Irradiated Thorium Research Articles

Automated preparation of Actinium-225 with accelerator-based production from irradiated natural thorium targets

Towards the goal of accelerator-based 225Ac production from proton and nitrogen irradiated natural thorium targets, this work hereby presents a self-designed preparation system that allows for continuous operation with improved multi-column ion-exchange method. The system with advantageous features validated the feasibility of the automatic separation and purification of actinium from hundreds of impurity nuclides. Through automated experiments with multiple targets, the recovery yield and the radionuclide purity of final 225Ac products was stabilized to be around 63 % and 92 % ∼ 96 %, respectively. The total content of stable impurities is measured to be less than 2.4 μg in the final actinium product. The radioactivity levels of 226Ac and 227Ac in the product are evaluated over time. Based on the results of evaluation, the optimal time to start separation is determined to be around the 10th day after the end of bombardment. Additionally, the 225Ac-PSMA-617 compound was synthesized with the 225Ac product that meets the requirement of nuclear medicine.

Radiation and Nuclear Physics Aspects of the Use of the Thorium Fuel Cycle in a Hybrid Fusion Facility

This paper presents the results of the experimental determination and computational simulation of the ambient dose equivalent rate for a metallic thorium cylindrical miniblock and the (n,2n), (n,f), and (n,γ) reaction rates in a thin 232Th metal foil irradiated with neutrons of the NG-24M generator spectrum. The ambient dose equivalent rate was determined by dosimeters-radiometers. The reaction rates were determined by the activation method using Ge spectrometers without destroying the irradiated samples. Computational simulations of ambient dose equivalent and reaction rates were performed, respectively, using the radiation transport codes PHITS, MCNP5, and KIR2, which use various nuclear data libraries: JEFF-3.2 and -3.3; JENDL4.0; ENDF/B-VII.0, -VII.1, and -VIII.0; ROSFOND; FENDL; and TENDL. The authors give an estimate of the 232U/233U relative accumulation upon natural thorium irradiation in a fusion facility blanket with defined neutron spectrum. The nonirradiated and irradiated thorium nuclide composition change simulation and visualization were performed using analytical solutions of an ordinary system of homogeneous linear differential equations describing nuclide transmutations.

A combined theoretical-experimental investigation of thermal transport in low-dose irradiated thorium dioxide

During reactor operation, nuclear fuels are subject to extreme temperature and irradiation conditions which can significantly degrade the fuel's thermal transport properties. The reduction in thermal conductivity of the fuel as a result of irradiation-induced lattice defects is arguably the most important fuel performance metric in regard to reactor efficiency and safety. Because thorium dioxide (ThO2) is suitable as a model system for more complex materials such as UO2 and its mixed oxides, we present a theoretical investigation of thermal conductivity reduction seen in defect-bearing thorium dioxide and compare directly to experimental measurements. Phonon-mediated thermal transport of the fuel is modeled by a solution to the Boltzmann transport equation (BTE) for phonons. A cluster dynamics (CD) model for lattice defect evolution during irradiation predicts defect densities which are used as input to the BTE for modeling phonon-defect scatterings. Phonon scatterings by lattice defects include those from point defects and vacancy clusters and interstitial clusters of various sizes. The CD model is benchmarked against structural defect characterization of irradiated thorium dioxide using electron microscopy. Thermal conductivity predicted by the BTE model is compared to measured values for irradiated thorium dioxide specimens below room temperature to isolate effects of phonon-defect scattering from intrinsic 3-phonon processes, which dominate at higher temperatures. The computed conductivity values are in partial agreement at temperatures close to room temperature while slight deviations are observed at the lowest measured temperatures, suggesting that implemented phonon-defect scattering cross-section expressions may not be adequate for low temperatures. The presented work provides a necessary investigation of the influence of irradiation induced defects on fuel performance and represents a first step toward a full characterization of phonon mediated thermal transport in irradiated materials with complex defect microstructure.

