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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 343 records · Page 19
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Demonstration of 151 Sm analysis using Trinity nuclear debris samples

Here, the long-lived fission product 151 Sm (T 1/2 ≈ 90 y) present in archived dissolved Trinity nuclear test debris samples was analyzed using liquid scintillation counting and the results were compared to the expected values based upon historical analysis results. The measured 151 Sm activities were found to agree with the expected values within experimental error. This demonstration validates the use of this technique in situations where the sample is too old for traditional shorter half-life radioactive fission product signatures such as 99 Mo (T 1/2 = 2.75 d), 97 Zr (T 1/2 = 16.7 h), 147 Nd (T 1/2 = 10.98 d), and 153 Sm (T 1/2 = 46.3 h), or too dilute for determining fissions through stable isotope ratio measurements.

151Sm↗
DOE OSTI2023

BlueSTEAl: A pair of silicon arrays and a zero-degree phoswich detector for studies of scattering and reactions in inverse kinematics

BlueSTEAl, the Blue (aluminum chamber of) Silicon TElescope Arrays for light nuclei, has been developed to study direct reactions in inverse kinematics, as well as scattering and breakup reactions using radioactive ion beams. It is a detector system consisting of a pair of annular silicon detector arrays and a zero-degree phoswich plastic scintillator. For typical binary reaction studies in inverse kinematics, light ions are detected by the Si array in coincidence with heavy recoils detected by the phoswich placed at the focal-plane of a zero-degree magnetic spectrometer. The Si array can also be used to detect light nuclei such as beryllium and carbon with clear isotope separation, while the phoswich can also be placed at zero degrees without a spectrometer and used as a high-efficiency beam counting monitor with particle identification capability at the rate of up to ~5 × 10 4 particles per second. This paper reports on the capabilities of BlueSTEAl as determined by recent experiments performed at the Texas A&M Cyclotron Institute. Furthermore, the device is also anticipated to be used in future experiments at other radioactive ion beam facilities.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2025

Constraining the Synthesis of the Lightest 𝑝 Nucleus 74 Se

We provide the first experimental cross section of the 73 As(p, γ) 74 Se reaction to constrain one of the main destruction mechanisms of the p nucleus 74 Se in explosive stellar environments. The measurement was done using a radioactive 73 As beam at incident energies of 3.1 and 3.7 MeV/nucleon. Along with the total cross section measurement, statistical properties of the 74 Se compound nucleus were extracted, constraining the reaction cross section across the entire Gamow window of the γ process. The impact of the experimentally constrained reaction rate on 74 Se production in Type II supernovae is investigated through Monte Carlo one-zone network simulations. The results indicate that the overproduction of 74 Se by Type II supernova models cannot be resolved by nuclear physics and point towards the description of the astrophysical conditions of the γ process.

59 ≤ A ≤ 89↗
DOE OSTI2020

An improved method for uranium separation from urine for rapid bioassay by inductively coupled plasma mass spectrometry

Bioassay monitoring is an important capability for any facility that works with significant amounts of radioactive material. The speed at which these analyses can be made is critical in the event of a potential exposure. Here, we present a new method that facilitates the rapid separation of uranium from a urine matrix for assay and isotopic determination by inductively coupled plasma mass spectrometry. This method is validated with archived urine samples that were previously analyzed using established, more time consuming methods. The sources of uncertainty for uranium assay and isotopic measurements are modelled using the Monte Carlo method to better understand the precision and accuracy of this method at the limit of quantification.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2020

Gas-phase molybdenum-99 separation from uranium dioxide by fluoride volatility using nitrogen trifluoride

Production of the important 99m Tc medical isotope parent, molybdenum-99 ( 99 Mo), via the fissioning of high- and low-enriched uranium (HEU/LEU) targets followed by target dissolution in acid and solution-phase purification of 99 Mo is time-consuming, generates quantities of corrosive radioactive waste, and can result in the release of an array of radionuclides to the atmosphere. An alternative 99 Mo purification method has been devised that has the potential to alleviate many of these issues. Herein, we demonstrate the feasibility of a rapid Mo/Tc gas-phase separation from UO 2 . The results indicate that volatile [ 99 Mo]Mo can be captured downstream of the reacted solid mixture on a column bed (trap) of alumina; the majority of the captured [ 99 Mo]Mo can be subsequently eluted from the alumina trap with a few milliliters of water. >1.0 × 10 5 single pass decontamination of U and the collected [ 99 Mo]Mo product is demonstrated. This simple thermo-fluorination technique has the potential to provide a rapid methodology for routine 99 Mo production.

