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At least 109 records · Page 6
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Rapid separation of radiopure yttrium-91 for tracer studies from mixed fission products and uranium

A novel, single step method for isolating 91 Y from irradiated uranium and mixed fission products has been developed based on the commercially available Eichrom LN resin. The separation procedure allows for loading an LN resin column with an irradiated uranium solution, containing mixed fission products, in dilute HCl where both uranium and yttrium are retained on the resin. Eluting dilute (0.5–1 M) HCl will strip the majority of fission products. Furthermore, increasing the concentration to 3 M HCl will elute 91 Y without the presence of any other fission products. Finally, uranium can be recovered by passing concentrated HCl through the column.

Chromatography↗
DOE OSTI2022

RAPID UNSEPARATED RARE EARTH ELEMENT ANALYSES BY ISOTOPE DILUTION MULTICOLLECTOR INDUCTIVELY COUPLED MASS SPECTROMETRY (ID-MC-ICP-MS)

We present a method for analyzing rare earth elements (REEs) by isotope dilution without separation of individual elements. Utilizing a Nu Plasma 2 multicollector inductively coupled plasma mass spectrometer (MC-ICP-MS) equipped with an Aridus II desolvation nebulizer, sample introduction tuning parameters (torch position, gas flows) were investigated and used to minimize oxide formation (e.g., NdO/Nd <0.05%) and remove the need to analyze the REEs as separated “cuts.” Ratios for spike/sample isotopes in mixed synthetic standards containing varying amounts of Ce, Nd, Sm, Eu, Gd, Dy, Er, and Yb were measured accurately within 1% of IUPAC values. La and Lu in the standards were more than 2% off from expected values due to more demanding interference corrections. A USGS BCR-2 rock standard was also processed with flux fusion, iron coprecipitation and a single column pass bulk separation followed by analysis. The concentrations for Ce, Nd, Sm, Eu, Dy, Er, and Yb fell within 2% of consensus values. A United States Geological Survey (USGS) BIR-1A rock standard was also processed in triplicate by hydrofluoric and nitric acid digestion, yielding rare earth concentrations within 3% of consensus values with 0.3% relative standard deviation (n=3). This demonstrates the ability to produce high quality REE patterns of comparable quality to traditional methods involving labor and time intensive REE separation, with sample preparation and analytical times that are comparable to analyses against isotope dilution by single-collector ICP-MS.

Shen, Steve D.↗
DOE OSTI2020

Measurement and microscopic description of odd–even staggering of charge radii of exotic copper isotopes

Nuclear charge radii globally scale with atomic mass number A as A 1/3 , and isotopes with an odd number of neutrons are usually slightly smaller in size than their even-neutron neighbours. This odd–even staggering, ubiquitous throughout the nuclear landscape1, varies with the number of protons and neutrons, and poses a substantial challenge for nuclear theory. Here, we report measurements of the charge radii of short-lived copper isotopes up to the very exotic 78 Cu (with proton number Z = 29 and neutron number N = 49), produced at only 20 ions s –1 , using the collinear resonance ionization spectroscopy method at the Isotope Mass Separator On-Line Device facility (ISOLDE) at CERN. We observe an unexpected reduction in the odd–even staggering for isotopes approaching the N = 50 shell gap. To describe the data, we applied models based on nuclear density functional theory and A-body valence-space in-medium similarity renormalization group theory. Through these comparisons, we demonstrate a relation between the global behaviour of charge radii and the saturation density of nuclear matter, and show that the local charge radii variations, which reflect the many-body polarization effects, naturally emerge from A-body calculations fitted to properties of A ≤ 4 nuclei.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2020

Optimal methods for preparation, separation, and determination of radium isotopes in environmental and biological samples

In recent years, radium has attracted considerable attention primarily because of the rapid increase in unconventional (fracking) drilling technology in the United States and around the world. One of the major radionuclides of interest in unconventional drilling wastes is radium isotopes ( 224 Ra, 226 Ra, 228 Ra). To access long-term risks associated with radium isotopes entering into the environment, accurate measurements of radium isotopes in environmental and biological samples are crucial. This article reviews many aspects of radium chemistry, which includes recent developments in radiochemical separations methods, advancements in analytical techniques followed by a more detailed discussion on the recent trends in radium determination.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2025

