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This progress report (Level 4 Milestone Number M4SF-22LL010302052) summarizes research conducted at Lawrence Livermore National Laboratory (LLNL) within the Crystalline Activity Number SF-22LL01030205. The research is focused on actinide and radionuclide sequestration in steel corrosion products.
We report the promising characteristics of the 7.2-h radiohalogen 211 At have long been recognized; including having chemical properties suitable for labeling targeting vectors ranging from small organic molecules to proteins, and the emission of only one α-particle per decay, providing greater control over off-target effects. Unfortunately, the impact of 211 At within the targeted α-particle therapy domain has been constrained by its limited availability. Paradoxically, the most commonly used production method – via the 209 Bi(α,2n) 211 At reaction – utilizes a widely available natural material (bismuth) as the target and straightforward cyclotron irradiation methodology. On the other hand, the most significant impediment to widespread 211 At availability is the need for an accelerator capable of generating ≥28 MeV α-particles with sufficient beam intensities to make clinically relevant levels of 211 At. In this review, current methodologies for the production and purification of 211 At – both by the direct production route noted above and via a 211 Rn generator system – will be discussed. The capabilities of cyclotrons that currently produce 211 At will be summarized and the characteristics of other accelerators that could be utilized for this purpose will be described. Finally, the logistics of networks, both academic and commercial, for facilitating 211 At distribution to locations remote from production sites will be addressed.
The use of accident tolerant fuel (ATF) cladding can increase coping times during and beyond design basis accidents. While such gains may be incremental, they provide a margin that can potentially be recovered to enable high-burnup (HBU) operation. Realizing such a margin requires demonstrating that the combination of HBU and ATF has not led to an overall increase in source term. This study investigates the influence of cladding technology (Zr-based, Cr-coated Zr, and FeCrAl) and fuel cycle length (18 and 24 months) on radiological dose at the boundary of the exclusion zone for a four-loop pressurized water reactor to investigate whether ATF claddings can provide such benefits. We analyze a recovered large break loss-of-coolant accident scenario to investigate the impact of transient timescale on the benefits of such coping time increases. The simulations have been performed using the MELCOR and MELCOR Accident Consequence Code System codes. For the cases analyzed, increased fuel cycle length did not necessarily increase radionuclide release and hydrogen generation, as these were found to be sensitive to the core power distribution. Similarly, off-site dose consequence is dominated by short-lived radionuclides that tend to saturate earlier in the burnup, so higher burnup operation did not necessarily increase the source term for the phenomena and transients analyzed here. Delays in recovery of the lowpressure safety injection system increase hydrogen production and radionuclide release, especially between 780 s and 1620 s, due to the nonlinear oxidation and core degradation behavior. Results show that Cr-coated Zr enhances safety by delaying heatup and gap release. Here, when uncertainty propagation on oxidation properties is considered, FeCrAl exhibits the lowest overall radionuclide release and off-site dose throughout the spectrum. However, while the considered “base model” performance is superior under delayed injection scenarios, upper-bound cases display hydrogen generation risk comparable to the Zr-based cladding.
To inform development of separation and generator technology relevant to the accelerator (and generator) production of 43Sc, 44Sc (from 44Ti), 45Ti, and 47Sc with fundamental, experimental exploration of the speciation and coordination chemistry of Ti and Sc.
