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Stacked target irradiations and radiochemical separations for simultaneous production of short-lived Ni and Ir isotopes
A method was developed for simultaneous production of neutron-deficient Ni and Ir isotopes by irradiation of stacked targets with a 50.5 MeV proton beam, followed by radiochemical separations to remove excess target material. The target, irradiation configuration, and post-irradiation radiochemical separations were developed and tested in a proof-of-principle irradiation. Radiochemical yields and purity of the products are discussed. Finally, the stacked target method, developed herein, can be used to produce other isotope pairs with appropriate beam tailoring.
Ozonolysis of selenium: an alternative oxidation method for radiochemical separations of arsenic and selenium
Many arsenic and selenium radiochemical separation methods use hydrogen peroxide (H 2 O 2 ) to oxidize selenite (Se(IV)) to selenate (Se(VI)), but this reaction is challenging to control and replicate. Here, this study introduces a simple ozonolysis apparatus with an in-line oxygen-ozone generator and impinger system for rapid selenium oxidation in aqueous samples as an alternative to H 2 O 2 . The system quickly converts Se(IV) to Se(VI), enabling successful separations of arsenic-73 and selenium-75 with high recoveries using ion-exchange chromatography (1-X8). Compared to traditional H 2 O2 treatment, ozone oxidation is a faster and more easily controlled method for selenium redox chemistry.
Radiochemical Separation and Quantification of Tritium in Metallic Radwastes Generated from CANDU Type NPP - 13279
As a destructive quantification method of {sup 3}H in low and intermediate level radwastes, bomb oxidation, sample oxidation, and wet oxidation methods have been introduced. These methods have some merits and demerits in the radiochemical separation of {sup 3}H radionuclides. That is, since the bomb oxidation and sample oxidation methods are techniques using heating at high temperature, the separation methods of the radionuclides are relatively simple. However, since {sup 3}H radionuclide has a property of being diffused deeply into the inside of metals, {sup 3}H which is distributed on the surface of the metals can only be extracted if the methods are applied. As an another separation method, the wet oxidation method makes {sup 3}H oxidized with an acidic solution, and extracted completely to an oxidized HTO compound. However, incomplete oxidized {sup 3}H compounds, which are produced by reactions of acidic solutions and metallic radwastes, can be released into the air. Thus, in this study, a wet oxidation method to extract and quantify the {sup 3}H radionuclide from metallic radwastes was established. In particular, a complete extraction method and complete oxidation method of incomplete chemical compounds of {sup 3}H using a Pt catalyst were studied. The radioactivity of {sup 3}H in metallic radwastes is extracted and measured using a wet oxidation method and liquid scintillation counter. Considering the surface dose rate of the sample, the appropriate size of the sample was determined and weighed, and a mixture of oxidants was added to a 200 ml round flask with 3 tubes. The flask was quickly connected to the distilling apparatus. 20 mL of 16 wt% H{sub 2}SO{sub 4} was given into the 200-ml round flask through a dropping funnel while under stirring and refluxing. After dropping, the temperature of the mixture was raised to 96 deg. C and the sample was leached and oxidized by refluxing for 3 hours. At that time, the incomplete oxidized {sup 3}H compounds were completely oxidized using the Pt catalysts and produced a stable HTO compound. After that, about a 20 ml solution was distilled in the separation apparatus, and the distillate was mixed with an ultimagold LLT as a cocktail solution. The solution in the vial was left standing for at least 24 hours. The radioactivity of {sup 3}H was counted directly using a liquid scintillation analyzer (Packard, 2500 TR/AB, Alpha and Beta Liquid Scintillation Analyzer). (authors)
Nuclear Physics Gamma-ray Imaging System for Real-Time Rare Isotope Harvesting, Monitoring and Radiochemical Separation – NP Imager Phase I Final Technical Report
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Radiochemical separations and experimental measurements of short-lived fission products from 14 MeV irradiation of depleted uranium
Here, this work focused on the experimental analysis of depleted uranium irradiated with 14 MeV neutrons to obtain fission yields for 20 fission products with half-lives ranging from 16.8 h to 30.1 years. Several analytes were identified where improvements to the available nuclear data would be beneficial. The results obtained from this experiment could be used to make improvements to cumulative fission yield data once confirmatory measurements are made. The required separations will be described along with a comparison to the reported Evaluated Nuclear Data File fission yields and other experimental measurements.
