Thermal Neutron Flux Research Articles

Production and radiochemical separation of Tb-161 as a feasible beta therapeutic radioisotope from reactor irradiated Gd target

Terbium-161 (161Tb) is a promising therapeutic radionuclide that has gained significant attention in the field of nuclear medicine in recent years. 161Tb has several favorable characteristics that make it a valuable candidate for targeted radionuclide therapy. The production of No-carrier-added 161Tb was carried out by the neutron activation of natural gadolinium target in the Egyptian Second Research Reactor (ETRR-2) at a thermal neutron flux position of 1.8 × 1014 n cm-2 s-1. The radioactivities of 161Tb as well as coproduced Gd radioimpurities were computed theoretically by the MCNPX2.7.0 code and verified by actual measurements, where accepted discrepancies were obtained. The purification of 161Tb from irradiated Gd target was developed by Chelex-100 resin. The elution performance was studied using different eluents, and 0.1 M HNO3 was found to be the best medium to obtain a high separation efficiency of more than 92% in a short time. The eluted 161Tb was of high chemical, radiochemical, and radionuclidic purities, indicating its potential for effective application in radiopharmaceutical preparation.

Sensitivity improvement of a deuterium-deuterium neutron generator based in vivo neutron activation analysis (IVNAA) system.

Our lab has been developing a deuterium-deuterium (DD) neutron generator-based neutron activation analysis (NAA) system to quantify metals and elements in the human body in vivo. The system has been used to quantify metals such as manganese, aluminum, sodium in bones of a living human. The technology provides a useful way to assess metal exposure and to estimate elemental deposition, storage and biokinetics. It has great potential to be applied in the occupational and environmental health fields to study the association of metal exposure and various health outcomes, as well as in the nutrition field to study the intake of essential elements and human health. However, the relatively low sensitivity of the system has greatly limited its applications. Neutron moderation plays an important role in designing an IVNAA facility, as it affects thermal neutron flux in irradiation cave and radiation exposure to the human subject. This study aims to develop a novel thermal neutron enhancement method to improve the sensitivity of the in vivo neutron activation analysis (IVNAA) system for elemental measurement but still maintain radiation dose. Utilizing a compact DD neutron source, we propose a new and practical moderator design that combines high density polyethylene with heavy water to enhance thermal neutrons by reducing thermal neutron absorption. All material dimensions are calculated by PHITS, a general-purpose Monte Carlo simulation program. The improvement of the new design predicted by the Monte Carlo simulation for the quantification of one of the elements, manganese was verified by experimental irradiation of manganese-doped bone equivalent phantoms. For the same radiation dose, a 67.9% thermal neutron flux enhancement is reached. With only 4.2% increase of radiation dose, the simulated thermal neutron flux and activation can be further increased by 84.2%. A 100% thermal neutron enhancement ratio is also achievable with a 20% dose increase. The experimental results clearly show higher manganese activation gamma ray counts for each specific phantom, with a significantly reduced minimum detection limit. Additionally, the photon dose was suppressed. The thermal neutron enhancement method can increase the number of useful neutrons significantly but maintain the radiation dose. This greatly decreased the detection limit of the system for elemental quantification at an acceptable dose, which will broadly expand the application of the technology in research and clinical use. The method can also be applied to other neutron medical applications, including neutron imaging and radiotherapy.

Effect of neutron beam properties on dose distributions in a water phantom for boron neutron capture therapy.

From the viewpoints of the advantage depths (ADs), peak tumor dose and skin dose, we evaluated the effect on the dose distribution of neutron beam properties, namely the ratio between thermal and epithermal neutron fluxes (thermal/epithermal ratio), fast neutron component and γ-ray component. Several parameter surveys were conducted with respect to the beam properties of neutron sources for boron neutron capture therapy assuming boronophenylalanine as the boron agent using our dose calculation tool, called SiDE. The ADs decreased by 3% at a thermal/epithermal ratio of 20-30% compared with the current recommendation of 5%. The skin dose increased with the increasing thermal/epithermal ratio, reaching a restricted value of 14 Gyeq at a thermal/epithermal ratio of 48%. The fast neutron component was modified using two different models, namely the 'linear model', in which the fast neutron intensity decreases log-linearly with the increasing neutron energy, and the 'moderator thickness (MT) model', in which the fast neutron component is varied by adjusting the MT in a virtual beam shaping assembly. Although a higher fast neutron component indicated a higher skin dose, the increment was <10% at a fast neutron component of <1 × 10-12Gycm2 for both models. Furthermore, in the MT model, the epithermal neutron intensity at a fast neutron component of 6.8 × 10-13Gycm2 was 41% higher compared with that of 2 × 10-13Gycm2. The γ-ray component also caused no significant disadvantages up to several times larger compared with the current recommendation.

