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At least 541 records · Page 30
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Anomalous isotope effect on the optical bandgap in a monolayer transition metal dichalcogenide semiconductor

Isotope effects have received increasing attention in materials science and engineering because altering isotopes directly affects phonons, which can affect both thermal properties and optoelectronic properties of conventional semiconductors. However, how isotopic mass affects the optoelectronic properties in 2D semiconductors remains unclear because of measurement uncertainties resulting from sample heterogeneities. Here, we report an anomalous optical bandgap energy red shift of 13 (±7) milli–electron volts as mass of Mo isotopes is increased in laterally structured 100 MoS 2 - 92 MoS 2 monolayers grown by a two-step chemical vapor deposition that mitigates the effects of heterogeneities. This trend, which is opposite to that observed in conventional semiconductors, is explained by many-body perturbation and time-dependent density functional theories that reveal unusually large exciton binding energy renormalizations exceeding the ground-state renormalization energy due to strong coupling between confined excitons and phonons. The isotope effect on the optical bandgap reported here provides perspective on the important role of exciton-phonon coupling in the physical properties of two-dimensional materials.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Fission product yields from the 238 U ( n,f ) reaction at E n = 4.6 MeV

Here, we present cumulative fission product yields from the 238 U (n, f) reaction at an incident energy of 4.6 MeV for fission products with half-lives from less than one second to several hours. We employed direct γ-ray spectroscopy to measure unseparated fission products with a conventional neutron activation analysis and a cyclic activation technique using the RApid Belt-driven Irradiated Target Transfer System (RABITTS). Yields were determined for 78 isotopes and isomers, covering the mass range 84 ≤ A ≤ 148 and the charge range from 33(As) ≤ Z ≤ 58(Ce). Our experimental cumulative yields are compared to the literature and evaluated data at E n ≈0.5 MeV from ENDF/B-VIII.0 and JEFF-3.3 libraries and with theoretical models using the GEF and FREYA codes at E n =4.6 MeV.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2025

Core Physics Characteristics of Extended Enrichment and High Burnup Boiling Water Reactor Fuel

This paper presents the highlights of boiling water reactor (BWR) core physics studies performed at Oak Ridge National Laboratory as part of a series of studies conducted to compare low-enriched uranium (LEU) with LEU+ fuel. The studies analyzed isotopic fuel content, lattice parameters (Phase 1), and core physics (Phase 2) to identify challenges in operation, storage, and transportation for BWRs and pressurized water reactors (PWRs). Because of a lack of publicly available lattice and core designs for modern BWR fuel assemblies and reactor cores, several optimized lattice designs were generated, and different core loading strategies were investigated. Twelve optimized lattice designs with 235 U enrichments ranging from 1.6% to 9% and gadolinia loadings ranging from 3 to 8 wt% were used to model axial enrichment and geometry variations in fuel assemblies for core designs. Each core shares a common set of approximations in design and analysis to allow for consistent comparisons between LEU and LEU+ fuel. The objective is to highlight anticipated changes in core behavior with respect to the reference LEU core. The results of this study show that the differences in LEU and LEU+ core reactor physics characteristics are less significant than the differences in lattice physics characteristics reported in the Phase 1 studies.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1988

Iodine-xenon studies of Allende inclusions - Eggs and the Pink Angel

The I-Xe systems of six Allende inclusions (five Eggs and the Pink Angel) appear to have been altered by nonnebular secondary processes. Evidence for this includes temperature-ordered variations in the initial I isotopic composition within several objects (with older apparent I-Xe ages associated with higher extraction temperatures) and the absence of primitive I-Xe ages. The span of apparent ages seen in Allende objects (10 Myr or more) is probably too long to reflect any nebular process, so at least some alteration probably occurred ont the parent body. The range in initial (Pu-244)/(U-238) ratios for the Eggs (0.003-0.014) includes the current best estimates of the bulk solar system value (0.004-0.007). For Egg 3, the Pu/U ratio varies by a factor of two between extractions, probably the result of fractionation of Pu from U among different phases.