The FRIB Decay Station: New Horizons with Rare Isotopes

In May 2022, the Facility for Rare Isotope Beams (FRIB), located on the campus of Michigan State University (MSU), began delivering exotic isotopes to an international community of scientists. New discoveries are now being reported from radioactive decay of neutron-rich nuclei near N = 20 and N = 28.FRIB is expected to produce roughly 80% of the unstable or radioactive isotopes predicted to exist up to uranium (Z = 92). The new user facility is supported by the U.S. Department of Energy, and it is operated by MSU. A high-power superconducting linear accelerator, shaped like a paper-clip, drives the production of these rare isotopes by colliding stable nuclei moving at half the speed of light with a rotating, water-cooled graphite tar-get. These collisions cause the primary stable beam to fragment into a wide variety of unstable nuclei, which can be subsequently filtered through a multistage magnetic separator, the Advanced Rare Isotope Separator, and transported to one of several experimental stations. The FRIB Decay Station initiator (FDSi) (see Figure 1) was developed to enable comprehensive radio-active decay studies of the exotic nuclei produced by FRIB and it was used in the first two experiments in 2022. Further, the FDSi is a highly reconfigurable multidetector system with two focal planes (FP1 for discrete spectroscopy and FP2 for total absorption spectroscopy) that can be optimized for the specific science goals of each experiment. It is designed, built, and operated by a community of users with the sup-port of U.S. funding agencies, including the Department of Energy and National Science Foundation.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2024

Production of High Specific Activity 72 Se/ 72 As, 117m Sn and 203 Pb for Research and Clinical Applications: Effective target design, recycling of target material and radioisotope separation

This grant involved four objectives for producing high specific activity radionuclides using reactor and accelerator technologies that would find use in medical, industrial and research applications. This grant continued a collaboration between the scientists at the University of Washington (UW), the University of Missouri (MU) and Brookhaven National Laboratory (BNL) and addresses the need for high specific activity radionuclides, both reactor- and accelerator-produced, and for the training and education of radiochemists in isotope production, separations and precursor syntheses. All three institutions were involved in the projects, but to different degrees. The chemistry graduate students (Ph.D. program at MU) may have participated in an internship at one of the other institutions for training and project translation/facilitation.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2020

Removal of High Specific Activity Fission Products from Uranyl Sulfate Waste Solutions

The Savannah River National Laboratory (SRNL) is currently providing support to SHINE Medical Technologies (SHINE) which plans to deploy a low energy, accelerator-based neutron source to fission low enriched U in a uranyl sulfate target solution for 99 Mo production. The 99 Mo is initially separated from the fission products and target solution by an extraction column. Subsequent washing of the column will generate waste solutions containing residual U and fission product activity. A small number of high specific activity fission products (e.g., 90 Sr, 137 Cs, and 144 Ce) in these streams will likely control the classification of the low level waste (LLW). If a sufficient amount of the high specific activity isotopes are separated from the SHINE waste streams and concentrated into a waste form, it would be possible to treat a majority of the wash solutions from the column operations as a lower class of LLW (Class A versus Class B or C or Class B versus Class C). The high specific activity fission product elements could then be dispositioned as a much smaller volume of waste rather than requiring the disposal of the entire waste stream at the higher waste classification. The Savannah River Site (SRS) has experience with using monosodium titanate (MST) and crystalline silicotitanate (CST) to remove Cs and Sr from high salt content waste solutions generated during the reprocessing of nuclear fuels and targets. Both of these materials have worked very well for their intended purposes at the SRS where the fission product elements are removed from highly alkaline waste. On the other hand, SHINE waste streams from the extraction column contain H 2 SO 4 which makes the solution acidic. Additionally, the SRS waste streams do not contain other fission product elements such as transition metals or lanthanides because they precipitate upon neutralization of the SRS waste and are not present in the supernate which is dispositioned as LLW following treatment. As such, there are inherent differences between SHINE and SRS waste treatment strategies. Savannah River National Laboratory was tasked with performing scoping studies to see if MST and CST would remove Sr, Cs, and Ce from an acidic mixed metal simulant solution. Batch contact experiments were performed using MST and two CST type materials. The MST material is a 15 wt % powder in 0.15 M NaOH slurry. The MST showed low adsorption for elements of interest from acidic solution. Furthermore, the powder size makes MST non-ideal for column operations. A CST IE-911 ion exchange material had high Cs adsorption, moderate Sr, and marginal Ce adsorption. Based on adsorption of all species, the ion exchange capacity was found to be 0.032 meq/mL. A bench-top column experiment to measure elemental breakthrough curves was performed using CST IE-911 where chromatographic separations of the mixed simulant were expected to occur. While most elements behaved as expected, the lanthanide series, containing Ce, broke though the column earlier than expected. The second CST material, CST R9120, displayed high adsorption for all elements in the acidic mixed simulant solutions in a batch contact study, and had a calculated loading capacity of 0.091 meq/mL. Future studies to develop a waste treatment flowsheet should focus on CST R9120 to treat SHINE waste solutions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

