Monte Carlo Radiation Transport Code Research Articles

A new ground level neutron monitor for space weather assessment

We report on a new ground-level neutron monitor design for studying cosmic rays and fluxes of solar energetic particles at the Earth’s surface. The first-of-its-kind instrument, named the NM-2023 after the year it was standardised and following convention, will be installed at a United Kingdom Meteorological Office observatory (expected completion mid 2024) and will reintroduce such monitoring in the UK for the first time since ca. 1984. Monte Carlo radiation transport code is used for the development and application of parameterised models to investigate alternative neutron detectors, their location and bulk material geometry in a realistic cosmic ray neutron field. Benchmarked against a model of the current and most widespread design standardised in 1964 (the NM-64), two main parameterisation studies are conducted; a simplified standard model and a concept slab parameterisation. We show that the NM-64 standard is well optimised for the intended large-diameter boron trifluoride (BF3\\documentclass[12pt]{minimal} \\usepackage{amsmath} \\usepackage{wasysym} \\usepackage{amsfonts} \\usepackage{amssymb} \\usepackage{amsbsy} \\usepackage{mathrsfs} \\usepackage{upgreek} \\setlength{\\oddsidemargin}{-69pt} \\begin{document}$$_3$$\\end{document}) proportional counters but not for multiple smaller diameter counters. The new design (based on a novel slab arrangement) produces comparable counting efficiencies to an NM-64 with six BF3\\documentclass[12pt]{minimal} \\usepackage{amsmath} \\usepackage{wasysym} \\usepackage{amsfonts} \\usepackage{amssymb} \\usepackage{amsbsy} \\usepackage{mathrsfs} \\usepackage{upgreek} \\setlength{\\oddsidemargin}{-69pt} \\begin{document}$$_3$$\\end{document} counters and has the added advantage of being more compact, lower cost and avoids the use of highly toxic BF3\\documentclass[12pt]{minimal} \\usepackage{amsmath} \\usepackage{wasysym} \\usepackage{amsfonts} \\usepackage{amssymb} \\usepackage{amsbsy} \\usepackage{mathrsfs} \\usepackage{upgreek} \\setlength{\\oddsidemargin}{-69pt} \\begin{document}$$_3$$\\end{document}.

Implementation of a Novel Monte Carlo Radiation Transport Code for Modeling Nonthermal Fe K$\alpha$ Line Generation in a Dense MagLIF Plasma

Implementation of a Novel Monte Carlo Radiation Transport Code for Modeling Nonthermal Fe K$\alpha$ Line Generation in a Dense MagLIF Plasma

The influence of Monte Carlo radiation transport codes and hadron physics models on stray neutron dose estimations in proton therapy: An extended intercomparison

PurposeSelecting a Monte Carlo (MC) radiation transport code and nuclear physics model affects the simulation of secondary neutron fluence energy spectra and the estimation of the neutron dose equivalent in proton therapy. This study compares the MC codes FLUKA, MCNPX and GEANT4 from previous work with two additional MC codes, TOPAS (with either the Binary Intra-nuclear Cascade (BIC) or the Bertini intra-nuclear cascade model (Bert)) and PHITS (with either the Intra-Nuclear Cascade of Liege (INCL) or the Bert model), to facilitate a more comprehensive comparison. Material and methodsSecondary neutron fluence energy spectra were scored at different locations inside a water phantom, exposed to a 10 × 10 cm2 parallel mono-energetic proton beam with energies of 110 MeV, 150 MeV, 180 MeV and 210 MeV. The neutron dose equivalent was estimated by applying fluence-to-dose equivalent conversion coefficients. ResultsDiscrepancies between MC codes and physics models reach 90 % for the total neutron fluence. Differences are most pronounced in the forward direction of the beam due to a notable presence of high-energy neutrons and the higher uncertainty on the neutron cross-section data for these neutrons. Nuclear model selection within PHITS (INCL or Bert) and TOPAS (BIC or Bert) shows a worse agreement for the Bert model than for the other MC codes and nuclear models. The discrepancies in the neutron spectrum simulations propagate into differences in the estimated neutron dose equivalents. They are more prominent for distal positions (up to a factor of 3) than for lateral positions (up to a factor of 2) and decrease at higher proton energies. ConclusionsThis study shows considerable differences in the out-of-field neutron fluence energy spectra calculated by different MC codes and nuclear physics models. Differences in the neutron dose equivalent reach a factor of 3. The discrepancies are most apparent in distal positions and at lower proton energies.

