Monte Carlo Radiation Transport Research Articles

Tritiated Steel Micro-Particles: Computational Dosimetry and Prediction of Radiation-Induced DNA Damage for In Vitro Cell Culture Exposures.

Biological effects of radioactive particles can be experimentally investigated in vitro as a function of particle concentration, specific activity and exposure time. However, a careful dosimetric analysis is needed to elucidate the role of radiation emitted by radioactive products in inducing cyto- and geno-toxicity: the quantification of radiation dose is essential to eventually inform dose-risk correlations. This is even more fundamental when radioactive particles are short-range emitters and when they have a chemical speciation that might further concur to the heterogeneity of energy deposition at the cellular and sub-cellular level. To this aim, we need to use computational models. In this work, we made use of a Monte Carlo radiation transport code to perform a computational dosimetric reconstruction for in vitro exposure of cells to tritiated steel particles of micrometric size. Particles of this kind have been identified as worth of attention in nuclear power industry and research: tritium easily permeates in steel elements of nuclear reactor machinery, and mechanical operations on these elements (e.g., sawing) during decommissioning of old facilities can result in particle dispersion, leading to human exposure via inhalation. Considering the software replica of a representative in vitro setup to study the effect of such particles, we therefore modelled the radiation field due to the presence of particles in proximity of cells. We developed a computational approach to reconstruct the dose range to individual cell nuclei in contact with a particle, as well as the fraction of "hit" cells and the average dose for the whole cell population, as a function of particle concentration in the culture medium. The dosimetric analysis also provided the basis to make predictions on tritium-induced DNA damage: we estimated the dose-dependent expected yield of DNA double strand breaks due to tritiated steel particle radiation, as an indicator of their expected biological effectiveness.

Layered CAD/CSG geometry for spatially complex radiation transport scenarios

Many spatially complex fission, fusion, and national security Monte Carlo (MC) radiation transport scenarios involve combining computer-aided design (CAD) models with constructive solid geometry (CSG) models. A layered geometry method has been implemented in the Shift MC code to address this need. With layered geometry, multiple CAD and/or CSG models can be clipped, translated, rotated, and placed in overlapping layers to form transport-ready geometries. The utility of this method is demonstrated with two problems: (1) a fixed-source simulation with a layered geometry consisting of a LiDAR-generated CAD model of the Combined Arms Collective Training Facility urban environment overlaid with CSG models of a mock hotel and a detector apparatus, and (2) a k-eigenvalue calculation using a layered geometry model of the Transformational Challenge Reactor consisting of CAD fuel elements placed in a CSG core. Tallied particle flux distributions match expectations, but tracking robustness must be improved prior to general-purpose use.

Boron-loaded deuterated liquid scintillator response characterization for neutron spectroscopy

This research presents the characterization methods and results for determining the pulse-integral response calibration, resolution, efficiency, and heavy charged particle light yield functions of boron-loaded deuterated liquid scintillators (BLDLS) through a combination of experimental neutron and gamma-ray measurements coupled with GEometry ANd Tracking (Geant4) Monte Carlo radiation transport simulations. These methods were used to characterize the scintillator light output, pulse-shape discrimination performance, and response matrix for a boron-loaded deuterated liquid scintillator to be used for future neutron spectrum unfolding measurements. These results were compared against an un-loaded deuterated liquid scintillator and a commercially-available hydrogen-based liquid scintillator, EJ-309, to benchmark the performance of the methods and scintillators to comparable measurements in literature. EJ-309 was shown to have only a 3.5 ± 2.3% and 8.1 ± 1.4% higher relative light output at 477 keV compared to Tol-D8 and Tol-D8−10B, respectively, a significant increase in performance over previous BLDLS results. Additionally, Tol-D8 and Tol-D8−10B have similar PSD performance, both exceeding that of EJ-309 as a function of light yield. Finally, full detector response matrices, used for neutron spectrum unfolding, were generated for each scintillator, and the condition of these response matrices were evaluated. The deuterated scintillators resulted in worse conditioned response matrices compared to EJ-309 for neutron energies ranging between 0.5 MeV to 6 MeV, primarily due to having a lower pulse-integral resolution at the smaller detector sizes used in this work. The boron-loaded scintillator shows a similarly conditioned response matrix to the un-loaded variant, thereby lowering the characterizable neutron energy threshold to thermal energies while also enhanced the neutron detection efficiency.