An alternative radiochemical separation strategy for isolation of Ac and Ra isotopes from high energy proton irradiated thorium targets for further application in Targeted Alpha Therapy (TAT)

Targeted Alpha Therapy (TAT) has shown very high potential for the treatment of cancers that were not responsive to other therapy options (e.g., β− therapy and chemotherapy). The main constraint to the widespread use of TAT in clinics is the limited availability of alpha-emitting radionuclides. One of the most promising candidates for TAT is 225Ac (t1/2 = 9.92 days), which can be used directly in combination with selective biomolecules (e.g., antibodies, peptides, etc.) or be a generator source of 213Bi (t1/2 = 45.6 min), another shorter-lived TAT radionuclide. Several strategies are currently under investigation to increase the supply of 225Ac. One of the most attractive options is the irradiation of natural thorium-232 targets with high-energy protons (≥100 MeV). However, there are several challenges associated with this production method including the development of an efficient radiochemical purification method.During irradiation of natural thorium with proton energy above 100 MeV, several Ra isotopes (223,224,225Ra) are produced. 223Ra (t1/2 = 11.43 days) is used for the treatment of bone metastases and can also be used as a generator source for 211Pb. Additionally, 225Ra (t1/2 = 14.9 days) can be a valuable source of isotopically pure 225Ac.In the present work, we address the radiochemical separation aspects of isolating Ac and Ra isotopes from irradiated thorium targets.

A Combined Theoretical-Experimental Investigation of Thermal Transport in Low-Dose Irradiated Thorium Dioxide

A Combined Theoretical-Experimental Investigation of Thermal Transport in Low-Dose Irradiated Thorium Dioxide

An integrated experimental and computational investigation of defect and microstructural effects on thermal transport in thorium dioxide

Advanced nuclear reactor concepts aim to use fuels that must withstand unprecedented temperature and radiation extremes. In these fuels, thermal energy transport under irradiation is directly related to fuel longevity, reactor safety, and is arguably one of the most important performance metrics. Here we provide a comprehensive, first-principles-informed treatment of phonon mediated thermal transport in a defect-bearing actinide oxide with direct comparison to experimental measurements. Pristine and proton irradiated thorium dioxide was chosen as a model system to treat the complexity of thermal transport in the presence of lattice defects. A thermal transport model is implemented using the linearized Boltzmann transport equation (LBTE) with input from first principles calculations and defect evolution models. Density functional theory is used to calculate phonon dispersion in thorium dioxide and used as an input to calculate both intrinsic and extrinsic, defect-induced relaxation times. In addition, a defect evolution model is benchmarked using microstructure characterization of as-irradiated thorium dioxide using a combination of electron microscopy and optical spectroscopy. The output of the LBTE is compared directly to mesoscopic measurements of thermal conductivity on length scales commensurate with defect accumulation. Parametric measurements of conductivity with irradiation dose and temperature suggest a saturation in the reduction of thermal conductivity with increasing defect generation, which is partially captured in our defect evolution model and LBTE framework. This comprehensive, atomistic- to meso-scale treatment provides the necessary basis to investigate thermal transport under irradiation in more complex systems that exhibit strong electron correlation.

232Th-Spallation-Produced 225Ac with Reduced 227Ac Content.

Recent clinical results have demonstrated remarkable treatment responses of late-stage cancer patients when treated with alpha-emitting radionuclides such as actinium-225 (225Ac). The resulting intense global effort to produce greater quantities of 225Ac has triggered a number of emerging technologies to produce this rare, yet important, radionuclide. Accelerator-based methods for increasing global 225Ac production capacity have focused on the high energy (>100 MeV) proton irradiation of thorium, despite the coproduction of the undesirable 227Ac byproduct at 0.1-0.3% of the 225Ac activity. We at TRIUMF have developed a process for the production of a 225Ra/225Ac generator from irradiated thorium that results in an 225Ac product with reduced 227Ac content. 225Ac was separated from irradiated thorium and coproduced radioactive spallation and fission products using a thorium peroxide precipitation method followed by cation exchange and extraction chromatography. Stable and radioactive tracer studies demonstrated the ability of this method to separate Ac from most other elements, providing a directly produced Ac product with measured 227Ac content of (0.15 ± 0.04)%. A second, indirectly produced Ac product with 227Ac content of <7.5 × 10-5% is obtained by repeating the final extraction chromatography step with the 225Ra-containing fraction. The 225Ra-derived 225Ac showed similar or improved quality compared to the initial, directly produced 225Ac product in terms of chemical purity and radiolabeling capability, the latter of which was comparable with other 225Ac sources reported in the literature.