Swindle, T. D.↗
DOE OSTI2021

Measurements of effective actinides neutron capture cross sections in a cadmium-filtered epithermal neutron spectrum

In this work, a total of 20 actinide samples were irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL): 233 U (1), 235 U (1), 236 U (1), 238 U (1), 237 Np (2), 239 Pu (1), 240 Pu (1), 242 Pu (2), 244 Pu (2), 241 Am (2), 243 Am (2), 244 Cm (2), and 248 Cm (2). The samples were loaded in a basket containing a 1.1-mm thick cadmium filter. The presence of the filter drastically changes the neutron spectrum in the samples and accentuates the reaction rates in the resonance region. Isotopic ratios were determined using Multi-Collector Inductively Coupled Plasma Mass Spectrometry (MC-ICP-MS) that allowed inferring effective capture cross sections. Detailed best-estimate MCNP models of the experiment using ENDF/B-VII.0 nuclear data were developed to calculate the effective neutron cross sections and, subsequently, compare them with the cross sections inferred from experimental results.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Measuring Signatures of Fuel Irradiation in Large Particle Samples

Understanding the operating history of a nuclear facility is one of the key goals in nuclear safeguards, enabling the verification of declared activities. This is accomplished via a number of mechanisms, including the analysis of samples collected at the facility. The best metrics for understanding and verifying the history of nuclear fuel are (1) actinide isotopic content and (2) actinide inter-element ratios. While techniques have long existed for measuring these values with high accuracy and precision in bulk samples, as samples become smaller – such as may be collected in environmental collection – new methods are required both for clean processing of pure samples, and for analysis of these small samples. This study describes a new extraction chromatography technique for recovering U, Pu, and Am from single small aliquots, and advanced mass spectrometry methods for high-precision and high-accuracy analysis of trace and ultra-trace levels of actinides. These techniques are applied to particle samples cut from fuel from a historic nuclear reactor. Ages calculated from Pu–Am chronometry of particles are in excellent agreement with the known operating history of the reactor.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

A Rapid Sequential Separation Method for Determination of Actinides and TENORM in Fracking Wastes - 20223

With recent advances in unconventional drilling technology in the US and around the world, the risks of environmental contamination and exposure due to technologically enhanced naturally occurring radioactive materials (TENORM) from fracking wastes have also surfaced. Although the Permian Basin (southeastern New Mexico, western Texas) has long been a modest producer of oil, the advent of fracking technology a decade ago turned it into hot property for producers. A 2018 assessment of undiscovered, technically recoverable continuous oil and gas resources by the U.S. Geological Survey estimates an average 46.3 billion barrels of oil and 281 trillion cubic feet of gas. This is likely to increase the use of unconventional drilling and TENORM generation. TENORM concentrations in oil and gas exploration and production waste can be several hundred to several thousand times higher than background TENORM concentrations. Both the thorium and uranium decay chains contribute to airborne radionuclides arising from the radon gas escaping the ground and subsequently decaying as airborne particulates. Additionally, radon decay products {sup 210}Pb and {sup 210}Po can build up in scale on the internal surfaces of oil and gas handling pipes and in sludge in refineries, becoming potential inhalation and ingestion hazards for workers. Furthermore, southeastern NM is also home to world's only licensed and operating transuranic nuclear waste repository, the DoE's Waste Isolation Pilot Plant (WIPP). Plutonium isotopes ({sup 239+240}Pu) and {sup 241}Am, are expected to account for more than 99% of the total radioactivity scheduled for disposal in the WIPP repository. Thus, an improved understanding of the environmental fate and transport of TENORM, liberated by unconventional drilling, is essential to assess how best to protect individuals and the environment. In this context, accurate measurement of TENORM and actinides in environmental and biological samples is essential. In this presentation, a new sequential method for the separation and pre-concentration of actinides (Pu, Am, Np) and TENORM (Po, U and Th) derived from oil and gas exploration is proposed. The TEVA method involves a rapid co-precipitation step to remove matrix interferences from the samples, followed by plutonium oxidation state adjustment to Pu (IV) and an incubation period of ∼ 1 hour at 50-60 deg. C to allow the resultant Po (II) to oxidize into Po(IV). The polonium, neptunium, thorium and plutonium are then separated on a TEVA column, while americium and uranium are separated on a TRU column. After separation, the alpha counting source was prepared by micro-precipitation with copper sulfide (CuS) for polonium and neodymium fluoride (NdF{sub 3}) for actinides. The efficiency and reliability of the procedures were tested by analyzing filter, drinking water and frack sand samples. The method is simple, robust and can be performed quickly with excellent removal of interferences, high chemical recovery and very good alpha peak resolution. (authors)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2023