Friday, June 17, 20222:30 PM - 3:30 PM PHSPP01 Presentation Time: 2:30 PM: Dosimetry for Beta Emitting Brachytherapy Sources

<h3>Purpose</h3> Dosimetry for photon-emitting brachytherapy sources is well established, to the point of commercial software for clinical treatment planning to account for material heterogeneities and patient-specific radiation scatter conditions. However, dosimetry standards for beta sources are simpler and have not been recently investigated. While the AAPM TG-149 and ICRU 72 Reports issue clinical guidance and reference data, a modern evaluation of beta-source dosimetry is missing. This investigation strives to identify the potential role for beta-emitting radionuclides as well as to examine theoretical sources for a deeper understanding of beta dosimetry. <h3>Materials and Methods</h3> Available beta-emitting brachytherapy sources include P-32, Sr/Y-90, and Ru/Rh-106 with Y-90 commonly used for selective internal radiation therapy (SIRT) of cancers in the liver. Energy spectra of these sources were taken from ICRU 72 with dose simulations performed using the MCNP v6.2 Monte Carlo radiation transport code. From ICRU 72, the maximum energies for P-32, Sr/Y-90, Y-90, and Ru/Rh-106 were 1.71, 2.28, 2.28, and 3.54 MeV, respectively. Monoenergetic sources of 0.5, 1, 1.5, 2, 3, and 4 MeV were also simulated to extend the energy range and scrutinize dose-distribution dependence on electron energy. Since 2D dose distribution are highly dependent on the design of a source, point sources were simulated to evaluate the depth-dose distribution as a function of radius, <i>r</i>, as well radial dose functions that removed the inverse-square effect. Dose was sampled in 15 cm diameter spheres comprised of water, tissue (ICRU 44 skeletal muscle), or bone (ICRU 44 cortical bone). To mass density effect was quantified in water for 0.5, 1, and 2 g/cm³ with 1 and 2 g/cm³ for tissue and bone, respectively. Observed dose differences were then ascribed to medium differences and not confounded by small differences in mass density. Dose was sampled with radial increments from 0.001 to 0.02 cm for direct comparisons to ICRU 72 results and to evaluate the effect of volume averaging. While only electron sources were considered, dose contributions from bremsstrahlung photons were quantified through comparing results of electron- and electron:photon-transport. Energy deposition was calculated using the MCNP *F8 pulse-height tally, then dividing by the voxel mass to obtain absorbed dose per electron history. With 10⁷ source electrons, statistical uncertainties (<i>k</i>=1) were <0.1% for <i>r</i><0.7 CSDA. When studying bremsstrahlung, 10⁸ source electrons produced 4% statistical uncertainties for <i>r</i>>1.2 CSDA. <h3>Results</h3> In the high dose-gradient region for radionuclide-based beta emitters, the calculated dose distributions in water were within 20% of those from the ICRU 72 Report, which used binned monoenergetic sources to approximate the continuous beta spectra. The radionuclide-based sources exhibited radial dose functions in water that diminished much more steeply than Pd-103, similar to a <0.01 MeV photon source. Sr/Y-90 had a bimodal distribution. Radial dose functions for radionuclide-based sources were inversely proportional to changes in mass density, differing from the scaling-factor models by Cross (1967) and the ICRU 56 Report over the large range of media and densities examined herein. Further, radial dose functions for monoenergetic sources demonstrated a Bragg peak with dose enhancement at <i>r</i>∼0.7 CSDA for all studied energies and media. A sampling of results for radionuclides and monoenergetic sources in water is given in the figure. <h3>Conclusions</h3> Beta dosimetry is largely governed by the mass density with medium composition being a much lesser effect, unlike for low-energy photons where the photoelectric effect is highly sensitive to <i>Z</i>_eff. Unlike for high-energy photon dosimetry, beta dosimetry is negligibly impacted by radiation scatter conditions due to their short range. While Ru/Rh-106 and Sr/Y-90 are proven for shallow plaque brachytherapy and Y-90 for SIRT, beta-emitting radionuclides do not have adequate range for interstitial implants like prostate brachytherapy since dose uniformity is much worse than with HDR Ir-192 or LDR Pd-103 brachytherapy.

