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9 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

SU-F-I-56: High-Precision Gamma-Ray Analysis of Medical Isotopes

Purpose: Advanced, time-resolved, Compton-suppressed gamma-ray spectroscopy with germanium detectors is implemented for assaying medical isotopes to study the radioactive decay process leading to a more accurate appraisal of the received dose and treatment planning. Lowell’s Array for Radiological Assay (LARA), a detector array that is comprised of six Compton-suppressed high-purity germanium detectors, is currently under development at UMass-Lowell which combines Compton-suppression and time-and-angle correlations to allow for highly efficient and highly sensitive measurements. Methods: Two isotopes produced Brookhaven Linac Isotope Producer (BLIP) were investigated. {sup 82}Sr which is the parent isotope for producing {sup 82}Rb is often used in cardiac PET. {sup 82}Sr gamma-ray spectrum is dominated by the 511keV photons from positron annihilation which prevent precise measurement of co-produced contaminant isotopes. A second project was to investigate the production of platinum isotopes. Natural platinum was bombarded with protons from 53MeV to 200MeV. The resulting spectrum was complicated due to the large number of stable platinum isotopes in the target, the variety of open reaction channels (p,xn), (p,pxn), (p,axn). Results: By using face-to-face NaI(Tl) counters 90-degrees to the Compton-suppressed germaniums to detect the 511keV photons, a much cleaner and more sensitive measurement of {sup 85}Sr and other contaminants was obtained. For the platinum target, we identified the production of {sup 188–189–191–195}Pt, {sup 191–192–193–194–195–196}Au and {sup 186–188–189–190–192–194–189–190–192–194}Ir. For example, at the lower energies (53 and 65MeV), we measured {sup 191}Pt production cross-sections of 144mb and 157mb. Considerable care was needed in following the process of dissolving and diluting the samples to get consistent results. The new LARA array will help us better ascertain the absolute efficiency of the counting system and more reliable production cross-sections. Conclusion: Modern HPGe spectroscopic techniques provide enhanced sensitivity, promising precise quantification of the quality of radioisotopes used in medical physics. Using new decay information may have non-trivial impact on treatment planning and dose-assessment.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2019

Nuclear Data Sheets for A=82☆

The evaluators present in this publication spectroscopic data and level schemes from radioactive decay and nuclear reactions studies for all isobars, Z=28(Ni) to Z=42(Mo), with mass number A=82. For {sup 82}Ni not much is known other than its identification. Most of the new data, i.e., after the last NDS (2003Tu03), are based on compilations included in XUNDL. Contributions of the various compilers in XUNDL are gratefully acknowledged.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2018

Target development for {sup 67}Cu, {sup 82}Sr radionuclide production at the RIC-80 facility

A high-current cyclotron C-80 capable of producing 40–80 MeV proton beams with a current of up to 200 μA has been constructed and commissioned at PNPI (Petersburg Nuclear Physics Institute). One of the main goals of cyclotron C-80 is the production of a wide spectrum of medical radionuclides for diagnostics and therapy. To date, the project development of a radioisotope facility RIC-80 (radioisotopes at cyclotron C-80) has been completed. The feature of the project is the use of a mass-separator combined with the ion-target device for obtaining ion beams of radioisotopes with a high purity of separation that is especially important for medical applications. The first results of а new high-temperature method for extracting {sup 82}Sr and {sup 67}Cu radioisotopes from irradiated targets have been presented.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2015

The radioisotope complex project “RIC-80” at the Petersburg Nuclear Physics Institute

The high current cyclotron C-80 capable of producing 40-80 MeV proton beams with a current of up to 200 μA has been constructed at Petersburg Nuclear Physics Institute. One of the main goals of the C-80 is the production of a wide spectrum of medical radionuclides for diagnostics and therapy. The project development of the radioisotope complex RIC-80 (radioisotopes at the cyclotron C-80) at the beam of C-80 has been completed. The RIC-80 complex is briefly discussed in this paper. The combination of the mass-separator with the target-ion source device, available at one of the new target stations for on-line or semi on-line production of a high purity separated radioisotopes, is explored in greater detail. The results of target and ion source tests for a mass-separator method for the production of high purity radioisotopes {sup 82}Sr and {sup 223,224}Ra are also presented.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2018

A New Method for Production of the Sr-82 Generator Radionuclide and Other Medical Radionuclides