EJUSTCO: Monte Carlo radiation transport code hybrid with ANN model for gamma-ray shielding simulation

EJUSTCO: Monte Carlo radiation transport code hybrid with ANN model for gamma-ray shielding simulation

Monte Carlo Radiation Transport for Astrophysical Transients Powered by Circumstellar Interaction

In this paper, we introduce SuperLite, an open-source Monte Carlo radiation transport code designed to produce synthetic spectra for astrophysical transient phenomena affected by circumstellar interaction. SuperLite utilizes Monte Carlo methods for semi-implicit, semirelativistic radiation transport in high-velocity shocked outflows, employing multigroup structured opacity calculations. The code enables rapid post-processing of hydrodynamic profiles to generate high-quality spectra that can be compared with observations of transient events, including superluminous supernovae, pulsational pair-instability supernovae, and other peculiar transients. We present the methods employed in SuperLite and compare the code’s performance to that of other radiative transport codes, such as SuperNu and CMFGEN. We show that SuperLite has successfully passed standard Monte Carlo radiation transport tests and can reproduce spectra of typical supernovae of Type Ia, Type IIP, and Type IIn.

Shutdown dose rate analysis with the Shift Monte Carlo radiation transport code and modular verification workflow

Calculation of the shutdown dose rate is crucial for safe fusion reactor operations. The Rigorous-two-step (R2S) method is a method that requires connected neutron transport, activation, and gamma transport. Shift has integrated variance reduction with a deterministic solver Denovo, supports multiple geometry formats, and is scalable. These features make it an attractive transport solver choice for an R2S workflow. An R2S workflow for the Shift Monte Carlo code is developed and compared to the existing Oak Ridge National Laboratory Shutdown Dose Rate Code Suite (ORCS) workflow. Also, a Python framework for integrating two R2S workflows is developed to mix and match each step in the R2S workflow for improved collaboration and verification experience. Results show that the Shift-Denovo R2S workflow and the ORCS workflow calculate the shutdown dose rate of the ITER Shutdown Dose Rate benchmark problem with an average relative error of 2.285%.

Radiological Characterization with a Fluence Conversion Coefficients–Based Method: A Practical Example of the Preparatory Studies to the Pilot Beam at the CERN Large Hadron Collider

At high-energy accelerator facilities like the ones that are part of the accelerator complex at the European Organization for Nuclear Research (CERN), Monte Carlo radiation transport codes are widely employed to face the challenges of estimating radionuclide production yields and activities with the aim of performing the radiological characterization of activated components. Indeed, it is of paramount importance to ensure adequate radiation protection during scheduled maintenance, transport, and handling of these components and to establish their proper disposal pathway once they ultimately reach the end of their useful life. This paper summarizes the principles of the fluence conversion coefficients method that was developed as a complementary approach for radiological characterization studies. Then, the Monte Carlo simulations in preparation to the pilot beam run at the Large Hadron Collider at CERN in 2021 are presented as a practical example of possible applications. Finally, the flexibility of the method and the most relevant operational radiation protection implications are discussed in relation to the provided example.

Current Status of TRIPOLI-4® Monte Carlo Radiation Transport Code on Adult and Pediatric Computational Phantoms for Radiation Dosimetry Study

TRIPOLI-4® is a general-purpose Monte Carlo radiation transport code developed by the Service d’Études des Réacteurs et de Mathématiques Appliquées at CEA-Saclay. It uses continuous-energy nuclear data to simulate neutron, photon, electron, and positron transport in fields like radiation shielding, reactor physics, and nuclear criticality safety. To study radiation protection dosimetry in human tissues and organs, male and female adult computational phantoms from the Medical Internal Radiation Dose–Oak Ridge National Laboratory phantoms family and the International Commission on Radiological Protection (ICRP) publication 110 were recently modeled and calculated using the geometry options of TRIPOLI-4. To easily use the ICRP 110 voxel-based phantoms in different exposure scenarios, a newly developed phantom option is available in TRIPOLI-4 and its display tool T4G. This new phantom option is helpful for modeling one or more phantoms and for improving calculation performance in real irradiation environments. The 2020 published pediatric computational reference phantoms are accessible from ICRP publication 143. Male and female pediatric phantoms are also verified with the new T4G tool and TRIPOLI-4 code. This paper reports on recent works using TRIPOLI-4 on adult and pediatric computational phantoms. The modeling methods of stylized and voxel-based phantoms, the graphic displays of modeled phantoms with T4G, and the verification procedures for single-phantom and two-phantom application cases are presented. Validation for external and internal dosimetry calculations has been performed. Calculation results on organ dose S values for nuclear medicine applications are presented for single female and single male voxel phantoms using 131I and 177Lu radiation sources. Effective dose calculations for two-phantom cases using 99mTc and 18F sources are compared with traditional H*(10) calculations from nuclear medicine patient to patient caregiver.