Challenges in correlating oxygen stable isotope ratios of hydrates on uranium ore concentrates to process waters

Exchange of oxygen stable isotopes (δ 18 O values) between precipitation waters and uranium oxides is governed by thermodynamics or kinetics. It has been assumed that meteoric waters can be related to precipitation waters in uranium ore concentrates and their calcination and reduced uranium oxide products. With this assumption, the δ 18 O values of uranium materials could provide forensic signatures that identify the production history and geolocation of nuclear materials. To further exploit the potential of δ 18 O values in nuclear material analysis, this study examines the oxygen stable isotope exchange in two UOCs, magnesium diuranate (MDU) and sodium diuranate (SDU). MDU and SDU were synthesized from solutions of uranyl nitrate hexahydrate using precipitation waters with unique oxygen isotope compositions. The structures of the MDU and SDU were analyzed using powder X-ray diffraction (p-XRD) and thermal mass loss curves, while the δ 18 O values of waters generated during thermal decomposition were analyzed using a thermogravimetric analyzer coupled to an isotope ratio infrared spectrometer (TGA-IRIS). By p-XRD, the MDU was uniform and amorphous across all syntheses with residual crystalline material incorporated as a minor component. Combined with the TGA results, all of the MDU is likely amorphous MgU 2 O 7 ·3H 2 O with MgO impurities present throughout. In contrast, the SDU synthesis resulted in multiple phases with many samples exhibiting crystalline phases including a combination of Na(UO 2 ) 4 O 2 (OH) 5 ·5H 2 O and Na 2 (UO2) 6 O 4 (OH) 6 ·8H 2 O with a Na 2 U 2 O 7 minor phase. A small fraction of the SDU samples were amorphous with no crystalline XRD peaks observed. Mass loss curves of the SDU samples revealed that the amorphous samples contained inclusions of similar crystalline phases compared to the crystalline materials. The uniformity of the MDU samples enabled highly reproducible measurements of δ 18 O values of the water vapor yielded for two dehydration events at 170 °C and 500 °C. In contrast, the multiphase composition of the SDU samples resulted in poor reproducibility in δ 18 O values. In conclusion, neither system revealed any correlation between the δ 18 O values of precipitation water, and the waters released during dehydration of the UOCs.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2022

Solvent Recovery and Management at the Savannah River Site H-Canyon Facility

NIOWAVE, Inc. is a domestic supplier of medical and industrial isotopes from uranium and radium. The Savannah River National laboratory (SRNL) is currently providing support to NIOWAVE, which plans to deploy a superconducting electron linear accelerator (LINAC) to fission uranium for Mo-99 production without the need for a nuclear reactor or HEU. The uranium from the Mo-99 production targets will be purified using a modified PUREX (Plutonium Uranium Reduction Extraction) solvent extraction process to recover the uranium in the product stream. The uranium will then be precipitated as an oxalate which is calcined to U 3 O 8 to fabricate pellets for new Mo-99 targets. In previous support provided to NIOWAVE, the SRNL demonstrated a solvent washing process to remove degradation products from the tributyl phosphate (TBP) solvent used in the modified PUREX process under development for uranium recovery. To supplement this technology demonstration, NIOWAVE requested the SRNL to provide summary information on the solvent recovery and management activities which are used at the Savannah River Site (SRS) H-Canyon facility. An existing reference document for the reprocessing of irradiated HEU fuels at the SRS was used as the primary reference for the solvent management activities; although, other reference documents were used to provide supplementary information. The information provided includes a brief summary of the solvent degradation issues which have been observed in the H-Canyon solvent extraction cycles and resulting process safety concerns. The solvent recovery processes for the three cycles of solvent extraction used in the H-Canyon were subsequently described including the process equipment which consists of the continuous and batch solvent washers, pumps, and tanks. A final section is provided on the monitoring and analysis of solvent quality based on the previous work performed at the SRNL for NIOWAVE and past research and development activities performed to support the solvent extraction processes in both the SRS F-Canyon and H-Canyon facilities.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2026

Summary of the Initial Post-Irradiation Characterization of HFIR-Irradiated Low-N and High-N HT-9 Steel

