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Final report for UW effort on project Automated Isolation of Astatine-211 and chemistry evaluation
Researchers at Sandia have developed an advanced radiation hydrodynamic simulation capability by coupling the ALEGRA and Sceptre codes.
Californium-252 (Cf-252) has been synthesized on a large scale (mg) since the 1970s for use as a portable neutron source. With a half-life of 2.645 years and a spontaneous fission fraction of 3.09%, Cf-252 emits 2.31 × 10 12 neutrons/gram second. Today most of the Cf-252 produced in the world is made at Oak Ridge National Laboratory (ORNL) using the High Flux Isotope Reactor (HFIR) and the Radiochemical Engineering Development Center (REDC). Because of its highly radioactive nature, all manipulations of Cf-252 need to be performed in hot cell facilities by highly trained personnel. The Cf-252 produced at ORNL is done at the REDC, which is conveniently located next to the HFIR. Cf-252 is sold as bulk wire, which has been described as palladium (Pd) metal incasing californium oxide (Cf 2 O 3 ). The process was developed by Mosley et al. at Savannah River Laboratory (SRL) where it was modeled after the thorium work done by Fuschillo et al. This process has been used since the 1970s to synthesize this unique wire, but its fundamental chemistry and materials characteristics were never openly reported. Here we aim to better understand the chemistry and the materials characteristics of Cf-252 wire production for process improvements and to better our fundamental understanding of the mid-late actinides. Though there are many challenges in working with radioactive elements such as Cf, the isotope Cf-252 is even more complicated due to its neutron emission. One challenge with neutron emission is the ability of neutrons to penetrate much deeper through materials, causing radiation damage. This makes using instrumentation to characterize the chemistry and materials properties of this process difficult. For this reason, this study will first focus on terbium (Tb) as a surrogate for Cf. Currently Tb is used as a carrier or filler in Cf production when small amounts of Cf are being processed. This, along with their similar size and charge, makes Tb a reasonable surrogate for this study.
Over the course of this two year grant, we have developed and optimized new methods for the production of clinical quality 76 Br, 77 Br, and 80m Br at the University of Wisconsin. This includes the synthesis of a novel isotopically enriched cobalt-selenium intermetallic compound, its fabrication into robust cyclotron targets, and the chemical isolation of the resulting bromine radionuclide. Compared with the previous state-of-the-art, these methods demonstrate a six-times larger 76 Br and three-times larger 77 Br production capacity for a 3-hour cyclotron irradiation. Additionally, the cyclotron targets are exceptionally robust to this process, allowing for their reuse for dozens of radiobromine productions.
In September 2020, the Laboratory’s Chemical Microscopy Facility tripled in size, expanding to 1,000 square feet. Located at TA-46, Building158, this facility houses a unique Laboratory resource. Operated by the Physical Chemistry and Applied Spectroscopy (C-PCS) group, this facility was established to renew the Laboratory’s rich history in radiobiology. The facility combines biological research capabilities with isotope production, as well as radioanalytical and chemical infrastructure. This facility provides space for Biosafety Level 2 (BSL-2) research, with alpha-emitting and other radionuclides of interest associated with biological systems. Equipped with a Class ll/A2 biological safety cabinet, a radiological fume hood and a host of other analytical equipment, this facility serves an effective resource to evaluate the efficacy of targeted alpha therapeutics, as well as the deleterious effects of radiological materials on living organisms. The facility receives funding through LDRD- 20180005DR, “Establishing a Radiotherapeutic Capability to Counter Biothreats,” and the BES Heavy Element Chemistry Program at the Laboratory.