Chemical Purification of Actinium-225 from Proton-Irradiated Thorium Targets

The chemical separation of actinium from proton-irradiated 232Th targets is being performed as part of a tri-laboratory (Oak Ridge National Laboratory, Brookhaven National Laboratory, and Los Alamos National Laboratory) effort focused on the accelerator production of 225Ac. Actinium-225 (t1/2 = 9.92 days) is produced by proton irradiation of a thorium metal target via the 232Th[p,x]225Ac reaction and can be used in targeted alpha therapy applications. The irradiation produces a suite of radioisotopes, including fission products, that must be chemically separated from actinium, as well as a vast excess of thorium. This separation is achieved by chemically processing irradiated thorium targets through a series of ion separation and extraction chromatography columns. This effort has produced purified 225Ac over 15 campaigns, providing material for use in targeted alpha therapy research and development.

Investigation of a tellurium-packed column for isolation of astatine-211 from irradiated bismuth targets and demonstration of a semi-automated system

Astatine-211 is an attractive radionuclide for use in targeted alpha therapy of blood-borne diseases and micrometastatic diseases. Efficient isolation methods that can be adapted to robust automated 211At isolation systems are of high interest for improving the availability of 211At. Based on the early studies of Bochvarova and co-workers involving isolation of 211At from irradiated thorium targets, we developed a method for 211At isolation from bismuth targets using tellurium-packed columns. Dissolution of irradiated bismuth targets is accomplished using HNO3; however, 211At is not captured on the Te column material in this matrix. Our method involves slow addition of aqueous NH2OH·HCl to the Bi target dissolved in HNO3 to convert to a HCl matrix. The amount of NH2OH·HCl was optimized because (1) the quantity of NH2OH·HCl used appears to affect the radiolabeling yield of phenethyl-closo-decaborate(2-) (B10)-conjugated antibodies and (2) reducing the volume of NH2OH·HCl solution can effectively shorten the overall isolation time. A proof-of-concept semi-automated process has been demonstrated using targets containing ~0.96 GBq (~26 mCi) of 211At. High isolation yields (88–95%) were obtained. Radiochemical purity of the isolated 211At was assessed by radio-HPLC. Concentrations of Bi and Te contaminants in the 211At and the astatinated antibodies were evaluated using ICP-MS.

ICP-AES Characterization of PHWR Irradiated Thoria Bundles for Fission Products

ICP-AES Characterization of PHWR Irradiated Thoria Bundles for Fission Products

The application of poorly crystalline silicotitanate in production of 225Ac

Actinium-225 (225Ac) can be produced from a Thorium-229/Radium-225 (229Th/225Ra) generator, from high/low energy proton irradiated natural Thorium or Radium-226 target. Titanium based ion exchanger were evaluated for purification of 225Ac. Poorly crystalline silicotitanate (PCST) ion exchanger had high selectivity for Ba, Ag and Th. 225Ac was received with trace amounts of 227Ac, 227Th and 223Ra, and the solution was used to evaluate the retention of the isotopes on PCST ion exchanger. Over 90% of the 225Ac was recovered from PCST, and the radiopurity was >99% (calculated based on 225Ac, 227Th, and 223Ra). The capacity of the PCST inorganic ion exchange for Barium and 232Th was determined to be 24.19 mg/mL for Barium and 5.05 mg/mL for Thorium. PCST ion exchanger could separate 225Ac from isotopes of Ra and Th, and the process represents an interesting one step separation that could be used in an 225Ac generator from 225Ra and/or 229Th. Capacity studies indicated PCST could be used to separate 225Ac produced on small 226Ra targets (0.3–1 g), but PCST did not have a high enough capacity for production scale Th targets (50–100 g).