Optimization study on the compact thermal neutron imaging system based on portable D-T neutron generator

The distribution of lithium in lithium-ion batteries is critical to their performance and lifespan. Studies have shown that thermal neutron imaging can effectively perform non-destructive testing of this distribution. However, most current thermal neutron imaging systems rely on expensive, bulky, and immobile nuclear reactors or spallation neutron sources, significantly limiting their application scenarios. Therefore, this study adopts a portable D-T neutron generator as the neutron source and uses Monte Carlo simulation methods to optimize the design of key components in the thermal neutron imaging system. Simulation results indicate that the optimized thermal neutron imaging system achieves a thermal neutron flux of 8.36 × 104 n·cm−2·s−1 and a thermal neutron content (TNC) of 14.4% at a collimation ratio of 50. Within an imaging range of Φ100 mm, the non-uniformity of the thermal neutron fluence is approximately 2.07%. Simulation imaging results of lithium and iron step wedges demonstrate that the compact thermal neutron imaging system designed in this study can effectively distinguish between iron and lithium of the same thickness. This finding provides an important reference for subsequent detection of lithium distribution within lithium-ion batteries.

The production and separation of 161Tb with high specific activity at the University of Utah

Targeted radiotherapy (TRT) is an increasingly prominent area of research in nuclear medicine, particularly in the context of treating cancerous tumors. One radionuclide of considerable interest for TRT is terbium-161 (t1/2 = 6.95 days), which undergoes beta emission and shares similar decay properties as 177Lu (FDA-approved as LUTATHERA® and PLUVICTO®). Besides beta emission, 161Tb also emits a significant number of conversion and Auger electrons further enhancing its therapeutic potential. Terbium-161 can be produced using nuclear reactors through an indirect neutron capture reaction, G64160dn,γG64161d→3.66min,β−T65161b, from 160Gd targets. However, a key challenge in utilizing 161Tb for TRT lies in effectively separating target and product materials to attain high specific activity for radiolabeling. Here, we detail the production of no-carrier added 161Tb using low flux research reactors (mean thermal (<0.625 eV) neutron flux: 1.356×1012n∙cm−2∙s−1) like the University of Utah TRIGA Reactor, using enriched 160Gd2O3 targets (1.5 ± 0.3 μCi of 161Tb per mg of 160Gd target per hour of irradiation). We also developed a separation technique based on cation exchange and extraction chromatography, suitable for mCi level irradiations with targets exceeding 200 mg. In a simulated full-scale irradiation, 161Tb was successfully isolated from large mass targets using cation exchange (AG 50W-X8, with 2-hydroxyisobutyric acid at 70 mM, pH 4.75) and extraction chromatography (LN Resin, 0.5–0.75 M HNO3) methods. This resulted in high apparent molar activities of [161Tb]Tb-DOTA (113 ± 3 MBq/nmol), demonstrating high purity 161Tb relevant for potential future preclinical applications.