Over 50 million nuclear medicine procedures are performed annually, leading to a multi-billion dollar market for radioisotope production. The demand for new medical and research isotopes is growing, and radioisotope supply is insufficient. Most radioisotope production today utilizes charged particle or low-energy neutron irradiation of a target. Isotope production with tens to hundred MeV incident energies is a relatively unexplored option. There is a tremendous opportunity associated with a growing number of suitable domestic and international facilities buttressed by hundred million dollar global investments (e.g., the Los Alamos and Brookhaven Isotope Production Facilities, the European Spallation Source in Lund, and the Korean Multi-purpose Accelerator Complex in Gyeongbuk). In part due to a lack of supporting nuclear data that would make modeling radioisotope yields and purities possible, these facilities do not utilize their high-energy neutron fluxes for isotope production. This project attempted to measure neutron reaction excitation functions relevant to the large-scale production of critical radioisotopes, enabling development of cost-efficient isotope production methods, contributing to the improvement of theoretical models, and enhancing the value of national isotope production facilities. Reactions which form 67 Cu, 32 Si, and alpha-emitting isotopes like 225 Ac were chosen for their consistent prioritization by expert panels, representation of diverse reaction mechanisms, and relative lack of supporting nuclear data. Accurate measurement of these data is presently made only using quasi-monoenergetic neutron beams, which are produced by bombarding thin lithium targets with protons at only a few laboratories in the world. Ultimately, the projects efforts were incompletely accomplished due to a combination of multi-year technical issues at the only laboratory in the world which can make the proposed measurements, iThemba Labs in South Africa, and the onset of a global pandemic, which precluded the significant effort and time required to develop an independent measurement capability in the United States.
This work reports an effective and scalable radiochemical separation process for isolating terbium from Gd 2 O 3 . The separation process uses three commercially available extraction chromatography resin columns, has been implemented on a computer-controlled chemistry module, and tested with 100 mg quantities of proton-irradiated nat Gd 2 O 3 . Here, the 4-hour separation procedure isolated radioterbium in 1.3 mL of 0.01 M HCl with 80 ±8% radiochemical yield and a Gd decontamination factor >(1.2 ± 0.3)·10 5 .
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Final Technical Report for DE-SC0023012
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Background: Radionuclides emitting Auger electrons (AEs) with low (0.02–50 keV) energy, short (0.0007–40 µm) range, and high (1–10 keV/µm) linear energy transfer may have an important role in the targeted radionuclide therapy of metastatic and disseminated disease. Erbium-165 is a pure AE-emitting radionuclide that is chemically matched to clinical therapeutic radionuclide 177Lu, making it a useful tool for fundamental studies on the biological effects of AEs. This work develops new biomedical cyclotron irradiation and radiochemical isolation methods to produce 165Er suitable for targeted radionuclide therapeutic studies and characterizes a new such agent targeting prostate-specific membrane antigen. Methods: Biomedical cyclotrons proton-irradiated spot-welded Ho(m) targets to produce 165Er, which was isolated via cation exchange chromatography (AG 50W-X8, 200–400 mesh, 20 mL) using alpha-hydroxyisobutyrate (70 mM, pH 4.7) followed by LN2 (20–50 µm, 1.3 mL) and bDGA (50–100 µm, 0.2 mL) extraction chromatography. The purified 165Er was radiolabeled with standard radiometal chelators and used to produce and characterize a new AE-emitting radiopharmaceutical, [165Er]PSMA-617. Results: Irradiation of 80–180 mg natHo targets with 40 µA of 11–12.5 MeV protons produced 165Er at 20–30 MBq·µA−1·h−1. The 4.9 ± 0.7 h radiochemical isolation yielded 165Er in 0.01 M HCl (400 µL) with decay-corrected (DC) yield of 64 ± 2% and a Ho/165Er separation factor of (2.8 ± 1.1) · 105. Radiolabeling experiments synthesized [165Er]PSMA-617 at DC molar activities of 37–130 GBq·µmol−1. Conclusions: A 2 h biomedical cyclotron irradiation and 5 h radiochemical separation produced GBq-scale 165Er suitable for producing radiopharmaceuticals at molar activities satisfactory for investigations of targeted radionuclide therapeutics. This will enable fundamental radiation biology experiments of pure AE-emitting therapeutic radiopharmaceuticals such as [165Er]PSMA-617, which will be used to understand the impact of AEs in PSMA-targeted radionuclide therapy of prostate cancer.
Molecular imaging is an important part of modern medicine which enables the non-invasive identification and characterization of diseases. With the advancement of radiochemistry and scanner technology, nuclear medicine is providing insight into efficient treatment options for individual patients. Titanium-45 (45Ti) is a lesser-explored radionuclide that is garnering increasing interest for the development of positron emission tomography (PET) radiopharmaceuticals. This review discusses aspects of this radionuclide including production, purification, radiochemistry development, and molecular imaging studies.