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

VENTSAR V3.0: A Python GUI for Estimating Contaminant Concentrations on or Near Buildings Due to Building Effects and Plume Rise and For Calculating Inhalation and Plume Shine Doses

VENTSAR: Originated as VENTAX (Smith and Weber 1983) on the IBM Mainframe at the Savannah River Site (SRS) as a Fortran program. Updated to VENTSAR XL V1.0 (Simpkins 1997) as a spreadsheet version using macros created in Microsoft Excel. Later updated to VENTSAR XL V2.0 (Dixon 2018) as a spreadsheet executing a modified version of the original VENTAX Fortran code. Estimates contaminant concentrations on or near a building from a release at a nearby location. Calculates concentrations for a given meteorological exceedance probability or for a given stability and wind speed combination. Can model a single building with or without a penthouse on top or a ground location from either a stack or ground release. Plume rise can be considered. Contaminant releases can be chemical or radioactive with downwind concentrations determined at user-specified distances. Wind passing over and around buildings creates a complicated dispersion pattern. Air-intake vents may be located on building roofs or near the ground downwind of a release source. An estimation of pollutant concentrations on or near a structure is important in determining expected pollutant levels. Meteorological data are selected based on a specific area of the site. Fortran-based VENTAX able to make fast, complex mathematical calculations. Not user-friendly VENTSAR XL V1.0 no longer supported due to obsolete Excel macros. VENTSAR XL V2.0 an interim solution using an Excel spreadsheet to interface with the modified VENTAX Fortran code. Requires user to have Microsoft Excel. Not an intuitive interface for someone unfamiliar with VENTSAR. Does not perform dose calculations. VENTSAR V3.0 goal to combine the strengths of previous versions of VENTSAR into a single, powerful yet user-friendly program with a Graphical User Interface (GUI). Not dependent upon a specific program or plug-ins. Self-contained package to be used on any computer running Microsoft Windows. User-friendly, intuitive GUI created in Python for parameter input. Executes same modified VENTAX Fortran code as VENTSAR XL V2.0. Outputs formatted text file of completed calculations for easy review and dissemination. Performs inhalation and plume shine dose calculations for up to 11 user-selected nuclides from a nuclide dose factor library of almost 500 nuclides. 12 test cases were created for verification of V1.0 and V2.0. Same test cases run in V3.0 to verify correct performance. V3.0 produced similar results to V1.0. Confirmed GUI did not alter calculations in any way. Comparisons of the test case results confirm the V3.0 GUI does not influence the VENTSAR calculations. Simply passes same input parameters to VENTAX Fortran code for execution. Provides end user an easy-to-use, intuitive tool to quickly make building effect and plume rise calculations, as well as, inhalation and plume shine dose calculations. No experience with or working knowledge of Fortran, Excel, command line, or macros required. Self-contained package allows VENTSAR be deployed to any Windows computer without the need for specific programs or plug-ins.

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

Modeling Approach for the Aluminum-clad Dry Storage Pilot using HFIR Fuel

To confirm that the dry storage of aluminum-clad research reactor spent nuclear fuel (ASNF) will remain within the safety envelope after applied drying schemes and that the resulting evolution of the gas space composition, temperature, and pressure conditions are understood, a dry storage pilot project is being established. The pilot will incorporate an instrumented lid for discrete interval or for on-demand gas composition and temperature monitoring of two DOE Standard Canisters (DSCs) loaded with three High Flux Isotope Reactor (HFIR) inner cores per DSC. Each DSC would be subjected to a separate alternative candidate drying scheme. Canisters will undergo 1 to 5 years of monitoring, including internal temperature and gas sampling to track pressure and composition changes. This report outlines the approach for modeling the ASNF-in-canister behavior in terms of evolving gas space conditions for the ASNF dry storage pilot using HFIR fuel. The ASNF has an adherent surface oxyhydroxide layer comprised of boehmite/bayerite that generates hydrogen when subjected to irradiation. Three-dimensional multi-physics computational fluid dynamics simulations will be executed to compute the thermal field within the DSC and provide inputs to a chemical model employed to compute pressure buildup as hydrogen is generated in the system. Implemented in Cantera, the chemical model solves gas phase and aluminum oxyhydroxide surface-mediated radiolysis reactions. Gas phase reactions are sourced from Wittman and Hanson (2015), whereas surface-mediated reactions are incorporated by fitting experimental data using an optimization algorithm (Abboud, 2023). Water radiolysis reactions from Wren and Ball (2001) are adopted with modifications as described in Abboud (2023c). Understanding the effect of the hydrogen buildup over time is important for long-term storage safety considerations. Modeling results will include the canister pressure, temperature, and composition evolution from the initial helium backfill with the addition of radiolytically-evolved chemical species (e.g., hydrogen and oxygen). The specific HFIR cores for the pilot program have not yet been selected, and the overall design is still in development. The CFD-chemical model used for this work will be based on prior models with necessary updates to allow for improved accuracy and efficiency. The experimental data obtained from the HFIR demonstration will be used to improve and validate the computational models to predict the ASNF-in-canister behavior.