CT organ dose calculator size adaptive for pediatric and adult patients

Background. Although computed tomography (CT) has played a critical role in medical care since its introduction in the 1970s, its potential long-term risk of adverse health effects has been of concern. It is crucial to accurately estimate the radiation dose delivered to the patient’s critical organs to ensure the dose is As Low As Reasonably Achievable. However, organ-level dose calculation tools for pediatric and adult patients with various body sizes are rare. We extended the existing CT organ dose calculator, NCICT 1.0, which is based on reference-size phantoms, to include body size-specific pediatric and adult phantoms. Methods. We calculated body size-specific organ doses normalized to CTDIvol by using a library of 158 pediatric and 193 adult computational human phantoms with various body sizes combined with a Monte Carlo radiation transport code, MCNP6. We also created a library of generic tube current modulation (TCM) profiles for the phantom library using a ray-tracing algorithm and implemented them into organ dose calculations. We validated organ doses for the body size-specific phantoms using those calculated from ten abdominal CT patients. We also evaluated potential dosimetric errors caused by only using reference phantoms for patients with different body sizes. Results. Organ dose coefficients and TCM profiles for 351 pediatric and adult body size-specific phantoms were implemented into NCICT 2.0. The dose coefficients from the ten abdominal CT patients agreed with those from the program within 13%. The organ doses for the overweight phantoms were overestimated by over 80% when only reference size phantoms were used. Conclusion. We confirmed that the upgraded dose calculator NCICT 2.0 could substantially reduce potential dosimetric errors caused by using only reference size phantoms. The program should be useful for the radiology community to accurately monitor organ doses for pediatric and adult CT patients with various body sizes.

Evaluation of different types of lithium fluoride thermoluminescent detectors for ring dosimetry in nuclear medicine

Nuclear medicine staff is primarily exposed to their fingers during radionuclide manipulation. About 20% of these workers have risk to exceed the hand dose limit. Therefore, monitoring with ring dosimeters in terms of Hp(0.07) is advised. Most ring dosimeters make use of lithium fluoride (LiF) thermoluminescent detectors (TLDs). Different types of LiF TLDs exist, each with their advantages and disadvantages. The goal of this work was to evaluate the performance of three different LiF TLD types (MCP-N, MTS-N and MCP-Ns) for ring dosimetry during manipulation of the more novel radionuclides Lu-177, Ga-68 and I-131 within radionuclide imaging and therapy, and to provide recommendations for the most accurate dosimetric follow-up of the dose to the hands. This was done by combining measurements with TLDs and theoretical modelling of the TLD responses using Monte Carlo radiation transport simulations and microdosimetric modelling of the TLD relative luminescence efficiency. It was concluded that MTS-N TLDs are best suited for measuring doses from Lu-177 and I-131 if more than 1 mm polycarbonate equivalent shielding is provided. However, MCP-N TLDs have less than 20% under-response and are thus also acceptable in case high sensitivity is important. MCP-Ns TLDs were found to be best suited for measuring doses from Ga-68, and from Lu-177 and I-131 if less than 1 mm polycarbonate equivalent shielding is provided. The recommended TLD type strongly depends on both the manipulated radionuclide and the amount of shielding between the radionuclides and the fingers (syringe thickness and potential additional shielding around the syringe). But the approach presented here for the theoretical modelling of the Hp(0.07) response can easily be applied to select the most appropriate TLD type for other radionuclides or other amounts of shielding.

A Review of Computational Phantoms for Quality Assurance in Radiology and Radiotherapy in the Deep-Learning Era

The exciting advancement related to the “modeling of digital human” in terms of a computational phantom for radiation dose calculations has to do with the latest hype related to deep learning. The advent of deep learning or artificial intelligence (AI) technology involving convolutional neural networks has brought an unprecedented level of innovation to the field of organ segmentation. In addition, graphics processing units (GPUs) are utilized as boosters for both real-time Monte Carlo simulations and AI-based image segmentation applications. These advancements provide the feasibility of creating three-dimensional (3D) geometric details of the human anatomy from tomographic imaging and performing Monte Carlo radiation transport simulations using increasingly fast and inexpensive computers. This review first introduces the history of three types of computational human phantoms: stylized medical internal radiation dosimetry (MIRD) phantoms, voxelized tomographic phantoms, and boundary representation (BREP) deformable phantoms. Then, the development of a person-specific phantom is demonstrated by introducing AI-based organ autosegmentation technology. Next, a new development in GPU-based Monte Carlo radiation dose calculations is introduced. Examples of applying computational phantoms and a new Monte Carlo code named ARCHER (Accelerated Radiation- transport Computations in Heterogeneous EnviRonments) to problems in radiation protection, imaging, and radiotherapy are presented from research projects performed by students at the Rensselaer Polytechnic Institute (RPI) and University of Science and Technology of China (USTC). Finally, this review discusses challenges and future research opportunities. We found that, owing to the latest computer hardware and AI technology, computational human body models are moving closer to real human anatomy structures for accurate radiation dose calculations.