Measurement and analysis of 233U from local thorium by using gamma spectrometry and DNCS methods

The measurement and analysis of 233U from irradiated local thorium sample by gamma spectrometry and fission-induced delayed neutron counting system (DNCS) methods have been done. The sample was a small high purity local ThO2 produced from the extraction of monazite sand at our research center. The main goal of this experiment was to explore possibilities of 233U determination in the irradiated thorium for its future use in SAMOP reactor. Some ThO2 samples of 0.1 gram were irradiated at the Kartini reactor at average neutron flux of 1012 n cm−2s−1 and cooled down for 55 days before counted by gamma spectrometry where 233U determined indirectly by gamma peak counting of 233Pa. The samples then re-irradiated to determine 233U by using DNCS method. The analysis result both by using gamma spectrometry and DNC methods were 10.10 ± 0.96 μg and 17.22 ± 1.85 μg of 233U in average. Of the two gamma spectrometry and DNCS methods both provided results in somewhat relatively good agreement with the calculated amounts of 233U using ORIGEN2 computer code i.e. 14.3 μg. The detection limit of DNCS and the efficiency of detection system measured are 0.006 μg and 6.39 ± 0.07% respectively, and the efficiency of gamma spectrometry system is 17.8%.

Conceptual potential of a pyroelectrochemical technology for the thorium engagement in the fast neutron fuel cycle

The use of thorium in combination with plutonium in nuclear power generation offers a solution to the problem of reducing the accumulation of long-lived transplutonium nuclides. Along with this, the existing uranium fuel cycle (UFC) has such disadvantage as the vulnerability to unauthorized use of nuclear materials. The thorium fuel cycle (TFC) is devoid of these drawbacks. The engagement of thorium in nuclear power is possible provided the availability of an appropriate technology for reprocessing irradiated thorium. A fuel cycle based on thorium oxide may not differ in principle from the already developed pyrochemical fuel cycle involving uranium and plutonium oxides. Thorium oxide is most commonly obtained in compact state by electrolysis of molten salts from thorium-containing electrolytes. The most thorough studies of physical and chemical and electrochemical behavior of thorium in molten haloids of alkali and alkaline-earth metals were conducted in the 1960ies and the 1970ies. Since extensive experimental material has been accumulated by now for justification of the use of pyroelectrochemical and chemical processes for regeneration of fuel in molten salts, then it has also been proposed that technologies for fuel reprocessing in molten chlorides of alkali metals should be applied resulting in a crystalline product that can be used for the fuel element fabrication. Unlike uranium and plutonium, thorium behavior in molten salt environments is less complex. In molten salts, thorium exists predominantly in the form of Th4+, and the mixture of uranium and thorium dioxides with ThO2 content reaching up to 50 % can be obtained by electrolysis of molten salts. Therefore, the existing amount of knowledge about the chemistry of thorium allows regarding the use of pyrochemical processes in production of thorium oxide as highly promising, and the available data on the physical and chemical properties of thorium and its compounds in high-temperature molten salts makes it possible to state that the pyroelectrochemical technology can be potentially used in production and reprocessing of thorium fuel.

Defining Processing Times for Accelerator Produced 225Ac and Other Isotopes from Proton Irradiated Thorium.

During the purification of radioisotopes, decay periods or time dependent purification steps may be required to achieve a certain level of radiopurity in the final product. Actinum-225 (Ac-225), Silver-111 (Ag-111), Astatine-211 (At-211), Ruthenium-105 (Ru-105), and Rhodium-105 (Rh-105) are produced in a high energy proton irradiated thorium target. Experimentally measured cross sections, along with MCNP6-generated cross sections, were used to determine the quantities of Ac-225, Ag-111, At-211, Ru-105, Rh-105, and other co-produced radioactive impurities produced in a proton irradiated thorium target at Brookhaven Linac Isotope Producer (BLIP). Ac-225 and Ag-111 can be produced with high radiopurity by the proton irradiation of a thorium target at BLIP.

Fluoride volatility experiments on irradiated thoria fuel at Canadian Nuclear Laboratories

Fluoride volatility experiments were completed on irradiated thoria fuel in a hot cell facility at Canadian Nuclear Laboratories (CNL). Uranium was removed from irradiated thoria fuel in significant quantities, with longer fluorination times resulting in greater uranium removal for tests up to 24 h. Some fission products were also removed via fluorination; fission products with fluorides having low boiling points (Mo and Ru) were removed in greater quantities than fission products with fluorides having higher boiling points (Zr and Rh).