Polymer Boron-Containing Composite for Protecting Astronauts of Manned Orbital Stations from Secondary Neutron Radiation

This article considers the prospects of using heat-resistant polyimide boron-containing composites to protect astronauts of manned orbital stations from secondary neutron radiation. Variant calculations are performed regarding neutron and gamma-quanta flux distributions in a polyimide composite material with different boron content used to reduce capture radiation. The dependences of spatial distributions of thermal neutron flux density and the gamma-quanta dose rate in a polyimide composite layer with a boron content of 0 to 5% are obtained. An experimental assessment of the energy distribution of neutron and gamma radiation behind the protective polyimide composite is carried out. The introduction of boron atoms in an amount of 3.0 wt.% shows the absence of bursts of secondary gamma radiation energy in the composite, which is due to the high cross-section of thermal neutron absorption by boron atoms. As a result, with a material layer thickness of 3–10 cm, the gamma-quanta dose rate decreases by 2–3 times. The differential thermal analysis method showed that the upper limit of the working temperature of the polyimide composite is 500 °C. The polyimide matrix filled with boron atoms can find effective application in the development of new radiation-protective polymer materials used in manned orbital stations.

Nuclear-excited source of coherent and incoherent radiation with direct nuclear pumping

Uranium fission fragments, as well as the products of 3He(n,p)3H and 10B(n,α)7Li nuclear reactions were utilized in the nuclear reactor for gas ionization and excitation. However, the 6Li(n,α)3H nuclear reaction was less examined. The use of lithium-6 as a surface source of excitation of the gas medium, due to the long path length of tritium nuclei in the gas, allows to excite large volumes of gas as opposed to using 235U or 10B.While investigating the luminescence of noble gases in the core of the IVG.1M research reactor, we noted an appearance of alkali metal lines and a sharp increase in the intensity of these lines at temperatures above 570 K. It was determined that the population of levels of lithium atoms has practically no effect on the population of the 2p-levels of atoms of noble gases. The selectivity of p- and s-levels deactivation by lithium atoms implies the possibility of creating inversion of population at 2p-1s transitions of noble gas atoms. Successful experiments to study the luminescence of gases upon excitation by 6Li(n,α)3H nuclear reaction products allow us to proceed to experiments to achieve the laser action threshold and study the lasing characteristics of gas mixtures at the IGR pulsed nuclear reactor with thermal neutron flux density up to 7∙1016 n/cm2s. For this purpose, an experimental device designs were proposed to perform experiments on the IGR reactor. A step-by-step procedure of fabrication of a nuclear-excited source for excitation of gas mixtures is provided. The results of reactor experiments aimed at determining the spectral and temporal characteristics of optical radiation during excitation of gas mixtures by 6Li(n,α)3H nuclear reaction products are presented.

Optimization and shielding design of a thermal neutron device based on D-D neutron generator with Geant4 toolkit

Neutron activation analysis is a highly sensitive non-destructive testing technique with important applications in industry, geoscience, medical therapy, etc. This work designed and optimized a thermal neutron device that utilized a portable D-D neutron generator, and the Monte Carlo method with the Geant4 toolkit was applied to simulation. The objective of the optimized design is to maximize the thermal neutron flux at the output surface and increase the utilization efficiency of the neutron generator. A parameter K was defined as a measure of the device's slowing capacity for neutrons and was used to determine the optimized device geometry. The simulation considered the contribution of different types and sizes of moderators and reflectors to the thermal neutron intensity to obtain the optimal size. The shielding protection of the device was then designed. The effectiveness of shielding with different thicknesses was evaluated using three dose reference points. The results indicated that the optimized device can achieve a maximum thermal neutron flux of 1.97 × 105 n∙cm−2∙s−1 at the output surface by using high-density polyethylene (HDPE) as the moderator and nickel as the reflector. It was determined that using 45 cm of HDPE and 9 cm of lead protection in sequence along the neutron head axis would reduce the dose rate at the reference point, located 5 cm from the surface of the device, below the safety limit of 2.5 μSv/h.

Introduction of the prompt γ-ray neutron activation analysis system at CARR and the first pilot experiment on boron-containing high-temperature alloys

A prompt γ-ray neutron activation analysis system has recently been developed at China advanced research reactor (CARR), the 60 MW research reactor in China Institute of Atomic Energy (CIAE). The system is set at the cold neutron beam guide with a thermal equivalent neutron flux at the sample position of 1.0 × 109 n·cm−2·s−1 with the power of 30 MW, and it is mainly composed of a neutron beam collimator, a sample chamber, a beam stopper, neutron and γ-ray shieldings and a detection system. The detection system can realize three modes of measurement: single, Compton suppression, and pair modes. The detection efficiency was calibrated up to 11 MeV using a set of radionuclides and the (n, γ) reactions of N and Cl. Boron, one of the most important elements in high-temperature alloy material studies, was analyzed in this work, as the first pilot experiment of the CARR-PGNAA system. The analytical sensitivity of 2000 cps/mg-B was obtained. The results verified the feasibility of the CARR-PGNAA system to measure boron in high-temperature alloys, and laid a foundation for the accurate quantification of boron in the next step.