PyMCGPU-IR Monte Carlo code test for occupational dosimetry.

PyMCGPU-IR is an innovative occupational dose monitoring tool for interventional radiology procedures. It reads the radiation data from the Radiation Dose Structured Report of the procedure and combines this information with the position of the monitored worker recorded using a 3D camera system. This information is used as an input file for the fast Monte Carlo radiation transport code MCGPU-IR in order to assess the organ doses, Hp(10) and Hp(0.07), as well as the effective dose. In this study, Hp(10) measurements of the first operator during an endovascular aortic aneurysm repair procedure and a coronary angiography using a ceiling suspended shield are compared to PyMCGPU-IR calculations. Differences in the two reported examples are found to be within 15%, which is considered as being very satisfactory. The study highlights the promising advantages of PyMCGPU-IR, although there are still several improvements that need to be implemented before its final clinical use.

Synthesizing Spectra from 3D Radiation Hydrodynamic Models of Massive Stars Using Monte Carlo Radiation Transport

Observations indicate that turbulent motions are present on most massive star surfaces. Starting from the observed phenomena of spectral lines with widths that are much larger than their thermal broadening (e.g., micro- and macroturbulence), and considering the detection of stochastic low-frequency variability (SLFV) in the Transiting Exoplanet Survey Satellite photometry, these stars clearly have large-scale turbulent motions on their surfaces. The cause of this turbulence is debated, with near-surface convection zones, core internal gravity waves, and wind variability being proposed. Our 3D gray radiation hydrodynamic (RHD) models previously characterized the convective dynamics of the surfaces, driven by near-surface convection zones, and provided reasonable matches to the observed SLFV of the most luminous massive stars. We now explore the complex emitting surfaces of these 3D RHD models, which strongly violate the 1D assumption of a plane-parallel atmosphere. By post-processing the gray RHD models with the Monte Carlo radiation transport code Sedona, we synthesize stellar spectra and extract information from the broadening of individual photospheric lines. The use of Sedona enables the calculation of the viewing angle and temporal dependence of spectral absorption line profiles. By combining uncorrelated temporal snapshots together, we compare the turbulent broadening from the 3D RHD models to the thermal broadening of the extended emitting region, showing that our synthesized spectral lines closely resemble the observed macroturbulent broadening from similarly luminous stars. More generally, the new techniques that we have developed will allow for systematic studies of the origins of turbulent velocity broadening from any future 3D simulations.

Monte Carlo simulation of shielding designs for a cabinet form factor preclinical MV-energy photon FLASH radiotherapy system.

A preclinical MV-energy photon FLASH radiotherapy system is being designed at Stanford and SLAC National Accelerator Laboratory. Because of the higher energy and dose rate compared to conventional kV-energy photon laboratory-scale irradiators, adequate shielding in a stand-alone cabinet form factor is more challenging to achieve. We present a Monte Carlo simulation of multilayered shielding for a compact self-shielding system without the need for a radiation therapyvault. A multilayered shielding approach using multiple alternating layers of high-Z and low-Z materials is applied to the self-shielded cabinet to effectively mitigate the secondary radiation produced and to allow the device to be housed in a Controlled Radiation Area outside of a radiation vault. The multilayered shielding approach takes advantage of the properties of high-Z and low-Z radiation shielding materials such as density, cross-section, atomic number of the shielding elements, and products of radiation interactions within each layer. The Monte Carlo radiation transport code, FLUKA, is used to simulate the total effective dose produced by theoperation. The multilayered shielding designs proposed and simulated produced effective dose rates significantly lower than monolayer designs with the same total material thickness at the regulatory boundary; this is accomplished through the manipulation of the locations where secondary radiation is produced and reactions due to material properties such as neutron back reflection in hydrogen. Borated polyethylene at five weight percent significantly increased the shielding performance as compared to regular polyethylene, with the magnitude of the reduction depending upon the order of the shieldingmaterial. The multilayered shielding provides a path for shielding preclinical FLASH systems that deliver MV-energy bremsstrahlung photons. This approach promises to be more efficient with respect to the shielding material mass and space claim as compared to shielded vaults typically required for clinical radiation therapy with MVphotons. This article is protected by copyright. All rights reserved.