Reference cladding systems for sodium fast reactors are based on the historical steel, HT-9. HT-9 is a Fe12Cr ferritic/martensitic steel with additions of Mo, W, V, and other minor elements and demonstrates low irradiation swelling and adequate mechanical properties. Extensive irradiation literature exists on the use of HT-9 as cladding for metal fuel, primarily irradiation on the U-Zr/HT-9 system from the Experimental Breeder Reactor-II (EBR-II) and Fast Flux Test Facility (FFTF) sodium fast reactor, and as a structural material from experiments in the FFTF. The large amount of historical data makes the U-Zr/HT-9 system the reference fuel specification for many nuclear reactor vendors that seek to license modern sodium-cooled fast reactors in the United States. However, it is yet unclear how variations in impurity content within HT-9 fundamentally affect irradiation performance at various irradiation temperatures. Recent work suggests that impurity content may noticeably alter the production of helium through nuclear transmutation. For these reasons, High-Flux Isotope Reactor (HFIR) irradiation of HT-9 steels with known variations in the impurity content is particularly timely to generate data to enable more accurate refinement of the chemical specification for nuclear-grade HT-9 material. This report summarizes the initial transmission electron microscopy characterization of HFIR-irradiated HT-9 steels following mechanical property measurements by the Advanced Fuels Campaign (AFC). This report includes qualitative results of the cavity, dislocation loop and cluster/precipitate microstructures as well as radiation-induced segregation. Quantitative results are being shared with partner institutions and will be included in more detail in a future report in FY2026.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Nuclear data uncertainty propagation applied to the versatile test reactor conceptual design

We report the Versatile Test Reactor (VTR) currently under development is a 300 MWth sodium-cooled fast reactor (SFR) fueled with ternary metal alloy fuel, which aims to accelerate the testing of advanced nuclear fuels, materials, instrumentation, and sensors in high flux environments that are necessary to license the next generation of advanced reactor concepts. To support the VTR design process, uncertainties associated with the nuclear data has been propagated through the reactor core neutronics calculation to global parameters of interest, such as the core multiplication factor, kinetic parameters, and various reactivity feedback coefficients, following the sensitivity based uncertainty propagation approach. By folding the sensitivity coefficients, separately computed by the generalized perturbation theory code PERSENT and Monte Carlo code Serpent 2, with the variance-covariance matrices from COMMARA-2.0, we obtain the reaction-wise, isotope-wise, and overall uncertainties for each response of interest due to nuclear data uncertainty. With Serpent 2, the statistical error of the uncertainty is obtained by propagating the statistical error of the sensitivity coefficients through the same process using a newly developed uncertainty propagation method. From both codes, the overall top uncertainty contributors are found to be the cross section of Fe-56 elastic scattering, Na-23 elastic scattering, and U 238 inelastic scattering. The large contributions of the Fe-56 elastic scattering cross sections to global parameters are due to its relatively large relative uncertainty of 5–10% in nuclear data and the large volume of Fe-containing reflector assemblies in the fairly compact VTR core design. Both codes agreed well for the overall uncertainty estimates of all responses of interest, except the delayed neutron fraction, prompt neutron generation time, and the coolant density feedback coefficient, where Serpent 2 yielded a much larger value than PERSENT due to the large statistical error of sensitivity coefficients. The calculated uncertainties are also compared to those associated with other SFR cores. Another outcome of this study is a variance-covariance matrix of reactivity coefficients, which can be used in the subsequent uncertainty propagation to the system level to investigate the impact of identified uncertainties on system responses in the safety analysis.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Assessment and Planning of HFIR Test Articles

Burnable absorbers (BAs) are introduced into nuclear fuels to aid in controlling the activity of the irradiated fuel at the beginning of life in a light water reactor (LWR). This improves the efficiency of the reactor since it allows for higher enrichments (> 5 % 235 U) to be used in a LWR, which increases the burnup limit and therefore the cycle length. The purpose of the BA is to absorb neutrons to prevent power peaking, and the concentration of BA is consumed under irradiation. In order to understand the effect of the BAs on the microstructure and thermophysical properties of the fresh and irradiated fuel, neutron irradiation tests will be carried out at the High Flux Isotope Reactor (HFIR). These tests will include Gd 2 O 3 doped UO 2 fuels with concentrations that vary from 4 to 10 weight % Gd 2 O 3 .