Bricks non-destructive simulation testing method utilizing neutron radiography facility based on a 7Li(p,n)7Be reaction

The evaluation of non-destructive neutron radiography (NR) for examining the internal composition of various structural materials, has been the focus of extensive research. This manuscript uses non-destructive testing to generate three-dimensional radiographs of three different brick structural materials: glass block, magnesia-chrome, and lead to evaluate their capability to withstand fast neutrons and gammas emitted from a source. When neutrons with thermal or epithermal spectrum are required, the optimum combination for an accelerator was simulated using a 2.8 MeV proton beam on a lithium target. The presented facility tested both thermal and fast neutron radiography. This study examined various aperture diameters and collimator lengths. It found that implementing a special fast neutron filter significantly increased the thermal neutron content (TNC) with minimal impact on the thermal neutron flux. For fast neutron radiography, the study evaluated parameters such as geometric unsharpness, fast neutron flux, and the percentage of the uncollided fast neutron reaching the object. Both neutrons and photons from the source were used to inspect faults in a glass brick.

Measurement of potassium in rats using an In vivo neutron activation analysis system

Abnormal levels of potassium are linked to several health conditions, including high blood pressure, cardiac dysfunction, kidney damage, and osteoporosis. Given the limited availability of in vivo measurement techniques, there is a need for novel methods to measure potassium to enhance the diagnosis and management of potassium metabolism related diseases. This study aimed to evaluate the feasibility of compact neutron generator based in vivo measurement system for quantification of potassium using rat carcasses. A cohort of thirty-nine rats (n = 20 males and 19 females, average weight 255 ± 15 and 163 ± 7 g) were sacrificed, and their carcasses were placed in polyethylene bottles. The rats were then positioned and irradiated in a carefully designed irradiation cave built alongside the neutron generator with an optimized thermal neutron flux and radiation dose ratio. The irradiation time was 10 min, followed by a 5-min decay and 2-h measurement using a high efficiency high purity germanium detector(HPGe). ResultsThe average potassium concentration in male and female rats was found to be comparable (male 2874 ± 161 and female 2866 ± 144 μg/g). A marginally positive correlation between potassium concentration and weight was found in female rats only (male(20) = 0.07, P = 0.76 and female r(19) = 0.34, P = 0.15). We assessed the influence of manganese toxicity on potassium levels and observed no significant impact. These results were consistent with our previous study in mice. ConclusionThis study suggests that in vivo neutron activation analysis could serve as a promising method to quantify potassium and to investigate the storage and metabolism of potassium in human and in animals.

Combined X-ray Diffraction Analysis and Quantum Chemical Interpretation of the Effect of Thermal Neutrons on the Geometry and Electronic Properties of CdTe

Objective: Substantiation of the result of the interaction of thermal neutrons with CdTe crystals by quantum-chemical methods and comparison of the experimental results with the results of quantum-chemical calculations. Methods: Single crystal samples of cadmium telluride (CdTe) were obtained by a modified Bridgman method. The plates cut from the single crystal were subjected to mechanical grinding with abrasive paper of the type P1,200-P4,000, mechanical polishing with silver paste. To clean the surface of the plate from contamination, it was washed in ethyl alcohol. In the experimental part of the work, the influence of low thermal neutron fluxes on the structural parameters of CdTe was studied. The samples were irradiated with a thermal neutron flux in the range from 2.64×107 neutron/cm2 to 1.85×109 neutron/cm2. To study the structure of CdTe crystals, the method of X-ray structural analysis was used, using a DRON-3 X-ray diffractometer. In the second part of the work, computer modeling of the process of interaction of thermal neutrons on CdTe crystals was carried out. Results: X-ray diffraction analysis of the crystals showed that as a result of irradiation with low thermal neutron fluxes, the structural parameters of CdTe improve, as evidenced by the ordering of the crystallites. In addition, in this region of the thermal neutron flux the resistance of the crystals decreases. Conclusion: Calculations of the crystal lattice parameters of CdTe are in good agreement with experimental data. The electronic and optical properties of CdTe have been studied. A decrease in the band gap after irradiation with thermal neutrons has been shown.

