2017-17232 Initial Phase

The objective of this investigation was to identify and differentiate potential non- mill-related water inputs to a shallow terrace groundwater system through the use of aqueous chemical and isotopic tracers at a former uranium- and vanadium-ore processing facility. Terrace groundwater in the vicinity of the Shiprock, New Mexico, site is hypothesized to be largely anthropogenic because natural rates of recharge in the terrace are likely insufficient to sustain a continuous water table in the terrace alluvial system, as observed in several analogue terrace locations east of the site and in response to post-mill dewatering efforts across the site. The terrace is composed of alluvial sand and gravel and weathered and unweathered Mancos Shale. Terrace groundwater exists and flows in the alluvium and to a much less extent in the Mancos Shale. Historical data established that in both the terrace and floodplain below the terrace, mill-derived uranium and sulfate is found primarily in the alluvium and the upper portion of the weathered Mancos Shale. Groundwater extraction is being conducted in the vicinity of former mill operations and in washes and seeps to dewater the formation and remove contamination, thus eliminating these exposure pathways and minimizing movement to the floodplain. However, past and present contribution of non-mill anthropogenic water sources may be hindering the dewatering effort, resulting in reduced remedy effectiveness. Groundwater source signatures can be determined based on chemical and isotopic ratios and are used to help identify and delineate both mill and non-mill water contributions. Aqueous chemical and isotopic tracers, such as 234U/238U activity ratios and uranium concentrations, δ34S sulfate and sulfate concentrations, tritium concentrations, and δ2Hwater and δ18O water are being used in this Phase I study. The aqueous chemical and isotopic analysis has identified areas on the terrace where groundwater is derived from mill-related activities and areas where the groundwater is associated with non-mill activities. A separate field effort of Phase II work will follow, including investigating additional locations for these isotopes and examination of δ18Osulfate , δ34Ssulfate , and chlorofluorocarbon signatures.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2021

Fusion gain from cylindrical liner-driven implosions of field reversed configurations

MagLIF experiments [M.R. Gomez et al., Phys. Plasmas 22, 056306 (2015)] on Z have demonstrated the basic principles of Magneto-Inertial Fusion (MIF) for wall confined plasmas. Other MIF schemes have been proposed based on the liner implosion of closed field magnetically confined plasmas such as Field Reversed Configurations (FRCs) [T. P. Intrator et al., Phys. Plasmas 15, 042505 (2008)]. We present a semi-analytical model of liner driven FRC implosions that predicts the fusion gain of such systems. The model predicts a fusion gain near unity for an FRC imploded by a liner driven by the Z Machine. We show that FRCs could be formed and imploded at the Z facility using the AutoMag liner concept [S. A. Slutz et al., Phys. Plasmas 24, 012704 (2017)]. Furthermore, an initial bias magnetic field can be supplied by the external magnets used in MagLIF experiments. The reverse field is then supplied by an AutoMag liner, which has helical conducting paths imbedded in an insulating substance. Experiments [Shipley et al., Phys. Plasmas 26, 052705 (2019)] have demonstrated that AutoMag can generate magnetic fields greater than 30 Tesla inside of the liner. We have performed 2D Radiation MHD simulations of the formation and implosion of an FRC, which are in good agreement with the analytical model. The FRC formation process could be studied on small pulsed power machines delivering about 1 MA.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2013

Off-gas Adsorption Model and Simulation - OSPREY

The absence of industrial scale nuclear fuel reprocessing in the U.S. has precluded the necessary driver for developing the advanced simulation capability now prevalent in so many other countries. Thus, it is essential to model complex series of unit operations to simulate, understand, and predict inherent transient behavior. A capability of accurately simulating the dynamic behavior of advanced fuel cycle separation processes is expected to provide substantial cost savings and many technical benefits. To support this capability, a modeling effort focused on the off-gas treatment system of a used nuclear fuel recycling facility is in progress. The off-gas separation consists of a series of scrubbers and adsorption beds to capture constituents of interest. Dynamic models are being developed to simulate each unit operation involved so each unit operation can be used as a stand-alone model and in series with multiple others. Currently, an adsorption model has been developed within Multi-physics Object Oriented Simulation Environment (MOOSE) developed at the Idaho National Laboratory (INL). Off-gas Separation and REcoverY (OSPREY) models the adsorption of offgas constituents for dispersed plug flow in a packed bed under non-isothermal and non-isobaric conditions. Inputs to the model include gas composition, sorbent and column properties, equilibrium and kinetic data, and inlet conditions. The simulation outputs component concentrations along the column length as a function of time from which breakthrough data can be obtained. The breakthrough data can be used to determine bed capacity, which in turn can be used to size columns. In addition to concentration data, the model predicts temperature along the column length as a function of time and pressure drop along the column length. A description of the OSPREY model, results from krypton adsorption modeling and plans for modeling the behavior of iodine, xenon, and tritium will be discussed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