Analysis of 161 Tb by radiochemical separation and liquid scintillation counting
The determination of 161 Tb activity is problematic due to its very low fission yield, short half-life, and the complication of its gamma spectrum. At AWE, radiochemically purified 161 Tb solution was measured on a PerkinElmer 1220 Quantulus TM Liquid Scintillation Spectrometer. Since there was no 161 Tb certified standard solution available commercially, the counting efficiency was determined by the CIEMAT/NIST Efficiency Tracing method. The method was validated during a recent inter-laboratory comparison exercise involving the analysis of a uranium sample irradiated with thermal neutrons. Lastly, the measured 161 Tb result was in excellent agreement with the result using gamma spectrometry and the result obtained by Pacific Northwest National Laboratory.
Radiochemical Separation Techniques in Classroom Settings
Not Available
Radiochemical Separation and Spectrometric Analyses of Americium and Plutonium [Thesis] [Slides]
Abstract not provided.
Target and Radiochemical Separations Development for (n,2n) Cross-Section Measurements of 73As at the National Ignition Facility
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Fundamental Data Supporting Novel Radiochemical Separations
Rapid chemical separation of actinide, lanthanide, and other radioactive elements is an important pursuit in the field of radioanalytical and nuclear chemistry. Vital to the development of advanced systems to achieve such separations is the determination of elements’ distribution coefficients (Kd values) for chromatographic resins. In this study, batch contacts of 68 elements were performed with various commercially available resins, and the resulting distribution coefficients were determined by Inductively Coupled Plasma – Mass Spectrometry Analysis (ICP-MS). An evaluation of the resulting Kd data for elements on the resins was performed. Finally, this work presents prototype flowsheets for streamlined separation of radioisotopes from a variety of complex matrices.
Final Results for 99 Mo, 111 Ag, 147 Nd, 140 Ba
Seven depleted uranium targets were irradiated at the ACRR reactor in three different irradiations occurring on June 4th, 5th, and 6th, 2019. The targets were received at LANL on July 10, 2019, unpackaged, and analyzed by autoradiography and gamma-ray spectrometry. Table 1 shows a chart of the Target ID, target mass, time the irradiation occurred, and the corresponding LANL sample ID. Six of the targets were dissolved and chemically separated radiochemical analyses were performed for a suite of radionuclides on four of the targets. The final results of the radiochemical separation and analysis of 99 Mo, 111 Ag, 147 Nd and 140 Ba for depleted U samples 4416-01, 4417-01, 4418-01 and 4418-03 are presented in Table 2. Associated errors represent 1σ uncertainty.
Preparing for Harvesting Radioisotopes from FRIB
In the second quarter of this grant, work has progressed smoothly at all three collaborating institutions. We have recently completed our first experiment at the NSCL under this grant successfully, where 79Kr was collected by cryotrapping from our water target apparatus. The three PI’s, one undergraduate (Boone Marois), two graduate students (Stacy Queern and Matt Scott) and one post-doc (Aranh Pen) were assisted by Dave Morrissey at the NSCL to perform this experiment. The experiment also provided the opportunity for a collaboration meeting of the PI’s to discuss future work on this proposal. Significant progress has been made on both novel radiochemical separations technology at the University of Missouri, and validating a radiochemical separation procedure for 48V at Washington University. The only change in the work-scope of the original proposal is the transition of the Washington University PI to the University of Alabama at Birmingham.
Annual Report, Fall 2016: Identifying Cost Effective Tank Waste Characterization Approaches
This report documents the activities that were performed during the second year of a project undertaken to improve the cost effectiveness and timeliness of SRNL’s tank closure characterization practices. The activities performed during the first year of the project were previously reported in SRNL-STI-2015-00144. The scope of the second year activities was divided into the following three primary tasks: 1) develop a technical basis and strategy for improving the cost effectiveness and schedule of SRNL’s tank closure characterization program; 2) initiate the design and assembly of a new waste removal system for improving the throughput and reducing the personnel dose associated with extraction chromatography radiochemical separations; and 3) develop and perform feasibility testing of three alternative radiochemical separation protocols holding promise for improving high resource demand/time consuming tank closure sample analysis methods.