Modelling Gd-diamond and Gd-SiC neutron detectors

A custom Monte Carlo (MC) computer model was developed to simulate thermal neutron absorption in, and subsequent photon and electron emission from, natural Gd with a view to using the material as a neutron conversion layer for neutron detectors. The MC code also modelled photon and electron detection with two dissimilar detectors: a thick (500 μm) single crystal diamond detector; and a thin (5.15 μm) commercial off the shelf (COTS) 4H–SiC photodiode detector. The detectors’ quantum detection efficiencies (QE) for hard X-rays and γ-rays were relatively low in comparison to their QE for electrons, thus making it possible to collect electron spectra from the Gd layer neutron conversion products which were not overwhelmed by photon emissions from the Gd. The MC code was utilised to determine the optimal thickness of Gd for the efficient detection of a thermal neutron flux. These radiation hard and spectroscopic detectors paired with natural Gd could find utility as robust and compact thermal neutron detectors for nuclear science and engineering, space science, and other applications.

Actinium-225 photonuclear production in nuclear reactors using a mixed radium-226 and gadolinium-157 target

BackgroundActinium-225 is one of the most promising radionuclides for targeted alpha therapy. Its limited availability significantly restricts clinical trials and potential applications of 225Ac-based radiopharmaceuticals. MethodsIn this work, we examine the possibility of 225Ac production from the thermal neutron flux of a nuclear reactor. For this purpose, a target consisting of 1.4 mg of 226Ra(NO3)2 (T1/2 = 1600 years) and 115.5 mg of 90 % enriched, stable 157Gd2O3 was irradiated for 48 h in the Breazeale Nuclear Reactor with an average neutron flux of 1.7·1013 cm−2·s−1. Gadolinium-157 has one of the highest thermal neutron capture cross sections of 0.25 Mb, and its neutron capture results in emission of high-energy, prompt γ-photons. Emitted γ-photons interact with 226Ra to produce 225Ra according to the 226Ra(γ, n)225Ra reaction. Gadolinium debulking and separation of undesirable, co-produced 227Ac from 225Ra was achieved in one step by using 60 g of branched DGA resin. After 225Ac ingrowth from 225Ra (T1/2 = 14.8 d), 225Ac was extracted from the 226Ra and 225Ra fraction using 5 g of bDGA resin and then eluted using 5 mM HNO3. ResultsMeasured activity of 225Ac showed that 6(1) kBq or 0.16(3) μCi (1σ) of 225Ra was produced at the end of bombardment from 0.9 mg of 226Ra. ConclusionThe developed 225Ac separation is a waste-free process which can be used to obtain pure 225Ac in a nuclear reactor.

Refueling Considerations for 99Mo Production in a CANDU Reactor

Operating CANDU reactors have the potential to produce significant quantities of molybdenum-99 (99Mo) because of their ability to be refueled online, high thermal neutron flux, and fuel design flexibility. A new molybdenum-producing fuel bundle (MPB), previously designed for CANDU reactors, has as its principal attribute that it is neutronically and thermal hydraulically equivalent to the standard 37-element fuel bundle typically used in CANDU reactors. Given that the typical irradiation time for MPBs is 20 days while the typical refueling period for a channel is on average 6 months, the refueling strategy needs to be adjusted to accommodate the shorter irradiation time of MPBs. This study evaluates a new refueling strategy suitable for employing the new MPBs in the core. A full-core, three-dimensional model is constructed in the diffusion code DONJON, and a fueling strategy for achieving the desired weekly yield of 99Mo is developed. The adequacy of the proposed refueling scheme is evaluated using a series of time-average calculations, which show that a small increase in the core reactivity (<0.4 mk) can be expected when irradiating a set of four MPBs in three different fuel channels in the inner region of the core. The small increase in the core reactivity can be managed by slightly increasing the discharge burnup in the non-MPB-bearing fuel channels, thus also improving slightly the fuel utilization in the reactor.