An Automated Ultraclean Ion Exchange Separation Method for the Determinations of 232Th and 238U in Copper using Inductively Coupled Plasma Mass Spectrometry

This work presents a novel automated analytical method developed for high throughput ultrasensitive determinations of Th and U in copper using inductively coupled plasma mass spectrometry (ICP-MS). The method is based on the use of ultra clean sample preparation procedures including the use of a fully automated off-line chromatography system for the extraction and pre-concentration of the analytes prior to ICP-MS analysis. The separation system is equipped with a single reusable chromatographic column that, with the rigorously clean procedures developed herein, provided a low carryover, low background, and reproducible automated separation method. Isotope dilution methods were used for quantitation. Method Detection Limits (MDLs) of 3.7 and 9.4 fg·g-1 were obtained for the quantitation of 232Th and 238U in copper, respectively, corresponding to activities of 0.0148 and 0.116 microBq·kg-1 for 232Th and 238U, respectively. The analytical method provided high and reproducible tracer recoveries. The use of an automated system significantly reduced the extremely tedious active work time required for the chemist (by ca. 80%) relative to manually performing the separation. This method could be adapted to other critical sample matrices requiring the utmost in high throughput clean chemistry procedures for automated ultrasensitive analyses.

Arnquist, Isaac J.↗
DOE OSTI2025

Preliminary Insights Into the Feasibility of Determining the Purification Date of Enriched Uranium by Direct Measurement of the 230 Th/ 234 U Ratio Using an All-Faraday Detector Configuration on the Neoma MC-ICP-MS

Rationale: Mass spectrometric measurement of the 230 Th/ 234 U ratio to calculate the purification age of enriched uranium is typically conducted via a combination of ion counters and faraday detectors, thus requiring an inter-detector calibration scheme. Here, our aim is to understand whether the pursuit of a simplified measurement scheme involving only faraday detectors is feasible. Methods: We investigate the possibility of determining U-Th model ages for two enriched uranium standards (NBL U630 and U850) by direct measurement of the 230 Th/ 234 U ratio (without chromatographic separation or isotope dilution) on a ThermoFisher Scientific Neoma MC-ICP-MS utilizing both solution and laser ablation (LA)-based sampling techniques and an all-faraday detector configuration. Results: For the solution mode analyses conducted on aliquots containing sub μg/mL total U, we produce composite average 230 Th/ 234 U model dates of May 19, 1988 (± 351 days), and March 26, 1961 (± 2.5 years) using the directly measured 230 Th/ 234 U ratios for the NBL U630 and U850 uranium standards, which have certified purification dates of June 6, 1988 (± 190 days), and December 31, 1957 (± 36.5 days), respectively. The ages produced by LA-based sampling of dried residues of the same standards deposited onto cotton TexWipes are less accurate and of poorer precision (June 23, 2004 ± 8.7 years for U630 and December 21, 1965 ± 7.9 years for U850) but still yield meaningful information in regards to the purification date. Conclusions: We believe that further refinement of the all faraday detector measurement approach to include development of a more robust Th/U relative sensitivity factor determination, signal cutoff selection, and data processing protocols will allow for this approach to be confidently applied to enriched uranium materials with unknown purification histories. Potential advantages of the method include the reduced sample handling and infrastructure requirements as well as the ability to simultaneously generate a broad picture of the uranium isotopic composition in tandem with the U-Th age determination.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Detailed in situ ATR-FTIR spectroscopy study of the early stages of C-S-H formation during hydration of monoclinic C3S