Advances in personal dosimetry towards real-time dosimetry

This objective of this paper is to give a short overview of the state-of-the-art of personal dosimetry, more specifically focussing on the advances towards real-time dosimetry. The use of active personal dosemeters has been common practice since a few decades, especially in the nuclear industry, but the large majority of the workers are still using passive personal dosemeters to monitor their doses and compare them to the legal dose limits. Besides passive personal dosimeters, many types of active personal dosemeters (APDs) exist that have equally good, or even better technical characteristics than those of passive dosemeters. Although improvements are also still needed for APDs (like for beta radiation and pulsed fields), the reason that they are not more often used for legal dosimetry has more to do with non-technical reasons, like the need to set-up a QC/QA system, and a higher cost. Recently new types of dosemeters have appeared on the market that are a mix between active and passive dosemeters. They can find their way to widespread use by being small and relatively cheap, while still offering some advantages for ALARA applications. The newest developments in personal dosimetry go in the direction of computational techniques. By using motion capture with cameras combined with fast Monte Carlo radiation transport simulations, the doses of workers can be calculated in near real time if the radiation source is known. So there are some exciting times ahead for personal dosimetry.

Simulation and experimental verification of ambient neutron doses in a pencil beam scanning proton therapy room as a function of treatment plan parameters

Out-of-field patient doses in proton therapy are dominated by neutrons. Currently, they are not taken into account by treatment planning systems. There is an increasing need to include out-of-field doses in the dose calculation, especially when treating children, pregnant patients, and patients with implants. In response to this demand, this work presents the first steps towards a tool for the prediction of out-of-field neutron doses in pencil beam scanning proton therapy facilities. As a first step, a general Monte Carlo radiation transport model for simulation of out-of-field neutron doses was set up and successfully verified by comparison of simulated and measured ambient neutron dose equivalent and neutron fluence energy spectra around a solid water phantom irradiated with a variation of different treatment plan parameters. Simulations with the verified model enabled a detailed study of the variation of the neutron ambient dose equivalent with field size, range, modulation width, use of a range shifter, and position inside the treatment room. For future work, it is planned to use this verified model to simulate out-of-field neutron doses inside the phantom and to verify the simulation results by comparison with previous in-phantom measurement campaigns. Eventually, these verified simulations will be used to build a library and a corresponding tool to allow assessment of out-of-field neutron doses at pencil beam scanning proton therapy facilities.

Experimental validation of Monte Carlo dosimetry for therapeutic beta emitters with radiochromic film in a 3D-printed phantom.

Validation of dosimetry software, such as Monte Carlo (MC) radiation transport codes used for patient-specific absorbed dose estimation, is critical prior to their use in clinical decision making. However, direct experimental validation in the clinic is generally not performed for low/medium-energy beta emitters used in radiopharmaceutical therapy (RPT) due to the challenges of measuring energy deposited by short-range particles. Our objective was to design a practical phantom geometry for radiochromic film (RF)-based absorbed dose measurements of beta-emitting radionuclides and perform experiments to directly validate our in-house developed Dose Planning Method (DPM) MC code dedicated to internal dosimetry. The experimental setup was designed for measuring absorbed dose from beta emitters that have a range sufficiently penetrating to ∼200 μm in water as well as to capture any photon contributions to absorbed dose. Assayed 177 Lu and 90 Y liquid sources, 13-450 MBq estimated to deliver 0.5-10 Gy to the sensitive layer of the RF, were injected into the cavity of two 3D-printed half-cylinders that had been sealed with 12.7 μm or 25.4 μm thick Kapton Tape. A 3.8 × 6 cm strip of GafChromic EBT3 RF was sandwiched between the two taped half-cylinders. After 2-48 h exposures, films were retrieved and wipe tested for contamination. Absorbed dose to the RF was measured using a commercial triple-channel dosimetry optimization method and a calibration generated via 6 MV photon beam. Profiles were analyzed across the central 1 cm2 area of the RF for validation. Eleven experiments were completed with 177 Lu and nine with 90 Y both in saline and a bone equivalent solution. Depth dose curves were generated for 177 Lu and 90 Y stacking multiple RF strips between a single filled half-cylinder and an acrylic backing. All experiments were modeled in DPM to generate voxelized MC absorbed dose estimates. We extended our study to benchmark general purpose MC codes MCNP6 and EGSnrc against the experimental results as well. A total of 20 experiments showed that both the 3D-printed phantoms and the final absorbed dose values were reproducible. The agreement between the absorbed dose estimates from the RF measurements and DPM was on average -4.0% (range -10.9% to 3.2%) for all single film 177 Lu experiments and was on average -1.0% (range -2.7% to 0.7%) for all single film 90 Y experiments. Absorbed depth dose estimates by DPM agreed with RF on average 1.2% (range -8.0% to 15.2%) across all depths for 177 Lu and on average 4.0% (range -5.0% to 9.3%) across all depths for 90 Y. DPM absorbed dose estimates agreed with estimates from EGSnrc and MCNP across the board, within 4.7% and within 3.4% for 177 Lu and 90 Y respectively, for all geometries and across all depths. MC showed that absorbed dose to RF from betas was greater than 92% of the total (betas + other radiations) for 177 Lu, indicating measurement of dominant beta contribution with our design. The reproducible results with a RF insert in a simple phantom designed for liquid sources demonstrate that this is a reliable setup for experimentally validating dosimetry algorithms used in therapies with beta-emitting unsealed sources. Absorbed doses estimated with the DPM MC code showed close agreement with RF measurement and with results from two general purpose MC codes, thereby validating the use of this algorithms for clinical RPT dosimetry.