The Rapid Sample Insertion/Extraction System for Gamma Irradiation, otherwise known as the "rabbit" system, was a four-week long project which included many different aspects; from coding an Arduino to building PVC piping to 3-D printing the "rabbit" capsules. The "rabbit" system is a system of PVC piping that allows a quick and efficient transfer of materials into/out of one of the irradiation chambers in the Gamma Irradiation Facility (GIF) with the use of a 3-D printed "rabbit." This "rabbit" encapsulates material to be irradiated and carries it from a position outside of the irradiation chamber to the basket inside of the chamber. The main purpose of this system is to save time and provide more exact data without any delays that normally occur when a person has to enter the chamber, retrieve data, and then analyze the data. This system should take measurements and retrieve the data instantaneously. The way in which the "rabbit" is sent through the PVC piping is with an advanced bi-directional, high-throughput pneumatic system, or a shop vacuum cleaner. When the vacuum is set to blow or suck then the "rabbit" will be pulled through the PVC piping to its intended destination and will hit sensors along the sides of the tubing when it reaches the end of the piping. These sensors tell the Arduino that the "rabbit" is finished moving throughout the tubing and stops a timer. Another timer is used to see how long the "rabbit" is being irradiated so when the "rabbit" reaches the sensors in the basket in the irradiation chamber another timer is started and it ends when the sensors no longer detect the "rabbit," which means that it has begun its journey back to the starting point.
This report was prepared to summarize the results of an inter- and intra-laboratory round robin of ASTM Method C1733-17 (now C1733-20a) Distribution Coefficients of Inorganic Species by the Batch Method. The testing used the sorption of cesium (Cs) to a standard reference soil from the San Joaquin Valley (Standard Reference Material 2709a). The data collected from the nine participants was used for a statistical analysis to assess the precision of the method and generate a precision and bias statement to update the C1733-20a standard.
Issue 1 highlights of the Los Alamos Physical Sciences Vistas newsletter include: Finding and aiding success through the Lab's employee scholarship fund; Volunteer vignettes: helping others throughout Northern New Mexico; X-ray topography instrument reimagined for use at Oregon State University; Potentially activated metal recycle project triples its goal; Adaptive 3D machine-learning method for 3D coherent diffraction imaging; and, Sigma team aims for excellence in safety, quality, productivity.
The soil composition used in the Scorpius MCNP6.2 (Ref. 1) model was recently recalculated. Originally, it was “determined using the United States Geological Survey (USGS) Mercury Core Library and the Nevada National Security Site U.S. Geological Survey Databases (NNSS USGS) and Nevada National Security Site Petrographic, Geochemical, and Geophysical Database (NNSS PGG),” but the details had been lost. Reference 2 documents the updated soil composition determination. Reference 2 estimated the grain density of the soil instead of the bulk density. This report corrects that error. I am indebted to Garrett Euler (EES-17) for reading Ref. 2, pointing out its errors, and guiding me through the correct calculations. This report is organized as follows. For completeness, the composition calculations from Ref. 2 are repeated: Sec. II discusses the data that are reported in the PGG Access database, and Sec. III uses the data in the PGG Access database to compute the composition of the soil. Section IV estimates the bulk density of the soil. Section V evaluates the soil wall thickness in the Scorpius model. Section VI estimates the effect of the soil density change on previously calculated results. Section VII presents a summary and conclusions. Input files and output files are listed in appendices.
Perform qualitative and quantitative assay of gamma, beta, and alpha-emitting radionuclides in a variety of matrices and over a wide range of activity levels.
Several 100 μg aliquots of NBL CRM 137 Pu sulfate tetrahydrate were prepared in 20 mL glass vials at some time in the past and are expected to have similar chemical composition and similar concentrations of contaminants as the bulk 250 mg CRM 137 Pu sulfate tetrahydrate units. Thus, these units were characterized and used to test Pu purification chemistry to be employed on the larger unit. For preliminary characterization, the residue of bottle 6 was dissolved in 4 M HNO3/0.01 M HF and transferred quantitatively to a pre-weighed 30 mL Savillex Teflon vial with a total solution mass of approximately 18 g. A gravimetric aliquot was removed from this solution for characterization and the remaining solution was dried down and reconstituted in 8 M HNO 3 to test the anion exchange purification column.