The early stages of tricalcium silicate hydration has been studied by in situ ATR-FTIR spectroscopy. A method for data reduction based on background correction and subtraction of spectra of pure components is presented, which makes it possible to describe the kinetic evolution of the calcium silicate hydrate phase upon hydration with great detail. The results obtained are in full agreement with recently reported, detailed solid-state NMR spectroscopy measurements. As ATR-FTIR spectroscopy does not require any isotope enrichment in order to increase the signal intensities and is a widely available technique, we suggest that our study opens up an avenue towards detailed in situ characterization of hydrating cementitious systems and hydrating systems in general.

36 MATERIALS SCIENCE↗
DOE OSTI2026

A Model of Hydrogen Solubility in Palladium-Silver Alloys

Fusion fuel cycle designs depend on palladium and palladium-silver alloys. The selective permeability of these is used to separate hydrogen isotopes from other elements. Predicting the solubility of hydrogen isotopes in palladium-silver alloys is important for the design of key unit operations for tritium processing. This paper presents a model of hydrogen isotope solubility in palladium and palladium-silver alloys based on Guggenheim’s quasichemical lattice theory. The model parameters were determined as functions of isotope molecular weight and the silver content of the alloy. The model predicts reasonable values for the solubility of protium, deuterium, and tritium over a wide range of temperatures, pressures, and alloy compositions.

08 HYDROGEN↗
DOE OSTI2025

Anode sensitivity calibration of photomultiplier tubes used as a readout of the ELOSS detector

We present and discuss the methodology and results for an anode sensitivity calibration of Photomultiplier Tubes (PMTs) that serve as the optical readout of the Energy Loss Optical Scintillation System (ELOSS). The ELOSS detector, under development at the Facility for Rare Isotope Beams (FRIB), is a gaseous Xe scintillation-based detector designed to identify the atomic number (Z) of nuclear reaction fragments by energy-loss measurement. Variation in the anode sensitivity from tube to tube of commercially available PMTs limits ELOSS from reaching its ultimate Z resolution. A methodical approach to calibrating PMT anode sensitivities to a single target value by systematic PMT gain corrections is presented. Although the calibration process is discussed in terms of its impact on ELOSS performance, the presented method is applicable to any optical detector system composed of a large number of PMTs needing anode sensitivity calibration.

Instrumentation for heavy-ion accelerators↗
DOE OSTI2026

Emissivity measurements of CuCrZr alloy

The Facility for Rare Isotope Beams (FRIB) heavy-ion accelerator, in user operation since 2022, produces rare isotope beams via interactions of high-intensity stable ion beams with a graphite production target. Approximately 20–40% of the primary beam power is deposited in the target, while the remaining 60–80% is absorbed by the beam dump. The minichannel beam dump (MCBD), currently operated at 20 kW and designed for operation up to 50 kW, uses CuCrZr alloy absorber plates. Thermal validation and thermal cycling tests of the MCBD were conducted at the Applied Research Laboratory (ARL) at Pennsylvania State University. Temperature measurements were obtained from an infrared (IR) camera. Since accurate temperature determination requires reliable emissivity values, the emissivity of CuCrZr was measured using the IR camera validated against thermocouple reference temperatures up to approximately 650 °C. The measurements were conducted under a vacuum level of approximately 10 -5 torr to minimize emissivity variations due to surface oxidation. The emissivity of CuCrZr was determined to be 0.057 ± 0.009 using a constant fit to the measured data over the surface temperature range from 100–650 °C.