Source Nuclide Density First-Order and Second-Order Sensitivities in Monte Carlo Codes

Application of perturbation capabilities for density sensitivities in Monte Carlo radiation transport codes has been limited because changing source nuclide densities or source material densities changes the intrinsic source, and in most Monte Carlo codes, the user-input source is independent of the user-input materials. The perturbation capability then has no way of accounting for changes in the intrinsic source. This paper derives the sensitivity of a response with respect to a source nuclide density in terms of a portion due to the transport operator and a portion due to the source rate density. The Monte Carlo perturbation method computes the portion due to the transport operator, and the portion due to the source rate density is computed in postprocessing using parameters from the precomputed intrinsic source calculation. This paper derives first- and second-order sensitivities. The equations require the response to be separated by contribution from each of the sources modeled. A test problem containing several (α,n) and spontaneous fission neutron sources verifies the method.

Enhancing Monte Carlo Workflows for Nuclear Reactor Analysis with Metamodel-Driven Modeling

Monte Carlo codes are essential components of many reactor physics simulation workflows as high-fidelity continuous-energy neutron transport solvers. Among Monte Carlo radiation transport codes, MCNP is particularly notable due to its diverse simulation capabilities, large user base, and long validation history. Despite being a powerful simulation tool, MCNP provides limited capabilities to allow automated execution, model transformation, or support for user-defined logic and abstractions that limit its compatibility with modern workflows. To better integrate MCNP into a modern scientific workflow, we have developed an intuitive yet full-featured MCNP Application Program Interface (API) in Python, named MCNPy, which provides a specialized set of classes for MCNP input development. Moreover, to guarantee that our reading, writing, and modeling capabilities remain self-consistent (and to render the huge scope of the MCNP API manageable), we have adopted a strategy of model-driven software development in which a generalized model of the MCNP input format has been created. From this generalized model, or “metamodel,” problem-specific implementations such as an engine for input validation or a codebase for programmatic operations may be automatically generated. Since MCNPy primarily acts as a Python front-end to the underlying Java API that directly interfaces with the metamodel, it is intrinsically linked to the metamodel and thus remains maintainable. With MCNPy, users can programmatically read, write, and modify any syntactically valid MCNP input file regardless of its origin. These capabilities allow users to automate complicated tasks like design optimization and model translation for nuclear systems. As examples, this work demonstrates the use of MCNPy to find the critical radius of a plutonium sphere and to translate a 9000+ line MCNP input file into a corresponding OpenMC model.

FLUKAVAL – A validation framework for the FLUKA radiation transport Monte Carlo code

The FLUKA general purpose radiation transport Monte Carlo code being developed and maintained by CERN (https://fluka.cern) has adopted modern software development standards including a formal quality assurance process. This includes the FLUKAVAL testing framework that takes into account the specific needs of testing a Monte Carlo radiation transport simulation code. FLUKAVAL allows to submit, process and validate a large number of test cases for a new FLUKA version and compare the results against previous versions and reference data. It produces a quantitative and qualitative comparison and compiles a summary report over all selected test cases, which is reviewed before the new FLUKA version is released. The overall validation requires only few manual steps and makes use of large computing clusters to run simulations in parallel.

Organ dose calculator for diagnostic nuclear medicine patients based on the ICRP reference voxel phantoms and biokinetic models

The exponential growth in the use of nuclear medicine procedures represents a general radiation safety concern and stresses the need to monitor exposure levels and radiation-related long term health effects in NM patients. In the current study, following our previous work on NCINM version 1 based on the UF/NCI hybrid phantom series, we calculated a comprehensive library of S values using the ICRP reference pediatric and adult voxel phantoms and established a library of biokinetic data from multiple ICRP Publications, which were then implemented into NCINM version 2. We calculated S values in two steps: calculation of specific absorbed fraction (SAF) using a Monte Carlo radiation transport code combined with the twelve ICRP pediatric and adult voxel phantoms for a number of combinations of source and target region pairs; derivation of S values from the SAFs using the ICRP nuclear decay data. We also adjusted the biokinetic data of 105 radiopharmaceuticals from multiple ICRP publications to match the anatomical description of the ICRP voxel phantoms. Finally, we integrated the ICRP phantom-based S values and adjusted biokinetic data into NCINM version 2. The ratios of cross-fire SAFs from NCINM 2 to NCINM 1 for the adult phantoms varied widely from 0.26 to 5.94 (mean = 1.24, IQR = 0.77–1.55) whereas the ratios for the pediatric phantoms ranged from 0.64 to 1.47 (mean = 1.01, IQR = 0.98–1.03). The ratios of absorbed dose coefficients from NCINM 2 over those from ICRP publications widely varied from 0.43 (colon for 99mTc-ECD) to 2.57 (active marrow for 99mTc-MAG3). NCINM 2.0 should be useful for dosimetrists and medical physicists to more accurately estimate organ doses for various nuclear medicine procedures.

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.

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.

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.

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.