Evaluation of SynPhase Lanterns for capturing Ac-225 from bulk thorium

Proton irradiation of a thorium (Th) target can produce Actinium-225 (Ac-225), but the irradiated thorium fissions results in the production of 100’s of other isotopes. SynPhase Lanterns containing sulfonic acid groups were evaluated for the purification of Ac-225 from Th and other fission metals. The SynPhase Lanterns were able to quantitatively and selectively capture Ac-225 and lanthanum (La) over Th. A second purification step would be needed to purify Ac-225 from trace amounts of Th and other fission products. The results from these studies were compared to similar studies performed with cation exchange resins, and the resins were superior at removing impurities.

Separation of Protactinium Employing Sulfur-Based Extraction Chromatographic Resins.

Protactinium-230 ( t1/2 = 17.4 d) is the parent isotope of 230U ( t1/2 = 20.8 d), a radionuclide of interest for targeted alpha therapy (TAT). Column chromatographic methods have been developed to separate no-carrier-added 230Pa from proton irradiated thorium targets and accompanying fission products. Results reported within demonstrate the use of novel sulfur bearing chromatographic extraction resins for the selective separation of protactinium. The recovery yield of 230Pa was 93 ± 4% employing a R3P═S type commercially available resin and 88 ± 4% employing a DGTA (diglycothioamide) containing custom synthesized extraction chromatographic resin. The radiochemical purity of the recovered 230Pa was measured via high purity germanium γ-ray spectroscopy to be >99.5% with the remaining radioactive contaminant being 95Nb due to its similar chemistry to protactinium. Measured equilibrium distribution coefficients for protactinium, thorium, uranium, niobium, radium, and actinium on both the R3P═S type and the DGTA resin in hydrochloric acid media are reported, to the best of our knowledge, for the first time.

Separation of 103Ru from a proton irradiated thorium matrix: A potential source of Auger therapy radionuclide 103mRh.

Ruthenium-103 is the parent isotope of 103mRh (t1/2 56.1 min), an isotope of interest for Auger electron therapy. During the proton irradiation of thorium targets, large amounts of 103Ru are generated through proton induced fission. The development of a two part chemical separation process to isolate 103Ru in high yield and purity from a proton irradiated thorium matrix on an analytical scale is described herein. The first part employed an anion exchange column to remove cationic actinide/lanthanide impurities along with the majority of the transition metal fission products. Secondly, an extraction chromatographic column utilizing diglycolamide functional groups was used to decontaminate 103Ru from the remaining impurities. This method resulted in a final radiochemical yield of 83 ± 5% of 103Ru with a purity of 99.9%. Additionally, measured nuclear reaction cross sections for the formation of 103Ru and 106Ru via the 232Th(p,f)103,106Ru reactions are reported within.

Chromatographic separation of the theranostic radionuclide 111Ag from a proton irradiated thorium matrix

Column chromatographic methods have been developed to separate no-carrier-added 111Ag from proton irradiated thorium targets and associated fission products as an ancillary process to an existing 225Ac separation design. Herein we report the separation of 111Ag both prior and subsequent to 225Ac recovery using CL resin, a solvent impregnated resin (SIR) that carries an organic solution of alkyl phosphine sulfides (R3P = S) and alkyl phosphine oxides (R3P = O). The recovery yield of 111Ag was 93 ± 9% with a radiochemical purity of 99.9% (prior) and 87 ± 9% with a radiochemical purity of 99.9% (subsequent to) 225Ac recovery. Both processes were successfully performed with insignificant impacts on 225Ac yields or quality. Measured equilibrium distribution coefficients for silver and ruthenium (a residual contaminant) on CL resin in hydrochloric and nitric acid media are reported, to the best of our knowledge, for the first time. Additionally, measured cross sections for the production of 111Ag and 110mAg for the 232Th(p,f)110m,111Ag reactions are reported within.

Simultaneous Separation of Actinium and Radium Isotopes from a Proton Irradiated Thorium Matrix

A new method has been developed for the isolation of 223,224,225Ra, in high yield and purity, from a proton irradiated 232Th matrix. Herein we report an all-aqueous process using multiple solid-supported adsorption steps including a citrate chelation method developed to remove >99.9% of the barium contaminants by activity from the final radium product. A procedure involving the use of three columns in succession was developed, and the separation of 223,224,225Ra from the thorium matrix was obtained with an overall recovery yield of 91 ± 3%, average radiochemical purity of 99.9%, and production yields that correspond to physical yields based on previously measured excitation functions.