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

Investigation of the excited states of 114 Sn using the GRIFFIN spectrometer at TRIUMF

The semi-magic 110-122 Sn isotopes display signs of shape coexistence in their excited 0 + states, which, in contrast to the spherical 0+ ground states, are deformed. Here, this paper investigates the nuclear structure of 114 Sn using the competing β + decay and electron capture of a radioactive beam of 114 Sb produced at the TRIUMF-ISAC facility using the GRIFFIN spectrometer. This study will allow for an in-depth understanding of the excited 0 + states in 114 Sn, by focusing on their decay patterns. In the present experiment, transitions at 856.2-keV and 1405.0-keV, which were observed in an earlier β + decay study but not placed in the 114 Sn level scheme, have been Physics, assigned to the level scheme in connection to the 0$^+_3$ level at 2156.0-keV. Properly assigning these transitions refines the level scheme and enhances our understanding of the nuclear structure in 114 Sn.

GRIFFIN↗
DOE OSTI2020

Measurements of Extinct Radionuclides in Filter Paper Samples

Radiochemical diagnostics rely on measurements of short-lived fission product isotopes. Once months to years have passed since isotope formation, the most useful short-lived signatures have decayed away and become extinct. Recent work at our lab has developed new methods to reconstruct the concentrations of extinct radionuclides using mass spectrometry to measure the stable daughter elements formed from radioactive decay. However, the applicability of these techniques to different types of samples has not been fully investigated.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2022

Report to NCSP on FY22 DANCE and NEUANCE measurements of 233 U(n, γ)

The experiment was performed at the end of the CY20 runcycle and CY21. A total of 20 days were measured using the 20 mg sample, and the 10 mg sample was placed in the beam for a total of 2 days. The rest of the beam time was used to measure radioactive γ sources for energy calibration, background measurements, and tests to define the 233 U windows required during the data taking.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2021

Development of an Activation Analysis Methodology to Support the Disposal of the High Flux Isotope Reactor Metal Pool Waste

During operation and maintenance of the High Flux Isotope Reactor (HFIR), metal waste from reactor components is generated. Due to space limitations in the pool fuel storage, it is necessary to remove the waste and send it to off-site disposal. Most of this waste was in the reactor and is radioactive through neutron activation. Prior to shipping and disposal at a waste disposal facility, the radiological activity of the waste must be determined. Therefore, a conservative methodology to characterize the radiological activity of the waste must be performed. The methodology used was based on being able to determine a conservative overall activity and radionuclide inventory based on radiological survey data and neutron activation products for the materials present.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Impact of UV-light and pH on the Fate of Tc, I, and U in Wetlands at Savannah River Site - 20230

The Savannah River Site (SRS) is one of the major nuclear facilities owned by the U.S Department of Energy. During the Cold War, these facilities produced large amounts of radioactive and hazardous waste. On site, three unlined seepage basins in the F-Area received approximately 1.8 billion gallons of low-level waste containing nitric acid, radionuclides, and dissolved metals due to plutonium and tritium production operations. The acidic nature of the waste created a source of relatively mobile radionuclides below the basins. Radionuclides previously disposed of within the F-Area, including uranium isotopes (U), technetium-99 (Tc-99), and iodine-129 (I-129), are moving with groundwater towards Four Mile Branch Wetland, where they are subsequently upwelling and interacting with natural organic matter (NOM). Many environmental factors including redox conditions, porewater composition, pH, light, and temperature affect the degradation of organic matter as well as interactions with Tc-99, U, and I-129. In particular, the high concentrations of nitrate from the acidic waste may increase the formation of reactive oxygen species (ROS) that impact both degradation of NOM and behavior of contaminants. In the presence of sunlight, I-129, Tc-99, and U speciation and complexation behavior may be affected by ROS and NOM degradation products in addition to the factors generally considered in subsurface systems in the absence of light. This research aims to determine whether the interactions between radionuclides, NOM, and nitrate affect the fate of I-129, Tc-99, and U and which processes are controlling their behavior. Experiments were conducted at variable pH in the presence of NaNO{sub 3} in order to determine the impact of light and pH on NOM degradation and to evaluate the impact on the fate of contaminants of concern. Soil samples high in NOM collected from two sites in the Southeastern United States (Savannah River Site and the Everglades) were studied. Batch experiments were conducted with NOM and the following aqueous contaminants, U, I-129, Tc-99 with exposure to ultraviolet (UV) light in an environmental chamber. This study was conducted in a sterile environment in order to exclude the potential for microbial degradation of organic matter. Results indicated photodegradation of NOM and significant interaction of radionuclides with NOM. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