Electrodeposition of Thin Silicon Films for Neutron Transmutation Doping

Thin silicon films were electrodeposited on glassy carbon (GC) from the KF-KCl (2:1)—75 mol% KI—1.5 mol% K2SiF6 melt under potentiostatic condition at 973 K. The synthesized films were single-phase, continuous, dense, and free from unwanted impurities. Neutron transmutation doping (NTD) of the samples was performed in the IVV-2M research reactor (RR) at a thermal neutron flux density of 1.8 × 1013 cm−2s−1 for 7.7 h in order to form the 31P isotope dopant. The irradiated samples were studied by scanning electron microscopy with energy-dispersive X-ray spectroscopy, X-ray diffraction, mass spectrometry, and gamma-ray spectrometry. Some excess of the minimum significant specific activity of the irradiated samples was explained by the formation of the 182Ta isotope due to the presence of tantalum traces in the GC substrate. The formation of the 31P isotope by the NTD process was confirmed. The calculated values of 31P concentration and electrical resistivity were 4.9 × 1016 cm–3 and 0.15 Ω·cm, respectively.

Optimization of conditions for topaz irradiation in the WWR-K reactor

Abstract Activation of impurities in topazes happens due to irradiation with thermal neutrons (induced radioactivity occurs), which complicates their further handling. This leads to the need for their long cooling for safe use. The Institute of Nuclear Physics (Kazakhstan) is conducting R&amp;D to develop a method for the effective formation of color centers in topaz during their irradiation in the WWR-K reactor. An irradiation capsule design has been developed in which optimized conditions for irradiating stones in the neutron field of the reactor are formed. The capsule uses shielding materials made of boron carbide and tantalum to cut off thermal neutrons, resulting in a reduction in induced radioactivity in topaz. The effectiveness of the irradiation capsule was tested in the core of the critical facility. As a result, thermal neutron flux is reduced by 5.7 times and the induced activity of the tantalum is reduced by 2.2 times.

Neutronic Analysis of Accident-Tolerant Cladding Materials in 3D Full Core BEAVRS PWR Benchmark Using OpenMC Code

After the Fukushima Daiichi nuclear accident in 2011, the performance of nuclear fuel during accidents became a matter of great concern. To address this, a new type of fuel technology called accident-tolerant fuel (ATF) has been developed with the goal of enhancing the ability of light water reactors (LWRs) to withstand severe accident conditions. Iron-based alloys have been suggested as potential candidates for fuel cladding due to their favourable thermomechanical properties, lower reactivity with steam, and lower hydrogen generation. This study evaluates the neutronic performance of C26M (a 2nd generation nuclear grade FeCrAl alloy), APMT™, 310SS, and 304SS cladding materials by comparing them with Zircaloy-4 cladding in a 3D PWR core at the beginning of the cycle (BOC) using OpenMC code. The results revealed that the neutronic penalty varied for different alternative cladding materials where C26M exhibited the lowest neutronic penalty value of -12551 pcm, while 310SS demonstrated the highest with a value of -17855 pcm. Additionally, important parameters in the reactor core such as neutron spectrum, reactivity coefficients, boron worth, control rod bank worth, power distribution, and radial thermal neutron flux distribution are evaluated and discussed. The analysis results showed that C26M provided a significantly higher level of neutronic performance compared to APMT™, 304SS, and 310SS. Although this study primarily focused on the neutronic performance of PWRs at BOC, future research should encompass fuel depletion analysis to delve deeper into the potential of alternative cladding materials.