A cyclotron (C-80) designed to produce protons with an energy of 40–80 MeV and beam intensity of 100–200 μA has been launched at the Petersburg Nuclear Physics Institute, National Research Center Kurchatov Institute. The main task of C-80 is the production of a wide range of medical radionuclides for diagnosis and therapy. A complex named Radioactive Isotopes on the C-80 Cyclotron (RIC-80) and intended to function on the C-80 beam was developed for this purpose. A brief description of the RIC-80 complex is given, and the results of the use of new methods and studies of target devices for the production of the generator radioisotope {sup 82}Sr; {sup 223}Ra, {sup 224}Ra, and {sup 225}Ac radioisotopes that undergo alpha-decay; {sup 67}Cu; and other medical radionuclides are presented. The use of a mass separator in the “on-line” mode to obtain ion beams of high-purity radioisotopes is a distinctive feature of the complex used in the project, and it is especially important for medical applications.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2022

Inorganic resins enable the increased purification efficiency of 82 Sr from rubidium targets for use in PET imaging isotope production

The 82 Sr/ 82 Rb generator system has found use in positron emission tomography (PET) imaging of myocardial perfusion. 82 Sr is produced by proton irradiation of Rb metal/RbCl salt followed by chemical separation. Inorganic ion-exchangers were synthesized by hydrothermal methods and evaluated for the purification of accelerator-produced 82Sr from a RbCl target and co-produced impurities like 75 Se, Fe, Cr, and Ni. A process consisting of removal of bulk RbCl using sodium nonatitanate (SNT), then removal of remaining tracers of Rb and Se using poorly crystalline silicotitanate (pCST) is suitable for 82 Sr separation and achieves higher purity of 82 Sr vs. commonly used organic resins.

43 PARTICLE ACCELERATORS↗
DOE OSTI2021

Effect of additives on carbon dioxide uptake and compressive strength of dry-cast concrete

The main thrust of the work reported herein is to study the effect of chemical additives to make beneficial use of carbon dioxide in enhancing the compressive strength of dry-cast concrete. The additives used in this investigation were calcium and magnesium oxides and silicates, nickel oxide, sodium hydroxide, sodium bicarbonate and strontium chloride. The experimental results showed that both the carbonation reactions and the compressive strength of dry-cast concrete mixtures were improved when using chemical additives. Experimental investigations for optimisation were also undertaken with selected additives to further increase the carbon dioxide uptake and improve the compressive strength. The best results were obtained when using calcium silicate in combination with calcium oxide. The early-age compressive strength of specimens made from the optimised mixture increased from 9·82 MPa to 13·32 MPa, and the specimens’ carbon dioxide uptake increased from 8·76% to 10·62% of cement weight.

Construction & Building Technology↗
DOE OSTI2020

Disulfonamide Ligands as f-Element Extractants from Alkaline High-Level Waste

Almost 100 million gallons of high level waste (HLW) has been generated from defense reprocessing programs to support nuclear weapons production. HLW is currently stored at the Hanford and Savannah River Sites (SRS) [1]. It contains radioactive components, such as {sup 137}Cs, {sup 99}Tc and {sup 90}Sr, as well as large amounts of non-radioactive species, including solvated cations and insoluble metal hydroxides. HLW has high concentration of soluble hydroxides ([OH{sup -}] = 2 M at SRS [2]), complexing inorganic anions ([NO{sub 3}{sup -}] ∼ 0.65 - 3.7 M at SRS [2]) and a high ionic strength. Even though the majority of actinide component in alkaline HLW is precipitated, studies of An(III) and Ln(III) complexation in highly alkaline solutions in the presence of high nitrate concentrations have showed that soluble nitrate complexes can be formed [3]. The current treatment of HLW at SRS currently consists of two processes: The Actinide Removal Process (ARP) and the Next-Generation Caustic-Side Solvent Extraction process (NG-CSSX). Strontium and actinides are removed by sorption on monosodium titanate through the Actinide Removal Process (ARP) commonly referred to as the alpha-strike step [4]. Then, cesium is selectively extracted from the alkaline media via the Caustic-Side Solvent Extraction (CSSX) process [5]. The CSSX solvent consists of a calix[4]arene-crown-6 extractant dissolved in an inert hydrocarbon matrix containing i) a solvent modifier (alkylphenoxy alcohol) which increases extractant solubility and prevents third phase formation, and ii) a suppressor (trioctylamine or guanidine) - that mitigates surfactant effects. The CSSX process removes {sup 137}Cs selectively and rapidly, yet the post- CSSX aqueous stream may still contain high amounts of An, thus requiring a second ARP treatment (referred to as 'alpha-finishing') for some tanks. Despite the success of the ARP process in removing Sr and An, it often represents the kinetic bottleneck of integrated processing, as it is slower than solvent extraction. Understanding the complexation of actinides by organic ligands that are compatible with the CSSX process could eventually lead to a combined caustic-side Cs/Sr/actinide extraction process with better economics due to a reduced amount of monosodium titanate and/or a shorter required contact time with titanate during ARP. This would ensure a low-activity waste (LAW) stream with no actinides without additional ARP processing post-CSSX. Disulfonamides were studied for extraction of Sm(III) from alkaline aqueous media of pH 10-14 into dichloromethane. Up to 82% of Sm(III) was extracted from solutions of pH 12.5 -13.5 and up to 84% from solutions of pH 10.5 - 11.5. These results show some resemblance with Am extraction results by calixarene ligands previously reported [10]. Kinetic studies demonstrated that even 5 min is enough to complete stripping, whereas extraction is time-limiting process and requires up to 20 h for efficient removal of Sm(III) from alkaline aqueous media. Determination of composition of complexes in solution after extraction by the equilibrium shift method showed a 1:1 Sm(III):dsa-2 complexation ratio for the extracted species.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