Electron backscattering for signal enhancement in a thin-film CdTe radiation detector.

Thin-film cadmium telluride (CdTe) offers high average electron density, direct detection configuration, and excellent radiation hardness, making it an attractive material for radiation detectors. Although a very thin detector provides capabilities to conduct high-resolution measurements in high-energy radiation fields, it is limited by a low signal, often boosted with a front metal converter enhancing X-ray absorption. An extension of this approach can be explored through the investigation of electron backscattering phenomenon, known to be highly dependent on the material atomic number Z. Adding an electron reflector in tandem with the back electrode is proposed to be utilized for the detector signal enhancement. We investigated the possibility of augmenting the fluence of electrons traversing CdTe thin film and thus increasing the detected signal pursuing two pathways: (1) adding a high-Z metal layer to the back of the detector surface, and (2) adding a top low-Z material to the detector layer to return its backscattered electrons. Copper (Cu) and lead (Pb) layers of varying thickness were investigated as potential metal back-reflectors, whereas polymethyl methacrylate (PMMA) water phantom material was tested as the top cover in multilayer detector structures. The Monte Carlo (MC) radiation transport package MCNP5 was first used to model a basic multilayer structure of a CdTe-sensitive volume surrounded by PMMA, under a 6-MV photon beam. Addition of Cu or Pb back-reflectors allowed for the analysis of the signal enhancement and associated changes in Compton electrons fluence spectra. Related backscattering coefficients were then calculated using EGSnrc MC user-code for monoenergetic electron sources. Analytical functions were established to represent the best fitting curves to the simulation data. Finally, electron backscattering data was related to signal enhancement in the CdTe sensitive layer based on a semiquantitative approach. We studied multilayer detector structures, decoupling the effects of PMMA and the back-reflector metals, Cu or Pb, on electron backscattering for electron energy range of up to 500keV or 1MeV depending on the choice of metal. Adding a 100-200-µm-thick metal film below the detector sensitive volume increased the fraction of reflected electrons, especially in the low, 100-200keV, energy range. The thickness dependence of backscattering coefficients from thin films exhibits saturations at values significantly exceeding the electron ranges. That effect was related to the large-angle electron scattering. A detailed simulation of energy deposition revealed that the modified structures using Cu and Pb increased energy deposition by ∼10% and 75%, respectively. We have also established a linear dependence between the energy deposition in the semiconductor layer and the fluence of backscattered electrons in the corresponding multilayer structure. The low-Z top layer in practically implemental thicknesses of tens of micrometers has a positive effect due to partial electron reflection back to the semiconductor layer. Signal enhancement in a thin-film CdTe radiation detector could be achieved using electron backscattering from metal reflectors. The methodology explored here warrants further studies to quantify achievable signal enhancement for various thin films and other small sensitive volume detectors.