The distribution and selective decontamination of carbon-14 from nuclear graphite

This work investigates 14 C release behaviour from UK Magnox reactor sourced irradiated graphite under relatively low temperature oxidation conditions: a 1% oxygen in argon atmosphere at temperatures ranging from 600°C to 700°C, and durations from 4 to 120 hours. A method is used for construction of a detailed 14 C release profile with mass loss. The 14 C release profile is predictable between samples, and an empirical release profile is derived to predict releases to higher mass losses than in this work. An accelerated 14 C release rate at small mass losses is observed, indicative of a 14 C-enriched surface region with a depleting concentration gradient into the material, and the source terms and complexities associated with predicting where the 14 C arose from are discussed. Selective decontamination of a fraction of the 14 C is possible, with the limitation of a reducing efficiency with mass loss and time. This work demonstrates a method for determining the distribution of 14 C between near-surface regions and the bulk material in differing sources of irradiated graphite, and provides data in support of making an informed assessment for the adoption of low temperature thermal treatment and isotopic reduction of irradiated graphite wastes.

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

Computational Optimization of 133m Xe Production via Neutron Irradiation in a TRIGA Reactor

Here, the Comprehensive Nuclear-Test-Ban Treaty bans all nuclear tests worldwide. As part of treaty compliance, the concentration of radioactive nuclides in the atmosphere is monitored to detect nuclear weapons tests. Radioactive noble gas fission products, specifically radioxenon, can vent into the atmosphere after a nuclear weapons test, even if the test is well contained underground or underwater. Radioxenon thus serves as a signal for nuclear weapons tests. All atmospheric monitoring systems require samples of radioxenon isotopes for detector calibration, quality control, and certification. Here, we present a novel, improved method for creating samples of 133m Xe via neutron irradiation of 132 Xe in the Washington State University TRIGA reactor. 132 Xe neutron absorption results in either 133 Xe or 133m Xe—thermal neutron absorption results in 133m Xe 12% of the time, while fast neutron absorption (above ~1 MeV) results in 133m Xe ~50% of the time. To optimize the production of 133m Xe via neutron absorption in 132 Xe in the thermal TRIGA reactor, spectral tuning using an irradiation chamber is required to maximize the fraction of fast neutrons being absorbed and minimize the number of thermal neutrons interacting with the 132 Xe. We used MCNP to tally 132 Xe absorptions with the isotopic tally function, flux tallies and neutron attenuation to estimate the number of neutrons reaching the 132 Xe through the irradiation chamber, and the adjoint importance function to improve the source strength estimate. Additionally, we performed a heat transfer analysis for safety considerations. It was determined that the use of a 96% enriched 10 B boron carbide chamber, placed next to the fuel elements in reactor position D8, increases the 133m Xe/ 133 Xe activity ratio from a baseline value of 0.3 to 1.0, a 233% increase. Additionally, it was determined that the alpha heating produced in the boron does not become an unmanageable problem in the Washington State University reactor.