GeoMelt{sup R} In-Container Vitrification (ICV){sup TM} for Fukushima Daiichi Water Treatment Secondary Wastes - 20212

The Japanese government is supporting development work implemented jointly by Veolia subsidiaries Kurion Japan, K.K., Veolia Nuclear Solutions (VNS), Inc., and Veolia Nuclear Solutions Federal Services, LLC for treating radioactive waste generated from Fukushima Daiichi Nuclear Power Station (NPS) water treatment using the GeoMelt{sup TM} In-Container Vitrification (ICV){sup TM} technology. The initial work consisted of glass formulation and engineering-scale testing which was completed in 2018, in the frame of an IRID (International Research Institute for Nuclear Decommissioning) program as part of a project subsidized by Japan's Ministry of Economy, Trade and Industry (METI). The Fukushima Daiichi NPS Mid- and Long-Term Road-map requires investigation of methods to stabilize solid wastes (and to immobilize radioisotopes in the wastes) generated as a result of emergency response and decommissioning activities. Cooling water treatment has resulted in a significant amount of solid and slurry secondary wastes (mostly adsorbents and ion-exchange materials) which will require processing at some point. GeoMelt{sup R} ICV{sup TM} is a joule-heated melter technology which uses a refractory-lined single-use container combining the melter and disposal container. There is no pouring required nor concerns with refractory corrosion which allows the process to accommodate a wide range of waste chemistries and high waste loadings. The testing described here consisted of three engineering-scale melts, each processing between 212 kg and 240 kg of waste simulants, glass formers, and non-radioactive cesium (Cs) and strontium (Sr) tracers. Continuous isokinetic stack sampling of off-gas emissions was performed for each test in order to calculate Cs and Sr retention in the glass wasteform. Single-pass retention of Cs in the final glass wasteform ranged from 91.46 to 99.30%, and single-pass retention of Sr ranged from 99.76 to 100%. Planned particulate recycle will increase these retention levels. Melt 1 processed a mixture of KUR-EH (a zeolite-based ion-exchange material), simulated Advanced Liquid Processing System (ALPS) Carbonate and Iron Slurries, and glass additives. Melt 2 processed a mixture of KUR-EH, KUR-TSG (a titanate-based adsorbent), and glass additives. Melt 3 processed a mixture of KUR-EH, simulated barium sulfate/iron ferrocyanide sludge (AREVA sludge), and glass additives. Waste loadings for these melts ranged from 70 weight percent (wt%) to 82 wt%. Vitrification produces a waste form much denser than the stored water treatment secondary waste wastes, resulting in significant volume reduction. Volume reductions for the three tests ranged from 74 to 79 vol%. Vitrification produces a chemically durable wasteform. Pacific Northwest National Laboratory (PNNL) tested three glass samples from each engineering-scale melt) by the Materials Characterization Center 1 (MCC-1) test, an international standard leach test of the chemical durability of nuclear waste glasses. PNNL also obtained one U.S. reference glass (EA Glass) and two Japan reference glasses P0798) and tested these under the same MCC-1 conditions (90 deg. C, 10 m-1, DIW, and 7, 14, 28-day) as the GeoMelt{sup R} ICV{sup TM} glasses. The GeoMelt{sup R} ICV{sup TM} glasses exhibited lower total normalized releases and 14- to 28-day normalized release rates than the three reference glasses. These results suggest that the GeoMelt{sup R} ICV{sup TM} glasses have durabilities on par with high-level waste glasses under standard test conditions. Post-melt process sampling and analysis indicated no Cs migration into the melter refractory materials and very little deposition of Cs or Sr onto the melter hood or off-gas piping internals. The results of the testing indicated good Cs retention in the glass, high volume reduction and waste loadings, and excellent chemical durability. These factors are important to minimize treatment costs and to protect workers and the environment. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