Attenuation of beta radiation in granular matrices: implications for trapped-charge dating

Abstract. Mineral grains within sediment or rock absorb a radiation dose from the decay of radionuclides in the host matrix. For the beta dose component, the estimated dose rate must be adjusted for the attenuation of beta particles within the mineral grains. Standard calculations, originally designed for thermoluminescence dating of pottery, assume that the grain is embedded in a homogenous medium. However, most current applications of trapped-charge dating concern sand- or silt-sized dosimeters embedded in granular sediment. In such cases, the radionuclide sources are not homogeneous but are localised in discrete grains or held on grain surfaces. We show here that the mean dose rate to dosimeter grains in a granular matrix is dependent on the grain-size distributions of the source grains and of the bulk sediment, in addition to the grain size of the dosimeters. We further argue that U and Th sources are likely to be held primarily on grain surfaces, which causes the dose rate to dosimeter grains to be significantly higher than for sources distributed uniformly throughout grains. For a typical well-sorted medium sand, the beta dose rates derived from surface U and Th sources are higher by ∼ 20 % and ∼ 30 %, respectively, compared to a homogenous distribution of sources. We account for these effects using an expanded model of beta attenuation – including the effect of moisture – and validate the model against Monte Carlo radiation transport simulations within a geometry of packed spheres.

KDSource, a tool for the generation of Monte Carlo particle sources using kernel density estimation

Monte Carlo radiation transport simulations have clearly contributed to improve the design of nuclear systems. When performing in-beam or shielding simulations a complexity arises due to the fact that particles must be tracked to regions far from the original source or behind the shielding, often lacking sufficient statistics. Different possibilities to overcome this problem such as using particle lists or generating synthetic sources have already been reported. In this work we present a new approach by using the adaptive multivariate kernel density estimator (KDE) method. This concept was implemented in KDSource, a general tool for modelling, optimizing and sampling KDE sources, which provides a convenient user interface. The basic properties of the method were studied in an analytical problem with a known density distribution. Furthermore, the tool was used in two Monte Carlo simulations that modelled neutron beams, which showed good agreement with experimental results.

Accelerated radiation transport modeling techniques for pencil beam computed tomography using gamma rays

Monte Carlo radiation transport modeling studies were performed for a compact, and high-resolution gamma-ray computed tomography system designed for imaging irradiated nuclear fuel. The system comprises a 60Co source – chosen for its highly penetrating 1173 keV and 1332 keV gamma rays – a pair of high-aspect-ratio pencil beam collimators, and an inorganic scintillator detector. Two acceleration methods are proposed to rapidly model a transmission type gamma-ray tomography system. The first, a variance reduction technique, is based on performing Monte Carlo simulations with a monodirectionally-biased source, sampled from a characteristic sub-volume of the full source volume. The second acceleration method is based on the deterministic calculations using the Beer–Lambert law and detector response characteristics. Comparison of simulations using acceleration approaches with analog simulations of the fully isotropic, full-volume equivalent, show that the Monte Carlo variance reduction technique gives quantitatively accurate predictions for large collimator aspect ratios while the deterministic calculations are semi-quantitative but converge close to the correct result as the collimator aspect ratio increases. As such, these techniques can be used to reduce the computational cost in generating simulated radiographs and tomographs by several orders of magnitude. Experimental validation efforts are currently underway and will be demonstrated in future work.

GATE/Geant4-based dosimetry for ex vivo in solution irradiation of blood with radionuclides

To establish a dose-response relationship between radiation-induced DNA damage and the corresponding absorbed doses in blood irradiated with radionuclides in solution under ex vivo conditions, the absorbed dose coefficient for 1 ml for 1 h internal ex vivo irradiation of peripheral blood (dBlood) must be determined. dBlood is specific for each radionuclide, and it depends on the irradiation geometry. Therefore, the aim of this study is to use the Monte Carlo radiation transport code GATE/Geant4 to calculate the mean absorbed dose rates for ex vivo irradiation of blood with several radionuclides used in Nuclear Medicine. MethodsThe Monte Carlo simulation reproduces the irradiation geometry of a blood sample of 7 ml mixed with 1 ml of a water equivalent radioactive solution in an 8 ml vial. The simulation was performed for ten different radionuclides: 18F, 68Ga, 90Y, 99mTc, 123I, 124I, 131I, 177Lu, 223Ra, and 225Ac. Two sets of simulations for each radionuclide were performed with 1x109 histories. The first set was simulated with a mass density of 1.0525 g/cm3 of the blood plus water mixture. The second set of simulations was performed with a mass density of 1 g/cm3 for comparison with previous studies. ResultsThe values of dBlood for ten radionuclides were calculated. The values range from 10.23 mGy∙ml∙MBq−1 for 99mTc to 15632.02 mGy∙ml∙MBq−1 for 225Ac. The maximum relative change compared to previous studies was 13.0% for 124I. ConclusionThis study provides a comprehensive set of absorbed dose coefficients for 1 ml for 1 h internal ex vivo irradiation of peripheral blood in a special vial geometry and radionuclides typically used in Nuclear Medicine. Furthermore, the method proposed by this work can be easily adapted to a variety of internal irradiation conditions and serve as a reference for future studies.