Design and simulation of neutron radiography system for an aqueous homogeneous solution reactor

AHR reactors are known for the production of radiopharmaceuticals. The ARGUS reactor is one of the most famous of these reactors, which has been designed originally for radioisotopes production. The ARGUS reactor produces a suitable neutron flux that can be used for other applications such as neutron radiography (NR). Here, a NR system is designed step by step using the Monte Carlo method. The length of the collimator, divergence angle, diameter of the aperture, and thermal column dimension are designed. The designed aperture and thermal column are selected from boron carbide and heavy water, respectively. The depth of the thermal column is changed until the ratio of thermal neutrons to total neutrons (TNC) reached above 90%. The hole inside the thermal column is designed and internal lining is done. The inner lining is made of cadmium and had a thickness of 1 mm, and the space between the inner and outer lining is filled with lead. To reduce the gamma dose to acceptable values, the position of the collimator is changed instead of using a gamma filter. By changing the depth of the hole and creating the convergence angle to the hole, the flux of thermal neutrons will be 1.8E+6 n.cm−2s−1 that satisfy the IAEA suggested values for NR. Then, the uniformity of the radiation field of the neutron flux output from the collimator is obtained. At the end, images of SI and BPI standards are recorded. The good quality and contrast of these images showed the acceptable design of the NR system for the ARGUS reactor.

A re-activation model for 169Yb intensity modulated brachytherapy sources accounting for spatiotemporal isotopic composition.

Intensity modulated brachytherapy based on partially shielded intracavitary and interstitial applicators is possible with a cost-effective 169Yb production method. 169Yb is a traditionally expensive isotope suitable for this purpose, with an average γ-ray energy of 93keV. Re-activating a single 169Yb source multiple times in a nuclear reactor between clinical uses was shown to theoretically reduce cost by approximately 75% relative to conventional single-activation sources. With re-activation, substantial spatiotemporal variation in isotopic source composition is expected between activations via 168Yb burnup and 169Yb decay, resulting in time dependent neutron transmission, precursor usage, and reactor time needed per re-activation. To introduce a generalized model of radioactive source production that accounts for spatiotemporal variation in isotopic source composition to improve the efficiency estimate of the 169Yb production process, with and without re-activation. A time-dependent thermal neutron transport, isotope transmutation, and decay model was developed. Thermal neutron flux within partitioned sub-volumes of a cylindrical active source was calculated by raytracing through the spatiotemporal dependent isotopic composition throughout the source, accounting for thermal neutron attenuation along each ray. The model was benchmarked, generalized, and applied to a variety of active source dimensions with radii ranging from 0.4 to 1.0mm, lengths from 2.5 to 10.5mm, and volumes from 0.31 to 7.85 mm3, at thermal neutron fluxes from 1×1014to1×1015 n cm-2 s-1. The 168Yb-Yb2O3 density was 8.5g cm-3 with 82% 168Yb-enrichment. As an example, a reference re-activatable 169Yb active source (RRS) constructed of 82%-enriched 168Yb-Yb2O3 precursor was modeled, with 0.6mm diameter, 10.5mm length, 3 mm3 volume, 8.5g cm-3 density, and a thermal neutron activation flux of 4 × 1014 neutronscm-2s-1. The average clinical 169Yb activity for a 0.99 versus 0.31 mm3 source dropped from 20.1 to 7.5Ci for a 4×1014ncm-2s-1 activation flux and from 20.9 to 8.7Ci for a 1×1015ncm-2s-1 activation flux. For thermal neutron fluxes ≥2×1014ncm-2s-1, total precursor and reactor time per clinic-year were maximized at a source volume of 0.99 mm3 and reached a near minimum at 3mm3. When the spatiotemporal isotopic composition effect was accounted for, average thermal neutron transmission increased over RRS lifetime from 23.6% to 55.9%. A 28% reduction (42.5 days to 30.6 days) in the reactor time needed per clinic-year for the RRS is predicted relative to a model that does not account for spatiotemporal isotopic composition effects. Accounting for spatiotemporal isotopic composition effects within the RRS results in a 28% reduction in the reactor time per clinic-year relative to the case in which such changes are not accounted for. Smaller volume sources had a disadvantage in that average clinical 169Yb activity decreased substantially below 20Ci for source volumes under 1mm3. Increasing source volume above 3 mm3 adds little value in precursor and reactor time savings and has a geometric disadvantage.