This report provides a summary of notes for building and running the Sandia Computational Engine for Particle Transport for Radiation Effects (SCEPTRE) code. SCEPTRE is a general- purpose C++ code for solving the li near Boltzmann transport equation in serial or parallel using unstructured spatial finite elements, multigroup energy treatment, and a variety of angular treatments including discrete ordinates and spherical harmonics. Either the first-order form of the Boltzmann equation or one of the second-order forms may be solved. SCEPTRE requires a small number of open-source Third Part y Libraries (TPL) to be available, and example scripts for building these TPLs are provided. The TPLs needed by SCEPTRE are Trilinos, boost, and netcdf. SCEPTRE uses an autotools build system , and a sample configure script is provided. Running the SCEPTRE code requires that the user provide a spatial finite-elements mesh in Exodus format and a cross section library in a format that will be described. SCEPTRE uses an xml-based input, and several examples will be provided.
This document includes details of the angular quadrature sets available in SCEPTRE for performing numerical integrations in the angular phase space. The angular dependence of the boundary and fixed-source terms an d initial angular flux are specified by angular index rather than by direction. It is, therefore, necessary to know the mapping from a specific direction to a direction index. This document includes angular quadrature weights and direction cosines for most of the quadrature sets available in SCEPTRE.
The production of californium-252 (Cf-252) has been of interest to the US Department of Energy (DOE) since the 1970s because of its ability to emit neutrons (2.31 x 10 12 neutrons/gram second). Today, Oak Ridge National Laboratory is responsible for the majority of Cf-252 sold around the world. This report aims to provide a complete summary of the chemical processing of the Cf and curium (Cm) from campaign 78 (C78) which started in January 2019.
The Chemical Operations Group has a stock of Reillex HPQ resin that was purchased from Reilly Industries in 1998. This report was written to give context to the currently available options for obtaining polyvinylpyridine based resin such as Reillex HPQ. The resin is no longer manufactured, and the only available options are holdings from Reilly Industries such as that at LLNL or a production run of Reillex HPQ by Vertellus in 2015. This Verellus production run was commissioned and purchased by Savannah River Site. This report and references establish LLNL’s current stock as a polyvinylpyridine based resin. Additionally, this report and references establish that the formulation of Reillex HPQ is only 70% polyvinylpyridine by design. The 30% divinylbenzene does not constitute an impurity but an integral crosslinking component of the resin itself.
Presentation outlines a summer project decontaminating gloveboxes.
Objectives: Provide experimental data to feed back in to models for high-throughput predictions of potential new scintillator materials and identify promising scintillator compounds to pursue further.
We have irradiated high-density polyethylene ribbons and cylinders to ~200 kGy at a dose rate of ~5 kGy day -1 , using a 60 Co source. The radiolytically-generated gas consists almost entirely of H 2 . Based on the time- and dose-dependence of the outgassing, we have determined the radiation- chemical yield for the evolved H 2 , as well as the diffusion coefficient of H 2 in HDPE. We find that the evolved gas yield G evolved depends on the interplay of sample geometry, experimental timescale, and the diffusion coefficient of H 2 in HDPE. When these factors facilitate completion of diffusion prior to gas quantification, such as for sub-mm-thickness ribbons, we measure a time- and dose-invariant value of G evolved = 0.0076 cm 3 std. H 2 J -1 that is in agreement with literature reports. When sample geometry and experiment timescale do not permit completion of diffusion, we measure a time- and dose-dependent value of G evolved . We have developed a model for the transient outgassing behavior of the cylinders, and fitted it to the experimental data. Based on these fits, we estimate a room-temperature diffusivity of D H 2 /HDPE ≈ 1.67 ± 0.12 x 10 -6 cm 2 s -1 for H 2 in HDPE, in good agreement with literature values. At present, the effects of dose rate and γ-photon energy on HDPE gas yield are not known and require further investigation.