Neutronic analysis of Indian helium-cooled solid breeder tritium breeding module for testing in ITER

India has proposed the helium-cooled solid breeder blanket concept as a tritium breeding module to be tested in ITER. The module has lithium titanate for tritium breeding and beryllium for neutron multiplication. Beryllium also enhances tritium breeding. A design for the module is prepared for detailed analysis. Neutronic analysis is performed to assess the tritium breeding rate, neutron distribution, and heat distribution in the module. The tritium production distribution in submodules is evaluated to support the tritium transport analysis. The tritium breeding density in the radial direction of the module is also assessed for further optimization of the design. The heat deposition profile of the entire module is generated to support the heat removal circuit design. The estimated neutron spectrum in the radial direction also provides a more in-depth picture of the nuclear interactions inside the material zones. The total tritium produced in the HCSB module is around 13.87 mg per full day of operation of ITER, considering the 400 s ON time and 1400 s dwell time. The estimated nuclear heat load on the entire module is around 474 kW, which will be removed by the high-pressure helium cooling circuit. The heat deposition in the test blanket model (TBM) is huge (around 9 GJ) for an entire day of operation of ITER, which demonstrates the scale of power that can be produced through a fusion reactor blanket. As per the Brayton cycle, it is equivalent to 3.6 GJ of electrical energy. In terms of power production, this would be around 1655 MWh annually. The evaluation is carried out using the MCNP5 Monte Carlo radiation transport code and FEDNL 2.1 nuclear cross section data. The HCSB TBM neutronic performance demonstrates the tritium production capability and high heat deposition.

Estimation of radioactivity and dose equivalent rate by combining Compton imaging and Monte Carlo radiation transport code

In a radiation environment, such as the decommissioning site of a nuclear power station, visualization of the distribution of radioactive substances and estimation of the dose equivalent rate around the site can help reduce the exposure dose of workers and plan their work. The author has developed a method of visualizing the existence of a radiation source using a gamma-ray imager, estimating its radioactivity, and estimating the dose equivalent rate around the source. A Compton camera, which is a gamma-ray imager, is used to visualize the existence of a137Cs radiation source and estimate its radioactivity, and a three-dimensional (3D) model of the region around the source is generated using a simultaneous localization and mapping device based on 3D light detection and ranging. Next, the dose equivalent rate around the source is calculated by importing the 3D model data and radioactivity information into a particle and heavy ion transport code system. The validity of the calculated dose equivalent rates is also confirmed by comparing them with values measured using a survey meter. This method can be used not only to simply visualize a source and calculate the dose equivalent rate around it but also to evaluate how addition of shielding or removal of contaminated objects can contribute to reducing the dose equivalent rate.

A Pipeline for Automated Voxel Dosimetry: Application in Patients with Multi-SPECT/CT Imaging After 177Lu-Peptide Receptor Radionuclide Therapy.