Accelerator Subsystems and Technologies↗
DOE OSTI2026

Physical thickness characterization of the FRIB production targets

The FRIB heavy-ion accelerator, commissioned in 2022, is a leading facility for producing rare isotope beams (RIBs) and exploring nuclei beyond the limits of stability. These RIBs are produced via reactions between stable primary beams and a graphite target. Approximately 20–40% of the primary beam power is deposited in the target, requiring efficient thermal dissipation. Currently, FRIB operates with a primary beam power of up to 20 kW. To enhance thermal dissipation efficiency, a single-slice rotating graphite target with a diameter of approximately 30 cm is employed. The effective target region is a 1 cm-wide outer rim of the graphite disc. To achieve high RIB production rates, the areal thickness variation must be constrained within 2%. This paper presents physical thickness characterizations of FRIB production targets with various nominal thicknesses, measured using a custom-built non-contact thickness measurement apparatus.

Instrumentation for radioactive beams (fragmentati↗
DOE OSTI2021

2021 Microbial U (Summary Report)

The US nuclear industry is on the cusp of significant growth in the areas of advanced reactors, SMRs, and microreactors. The existing commercial nuclear reactor fleet is considering expanding the life of the reactor cores by slightly increasing enrichment of their fuel. The common thread is the demand for High Assay Low Enriched Uranium (HALEU). A technical challenge exists with a blend down of Highly Enriched Uranium (HEU) from spent nuclear fuel (SNF) to meet the American Society for Testing and Materials (ASTM) and the “Y12” specifications. U-236 impacts reactor performance and is not desired in fresh fuel cores. In order to meet tight specifications, U-236 must be diluted by addition of fresh HEU. Fresh HEU is highly sought after and not plentiful in the US. A solution to this issue lies in the application of a technology to selectively remove the U-236 with non-hazardous microorganisms, which will be evaluated in this project by experimentally separating different isotopes of uranium.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Process for the separation and purification of medical isotopes

The invention provides a method for isolating medical isotopes, the method having the steps of dissolving titanium nuclear targets to create a solution; contacting the solution with a resin so as to retain the isotopes on the resin and generate an eluent containing titanium; contacting the isotope-containing resin with acid of a first concentration to remove impurities from the resin; and contacting the isotope-containing resin with an acid of a second concentration to remove isotope from the resin.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2020

Management of Fissionable Equivalent Mass Materials at ORNL - 20146

The US DOE manages an inventory of materials that contains a range of long-lived radioactive isotopes that were produced from the 1960's through the 1980's by irradiating targets in production reactors to produce special heavy isotopes for US DOE programmatic use, scientific research, and industrial and medical applications. Since the production reactors and enrichment facilities that produced many of these materials have been shut down, they are considered unique materials that are not likely to be produced again. ORNL uses these materials in US DoE's center for production, storage, and distribution of TRU isotopes (plutonium through californium) and other related nuclear research programs. ORNL also operates the High Flux Isotope Reactor, which provides a high neutron source for production of isotopes for medical, industrial, and nuclear research programs. As a result, ORNL has an inventory of radioisotopes that are being managed for ongoing research programs and being held for reuse because they have potential intrinsic value to US DOE. An initiative is underway at ORNL to better manage these materials, particularly focusing on those with high fissionable equivalent mass that could impact the ability to perform ongoing or new research projects. This paper describes the actions ORNL is taking to manage these inventories, many of which consist of TRU materials, through reuse and disposal. (authors)

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

Observation of New Isotopes in the Fragmentation of Pt 198 at FRIB

Five previously unknown isotopes ( 182,183 Tm, 186,187 Yb, 190 Lu) were produced, separated, and identified for the first time at the Facility for Rare Isotope Beams (FRIB) using the Advanced Rare Isotope Separator (ARIS). The new isotopes were formed through the interaction of a 198 Pt beam with a carbon target at an energy of 186 MeV/u and with a primary beam power of 1.5 kW. Event-by-event particle identification of A, Z, and q for the reaction products was performed by combining measurements of the energy loss, time of flight, magnetic rigidity Bρ, and total kinetic energy. The ARIS separator has a novel two-stage design with high resolving power to strongly suppress contaminant beams. Finally, this successful new isotope search was performed less than one year after FRIB operations began and demonstrates the discovery potential of the facility which will ultimately provide 400 kW of primary beam power.

07 ISOTOPE AND RADIATION SOURCES↗