Pursuing Cost Effective Tank Waste Characterization at the Savannah River Site - 16172
The U. S. Department of Energy Office of Environmental Management has tasked the Savannah River National Laboratory (SRNL) with developing strategies and technologies to understand, optimize, scale, and speed up tank waste characterization. This scope is part of a multi-year project with the end goal of implementing programmatic changes that accelerate tank waste processing and tank closure schedules, and at the same time reduce characterization costs, while maintaining data integrity. The project is currently in its second year. During the first year of the project, SRNL completed a series of activities focused on understanding the range of current characterization practices, needs, gaps, risks, and potential alternative approaches. Based on the results, conclusions were drawn regarding aspects of the programs that are most costly and time consuming; particular waste constituents that drive costs, schedule, and program risks; alternative characterization approaches (other than sampling and analysis) that are less expensive and more rapid; new laboratory methods holding the greatest promise for reducing characterization costs and schedule; and in whole, the primary areas where current characterization practices can be improved. SRNL's current scope is focusing on three primary activities: 1) development of a technical basis/strategy for improving the cost effectiveness and schedule of SRNL's tank closure characterization program; 2) design/assembly of hardware, plumbing, and software for automating select radiochemical separation and waste removal processes; and 3) development and testing of alternative radiochemical separation protocols most likely to improve high resource demand/time consuming analysis methods (such as Ra-226, Pa-231, Tc-99/I-129, and/or Y/trivalent actinide separations). (authors)
Rescue of Unique Liquid IP-2 Portable Tanks at Savannah River Site – 25078
Westinghouse Savannah River Company, the SRS Management and Operations Contractor at the time, procured six large IP-2 portable tanks capable of transporting liquids in 2003 to support the layup of the F Canyon at SRS. F Canyon was one of two full-scale hardened radiochemical separations facilities built at SRS in the 1950s as part of the initial construction of the Site. F Canyon and H Canyon provided actinide separations capabilities for SRS through the end of the Cold War. Following the end of the Cold War era, the decision was eventually made that only one of these radiochemical separations facilities at SRS would still be needed moving forward. As H Canyon was considered better suited for ongoing SRS missions, F Canyon was placed into a layup state in the first decade of the 21st Century, with F Area deactivation completing in 2006. This process resulted in the creation of excess, radioactively contaminated solutions, which required disposition. These solutions included contaminated organic solutions as well as DU-bearing aqueous solutions. The six portable tanks were procured to support the transportation of these solutions.
Rapid Automated Dissolution and Analysis Techniques for Radionuclides in Recycle Process Streams
The analysis of process samples for radionuclide content is an important part of current procedures for material balance and accountancy in the different process streams of a recycling plant. The destructive sample analysis techniques currently available necessitate a significant amount of time. It is therefore desirable to develop new sample analysis procedures that allow for a quick turnaround time and increased sample throughput with a minimum of deviation between samples. In particular, new capabilities for rapid sample dissolution and radiochemical separation are required. Most of the radioanalytical techniques currently employed for sample analysis are based on manual laboratory procedures. Such procedures are time- and labor-intensive, and not well suited for situations in which a rapid sample analysis is required and/or large number of samples need to be analyzed. To address this issue we are currently investigating radiochemical separation methods based on extraction chromatography that have been specifically optimized for the analysis of process stream samples. The influence of potential interferences present in the process samples as well as mass loading, flow rate and resin performance is being studied. In addition, the potential to automate these procedures utilizing a robotic platform is evaluated. Initial studies have been carried out using the commercially available DGA resin. This resin shows an affinity for Am, Pu, U, and Th and is also exhibiting signs of a possible synergistic effects in the presence of iron.
Extraction of Pm-147 from savannah river site nuclear material management programs
The Savannah River Site (SRS) processes used-nuclear fuel to recover uranium. There are many interesting fission products lost in the raffinate from this separation including 147 Pm. Here, the goal of this work was to recover 147 Pm from SRS waste solutions. A radiochemical separation method typically used for the determination of 147 Pm and 151 Sm was modified to purify significant quantities of 147 Pm from the H Canyon waste solutions. The developed scheme was tested on a small aliquot of waste solution from the processing of used fuel from the High Flux Isotope Reactor at Oak Ridge National Laboratory.