Patient-specific dosimetry in radiopharmaceutical therapy (RPT) is impeded by the lack of tools that are accurate and practical for the clinic. Our aims were to construct and test an integrated voxel-level pipeline that automates key components (organ segmentation, registration, dose-rate estimation, and curve fitting) of the RPT dosimetry process and then to use it to report patient-specific dosimetry in 177Lu-DOTATATE therapy. Methods: An integrated workflow that automates the entire dosimetry process, except tumor segmentation, was constructed. First, convolutional neural networks (CNNs) are used to automatically segment organs on the CT portion of one post-therapy SPECT/CT scan. Second, local contour intensity-based SPECT--SPECT alignment results in volume-of-interest propagation to other time points. Third, dose rate is estimated by explicit Monte Carlo (MC) radiation transport using the fast, Dose Planning Method code. Fourth, the optimal function for dose-rate fitting is automatically selected for each voxel. When reporting mean dose, we apply partial-volume correction, and uncertainty is estimated by an empiric approach of perturbing segmentations. Results: The workflow was used with 4-time-point 177Lu SPECT/CT imaging data from 20 patients with 77 neuroendocrine tumors, segmented by a radiologist. CNN-defined kidneys resulted in high Dice values (0.91-0.94) and only small differences (2%-5%) in mean dose when compared with manual segmentation. Contour intensity-based registration led to visually enhanced alignment, and the voxel-level fitting had high R 2 values. Across patients, dosimetry results were highly variable; for example, the average of the mean absorbed dose (Gy/GBq) was 3.2 (range, 0.2-10.4) for lesions, 0.49 (range, 0.24-1.02) for left kidney, 0.54 (range, 0.31-1.07) for right kidney, and 0.51 (range, 0.27-1.04) for healthy liver. Patient results further demonstrated the high variability in the number of cycles needed to deliver hypothetical threshold absorbed doses of 23 Gy to kidney and 100 Gy to tumor. The uncertainty in mean dose, attributable to variability in segmentation, averaged 6% (range, 3%-17%) for organs and 10% (range, 3%-37%) for lesions. For a typical patient, the time for the entire process was about 25 min (∼2 min manual time) on a desktop computer, including time for CNN organ segmentation, coregistration, MC dosimetry, and voxel curve fitting. Conclusion: A pipeline integrating novel tools that are fast and automated provides the capacity for clinical translation of dosimetry-guided RPT.

The design and optimization of a radiation shield for the Tehran Research Reactor's neutron diffraction facility

In this paper, a novel optimal radiation shield for the Tehran research reactor's neutron powder diffraction facility was designed to reduce undesirable neutron and gamma dose rates. A layer-by-layer methodology was considered based on the design approach to achieve a dose rate of less than 30 μSv/hr. MCNPX was used for this purpose as a Monte Carlo radiation transport code. A combination of neutron and gamma absorbers and moderators was employed to create this shield. The results indicated that the shield's six distinct layers met the design objective, totaling 65 cm in thickness. This newly designed shield comprises 15 cm paraffin, 15 cm Portland iron concrete, another 15 cm paraffin, 5 cm solid boric acid, another 10 cm Portland iron concrete, and 10 cm lead. The neutron and photon dose rates are 9.35 and 12.81 μSv/hr, respectively, which is one-sixteenth and one-third of the current facility shield. Validation of the simulation was accomplished by comparing the results for the current shield to experimental data.

Development of a GPU-accelerated Monte Carlo dose calculation module for nuclear medicine, ARCHER-NM: demonstration for a PET/CT imaging procedure

Objective. This paper describes the development and validation of a GPU-accelerated Monte Carlo (MC) dose computing module dedicated to organ dose calculations of individual patients undergoing nuclear medicine (NM) internal radiation exposures involving PET/CT examination. Approach. This new module extends the more-than-10-years-long ARCHER project that developed a GPU-accelerated MC dose engine by adding dedicated NM source-definition features. To validate the code, we compared dose distributions from the point ion source, including 18F, 11C, 15O, and 68Ga, calculated for a water phantom against a well-tested MC code, GATE. To demonstrate the clinical utility and advantage of ARCHER-NM, one set of 18F-FDG PET/CT data for an adult male NM patient is calculated using the new code. Radiosensitive organs in the CT dataset are segmented using a CNN-based tool called DeepViewer. The PET image intensity maps are converted to radioactivity distributions to allow for MC radiation transport dose calculations at the voxel level. The dose rate maps and corresponding statistical uncertainties were calculated at the acquisition time of PET image. Main results. The water-phantom results show excellent agreement, suggesting that the radiation physics module in the new NM code is adequate. The dose rate results of the 18F-FDG PET imaging patient show that ARCHER-NM’s results agree very well with those of the GATE within −2.45% to 2.58% (for a total of 28 organs considered in this study). Most impressively, ARCHER-NM obtains such results in 22 s while it takes GATE about 180 min for the same number of 5 × 108 simulated decay events. Significance. This is the first study presenting GPU-accelerated patient-specific MC internal radiation dose rate calculations for clinically realistic 18F-FDG PET/CT imaging case involving autosegmentation of whole-body PET/CT images. This study suggests that the proposed computing tools—ARCHER-NM— are accurate and fast enough for routine internal dosimetry in NM clinics.