This project resulted in the synthesis and utility of new solid-phase extractants using diglycolamide (DGA) extractants grafted onto mesoporous silica. We used DGA ligands that offer multidentate coordination sites for lanthanides and offer pre-arranged binding sites that may facilitate radiolanthanide metal binding. Also, the Hunter graduate student worked at University of Utah on production and separation of 161 Tb with high specific activity resulting in a publication. These are reported in the full text upload.
“Isotope harvesting” is a technique that offers access to exotic radionuclides created as by-products during nuclear science research. Ongoing exploratory work at the National Superconducting Cyclotron Laboratory (NSCL) is directed towards the production and extraction of rare radionuclides from a flowing-water target and intends to pave the way for future harvesting efforts at the upcoming Facility for Rare Isotope Beams (FRIB). Here in this paper, we present the collection of 62 Zn from an aqueous matrix irradiated with a 150 MeV/nucleon 78Kr beam, while synergistically capturing other gaseous reaction products. In addition to the production rate for 62 Zn (9.08(30)E-5 62 Zn per incoming 78 Kr), the rates of formation for several other radionuclides were determined as well. The purification of 62 Zn from a large number of co-produced radionuclides was performed by anion exchange chromatography, allowing the isolation of 80.5(5.2) % of the generated 62 Zn. With the decay of 62 Zn the radioactive daughter 62 Cu is generated, and with the isolation of pure 62 Cu eluate, the principle of a medical radionuclide generator could be demonstrated. To illustrate the applicability of the obtained 62 Zn, the isolated product was used in free and DTPA-labelled form in a proof of principle plant uptake study with garden cress employing phosphor imaging for visualization.
The development of the Second Target Station (STS) target system at the Spallation Neutron Source (SNS) at Oak Ridge National Laboratory (ORNL) is well underway. The target system at STS consists of a rotating target disk that contains 21 segments of tungsten clad in tantalum clad in steel. A key aspect of the design of the target system is to account for the delayed heating and material damage caused by the delayed dose from decaying radionuclides. These radionuclides are a product of either spallation reactions or transmutation of the nuclei in the target system. These radionuclides build up in the target system components over the lifetime of the facility, and the radiation that is emitted can deposit energy in the components causing significant component heating and material damage. Monte Carlo N-Particle (MCNP) Version 6.2 transports the various particle species and calculates the spallation products and neutron fluxes throughout the target system. These spallation products and neutron fluxes along with the material definition of each component are relayed to the CINDER2008 transmutation code to calculate the radionuclide inventories and the corresponding decay gamma emission spectra. MCNP6.2 coupled with CINDER2008 is the computational method-of-choice for the analysis discussed in the following sections of this report. The analysis focuses on validating major assumptions in calculating the radionuclide inventory in the STS target system: all of the target segments are fresh, unirradiated material when the protons are incident on the segment, the average of the 21 segments of the target is sufficient to represent a single segment, and that averaging the proton pulse structure over time does not significantly affect the radionuclide inventory. The position-averaged, high-fidelity, and single-tally computational methods are used to validate the assumptions and provide a point of comparison to evaluate how the assumptions impact the radionuclide inventories. A more detailed explanation of the three computational methods is provided in Section 2. The position-averaged and single-tally methods are less computationally expensive when compared with the high-fidelity method where 54,000 individual calculations are needed to calculate 1 hr of STS operation. Section 3 details the comparison of the three methods to show that the assumptions made in the position-averaged method do not significantly impact the radionuclide inventory after 1 hr of operation. The discussions and results in this report are for 1 hr of operation. Due to the computational cost associated with calculating the transmutation and activation using the high-fidelity method, only 1 hr of operation has been calculated. The discrepancies observed after 1 hr of operation are not extrapolated out to longer operational times, and this report does not address how the discrepancies between the computational methods may manifest for longer operational periods.