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Understanding Microscopic Mechanisms of LeTID and LID and their Unifying Features by Electron Paramagnetic Resonance

LID degradation involves not only creation of ~ 10^12 cm^-3 recombination centers, but also ~ 10^16 cm^-3 shallow negative-U traps. In Ga-doped Si, LID EPR defects don't appear, but some traps are still created. LeTID: Si DB and H-hyperfine EPR signatures. We postulate that the defect responsible for LeTID is a partially hydrogenated (multivacancy) with a Si dangling bond and H in the vicinity. O involvement is possible yet unclear. We prove that H is related to the structure of the LeTID defect with isotope experiments and its EPR signal is comparable and linear with the Si DB signal upon LeTID degradation. Working on simulating these results with DFT to obtain more detailed defect structure.

Cz Si↗
DOE OSTI2025

Fabrication and Characterization of MiniFuel Specimen

Burnable absorbers (BAs) are used in UO 2 fuels to tune the reactivity in a light water reactor (LWR) particularly at the beginning of life of the fuel. BAs are elements that have high neutron absorption cross sections and because of this property can prevent power peaking. The addition of BAs reduces the reactivity at the beginning of life, and this in turn allows for the increase of enrichments from > 5 % 235 U up to ~ 10 %. This also enables the increase of the burnup limit to 75 GWd/MTU and the cycle length to 24 months. The BA are eventually consumed under irradiation, and their concentration is diminished at roughly 10-20 GWd/MTU. The objective of this report is to fabricate UO 2 fuels that have been doped with BAs (Gd 2 O 3 in this case) and subsequently irradiate them in the high flux isotope reactor (HFIR) at Oak Ridge National Laboratory. The test specimen that are fabricated include 4 and 8 weight % Gd 2 O 3 doped UO 2 to investigate the influence of the BAs on the thermophysical properties and microstructure of the fresh and irradiated fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1988

The early archaean crustal history of West Greenland as recorded by detrital zircons

The isotope systematics of some of the oldest samples on Earth from both Greenland and Australia was discussed. The antiquity was confirmed of the 4.1 to 4.2 Ga zircons from Western Australia; the model Lu-Hf age of these zircons, as measured with the ANU ion probe is 4.14 + or - 0.24 Ga, although the oldest preserved rock units there are anorthosites with a Lu-Hf model age of about 3.73 Ga. U-Pb ion probe ages of detrital zircons ranging between 2.87 and 3.89 Ga from an Akilia association quartzite was reported, whose age of deposition is probably around 3.8 Ga. It was argued that the younger age in this range are discordant because of late Pb-loss, probably associated with a high grade metamorphic event at about 3.6 Ga. It was also argued that the earliest crust in West Greenland and elsewhere is about 3.9 Ga, but in some places, such as Western Australia, crustal evolution took place much earlier, perhaps starting as far back as 4.3 Ga. This would account for the presence in that terrane of abundant K rich granitoid, the paucity of tonalitic and trondhjemitic materials, and the existence of Eu anomalies in early Archean sediments.

Kinny, P. D.↗
NASA NTRS1991

Stable hydrogen and carbon isotope ratios of extractable hydrocarbons in the Murchison meteorite

A fairly fool-proof method to ensure that the compounds isolated from meteorites are truly part of the meteorites and not an artifact introduced by exposure to the terrestrial environment, storage, or handling is presented. The stable carbon and hydrogen isotope ratios in several of the chemical compounds extracted from the Murchison meteorite were measured. The results obtained by studying the amino acids in this meteorite gave very unusual hydrogen and carbon isotope ratios. The technique was extended to the different classes of hydrocarbons and the hydrocarbons were isolated using a variety of separation techniques. The results and methods used in this investigation are described in this two page paper.

Krishnamurthy, R. V.↗
NASA NTRS1998

Detection of atomic deuterium in the upper atmosphere of Mars

High-resolution spectroscopy of Mars' atmosphere with the Hubble Space Telescope revealed the deuterium Lyman alpha line at an intensity of 23 +/- 6 rayleighs. This measured intensity corresponds to HD/H2 = 1.5 +/- 0.6 x 10(-4), which is smaller by a factor of 11 than HDO/H2O. This indicates that fractionation of HD/H2 relative to that of HDO/H2O is not kinetically controlled by the rates of formation and destruction of H2 and HD but is thermodynamically controlled by the isotope exchange HD + H2O left and right arrow HDO + H2. Molecular hydrogen is strongly depleted in deuterium relative to water on Mars because of the very long lifetime of H2 (1200 years). The derived isotope fractionation corresponds to an estimate of a planetwide reservoir of water ice about 5 meters thick that is exchangeable with the atmosphere.

Flight Experiment↗