37 - INORGANIC, ORGANIC, PHYSICAL AND ANALYTICAL C↗
DOE OSTI2021

Temperature Effect of Cesium Exchange onto Crystalline Silicotitanate in AP-107 and AP-105 Hanford Tank Wastes and Two Simulants

Washington River Protection Solutions, LLC (WRPS) is charged with the development of the Tank Side Cesium Removal (TSCR) system to process Hanford tank waste supernates in preparation for vitrification. In addition to a filtration step, TSCR will remove cesium (Cs) using ion exchange columns filled with crystalline silicotitanate (CST) ion exchange media. CST is produced by Honeywell UOP, LLC. The documented safety analysis (DSA) developed for the TSCR system limits a single column loading to 141,600 Ci 137 Cs. Given a 137 Cs isotopic mass fraction of 20% and the planned CST bed size of 596 L (157.5 gal) in a TSCR column, this equates to 0.10 mmole Cs per g CST (Cs distribution coefficient, K d , 1400 mL/g). Factors that influence Cs uptake by CST include (but are not limited to) (1) CST production (lot-to-lot variations), (2) contact temperature, (3) contact duration, (4) competitors in the tank waste feed, (5) anionic composition of the tank waste feed, and (6) the 137 Cs isotopic mass fraction (differs slightly among tank wastes and decreases with time).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

UNESE Argon-39 Measurement Techniques: Developing an above-ground Argon-39 Measurement Capability

The Underground Nuclear Explosion Signatures Experiment (UNESE) sought to use 37 Ar as a tracer for measuring noble-gas migration in the soil surrounding historic Underground Nuclear Explosions (UNE). One unexpected observation was the presence of the much longer-lived isotope 39 Ar from historic UNEs. Quantifying the activity of 39 Ar proved difficult due to the lack of capability to measure significantly-above-background levels of 39 Ar and a general discomfort to repeatedly expose Ultra-Low-Background Proportional Counters (ULBPCs) to significant radioactivity. Because the whole-air samples collected for the 37 Ar tracer measurement were already being measured on the above-ground argon capability, it was decided to expand that capability to include 39 Ar. This document describes the efforts required to achieve quantitative reporting of the 39 Ar backgrounds measured during experiments at the sites of the historic Barnwell and Disko Elm UNEs.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

Study of cosmogenic activation above ground for the DarkSide-20k experiment

The activation of materials due to exposure to cosmic rays may become an important background source for experiments investigating rare event phenomena. DarkSide-20k, currently under construction at the Laboratori Nazionali del Gran Sasso, is a direct detection experiment for galactic dark matter particles, using a two-phase liquid-argon Time Projection Chamber (TPC) filled with 49.7 tonnes (active mass) of Underground Argon (UAr) depleted in 39 Ar. Despite the outstanding capability of discriminating γ / β background in argon TPCs, this background must be considered because of induced dead time or accidental coincidences mimicking dark-matter signals and it is relevant for low-threshold electron-counting measurements. Here, the cosmogenic activity of relevant long-lived radioisotopes induced in the experiment has been estimated to set requirements and procedures during preparation of the experiment and to check that it is not dominant over primordial radioactivity; particular attention has been paid to the activation of the 120 t of UAr used in DarkSide-20k. Expected exposures above ground and production rates, either measured or calculated, have been considered in detail. From the simulated counting rates in the detector due to cosmogenic isotopes, it is concluded that activation in copper and stainless steel is not problematic. The activity of 39 Ar induced during extraction, purification and transport on surface is evaluated to be 2.8% of the activity measured in UAr by DarkSide-50 experiment, which used the same underground source, and thus considered acceptable. Other isotopes in the UAr such as 37 Ar and 3 H are shown not to be relevant due to short half-life and assumed purification methods.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2021

Hydride Bed Helium-3 Recovery and Partial Regeneration

Savannah River Tritium Enterprise has used LaNi 4.25 Al 0.75 (LANA75) hydride beds to store hydrogen isotopes for over two decades. A benefit of using LANA75 is that the 3He generated from tritium decay is retained in the hydride material, allowing the hydride beds to deliver high-purity product gas. A disadvantage is that the 3He accumulates in the LANA75 material over time, which forms a heel that cannot be removed under normal operating conditions. The heel traps hydrogen in the bed, slowly reducing the operational capacity of the bed as the heel grows. Eventually, the 3 He begins to release from the material, preventing the delivery of high-purity product. The hydride beds are replaced when (1) operational capacity is reduced such that it is impactive to routine operations, and/or (2) product purity is not maintained due to 3 He release.Several beds were operated beyond their design life. One of these beds was selected to undergo heating beyond its normal operating temperature to evaluate the possibility of removing a portion of the hydrogen and helium heel to improve bed function until a replacement could take place. This bake-out removed a portion of the hydrogen and helium heel, and preliminary data indicate that bake-outs may partially regenerate the beds. The bed’s performance will continue to be monitored, and additional bake-outs will likely be performed. Performing bake-outs results in increasing the recovery of 3 He, more efficient end-of-life activities (such as isotopic exchange), and extension of the useful service life of the bed.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