This paper will document the air emissions source term for the TA-53 Building 0984 Isotope Production Facility (IPF). The facility has historically been managed as a minor source under Radionuclide NESHAP regulations, but began using a wider variety of irradiation targets and a higher beam current. To support expanded generation of actinium-225 for medical diagnostic and therapeutic uses, IPF is now irradiating thorium targets and the beam current is planned to increase from 250 microamperes up to 600 microamperes. These changes merit a review of the anticipated source term under the new proposed operational conditions. The end result of the review indicates the main stack (Exhaust Stack ES-1) needs to be managed as a major source with continuous emissions monitoring. No additional advance notifications or permitting is required as a result of these operational changes.
A pour stream (PS) sample taken during processing of Sludge Batch 8 (SB8, Macrobatch 10) was analyzed by Savannah River National Laboratory (SRNL) to verify compliance with the Defense Waste Processing Facility (DWPF) Glass Product Control Program. The following conclusions were drawn from these analyses; The sum of oxides for the SB8 PS glass (97.41 wt%) is within the Product Composition Control System (PCCS) limits (100 ± 5 wt%); The average Waste Dilution Factor (WDF) calculated from the Tank 40 dried sludge analysis results for SB8 is 1.9 ± 0.1. In general, the radionuclide concentration of the SB8 PS glass sample calculated using this WDF is in good agreement with the measured radionuclide content; The Product Consistency Test (PCT) results indicate that the SB8 PS glass meets the Waste Acceptance Product Specifications (WAPS) for durability, with normalized release rates for boron, lithium and sodium all at least two standard deviations lower than those of the Environmental Assessment (EA) glass; The average measured density of the SB8 PS glass is 2.72 g/cm 3 , consistent with previous PS samples; and, The average measured Fe 2+ /ΣFe ratio of the SB8 PS glass is 0.15, within the target range of 0.09- 0.33, and closely resembling the predicted REDOX value of 0.15.
Abstract The radioscandium isotopes, 43 Sc and 47 Sc, compose a promising elementally matched theranostic pair that can be used for the development of imaging and therapeutic radiopharmaceuticals with identical structures. This study aimed to investigate the production of high radionuclidic purity 43 Sc from enriched [ 46 Ti]TiO 2 targets and 47 Sc from enriched [ 50 Ti]TiO 2 targets and establish a target recycling technique. Enriched [ 46 Ti]TiO 2 targets were irradiated with 18 MeV protons, and enriched [ 50 Ti]TiO 2 targets were bombarded with 24 MeV protons. 43 Sc and 47 Sc were purified using ion chromatography attaining recovery yields of 91.7 ± 7.4% and 89.9 ± 3.9%, respectively. The average radionuclidic purity for 43 Sc was 98.8 ± 0.3% and for 47 Sc 91.5 ± 0.6%, while the average recovery of enriched titanium target material was 96 ± 4.0%. The highest apparent molar activity for [ 43 Sc]Sc-DOTA was 23.2 GBq/µmol and 3.39 GBq/µmol for [ 47 Sc]Sc-DOTA. This work demonstrates the feasibility of using enriched recycled [ 46 Ti]TiO 2 and [ 50 Ti]TiO 2 targets to produce high purity 43 Sc and 47 Sc as an elementally matched theranostic isotope pair.
Completion of the Proton Power Upgrade Project for the Spallation Neutron Source (SNS) accelerator at Oak Ridge National Laboratory opens an opportunity to utilize reserve beam power of more than 100 kW for applications beyond neutron production. One of these applications is the production of critical radionuclides. To demonstrate the feasibility of using the reserve beam power to produce radioisotope at SNS, a design concept of a small-scale experimental target station in the Linac Dump area has been developed. This experimental facility will provide isotope yield benchmarking data using protons in the GeV range. It will also enable additional research and development in isotope handling and radiochemical separation. The target station consists of a target module enclosed in a vessel and concrete shielding. Particle transport calculations and thermo-mechanical simulations are used to determine beam parameters, decay time, isotope yield, shielding dimensions, and target design parameters. Calculations verified that the irradiated capsule can be handled manually using hands-off tools and transported to a hot cell in a shielded container for post-irradiation characterizations.