Monte Carlo Particle Transport Code Research Articles

Investigation on the impact of the spatial arrangement of individuals in a standing-type in-vivo monitoring system

An automated standing type Quick Scan Whole Body Monitor (QS-WBM) has been developed for the measurement of internal radioactive contamination due to high energy photon (HEP) emitters (Eϒ >200keV). Individuals are monitored while standing on a platform inside QS-WBM at specified reference position. Instances may occur where individuals deviate from their monitoring position, potentially leading to errors in the measurement of the body content. Positional inaccuracies are estimated in three potential scenarios: lateral displacement of the individual, movement perpendicular to the detector face, and forward bending of the torso. The Monte Carlo particle transport code FLUKA is used for estimation of efficiencies for selected radionuclides across various geometrical positions for adult male radiation workers. Approximately, 8% variation in calibration factors (CFs) is noted for lateral movements up to 15 cm from the centerline regardless of the energy, while perpendicular movements from reference position exhibit variations ranging from 14 to 60 %. This study helps in quantifying uncertainties in the estimation of body content and dose, arising from deviations in position of individuals in QS-WBM enclosure during monitoring.

Integrating ytopt and libEnsemble to autotune OpenMC

Ytopt is a Python machine-learning-based autotuning software package developed within the ECP PROTEAS-TUNE project. The ytopt software adopts an asynchronous search framework that consists of sampling a small number of input parameter configurations and progressively fitting a surrogate model over the input-output space until exhausting the user-defined maximum number of evaluations or the wall-clock time. libEnsemble is a Python toolkit for coordinating workflows of asynchronous and dynamic ensembles of calculations across massively parallel resources developed within the ECP PETSc/TAO project. libEnsemble helps users take advantage of massively parallel resources to solve design, decision, and inference problems and expands the class of problems that can benefit from increased parallelism. In this paper we present our methodology and framework to integrate ytopt and libEnsemble to take advantage of massively parallel resources to accelerate the autotuning process. Specifically, we focus on using the proposed framework to autotune the ECP ExaSMR application OpenMC, an open source Monte Carlo particle transport code. OpenMC has seven tunable parameters some of which have large ranges such as the number of particles in-flight, which is in the range of 100,000 to 8 million, with its default setting of 1 million. Setting the proper combination of these parameter values to achieve the best performance is extremely time-consuming. Therefore, we apply the proposed framework to autotune the MPI/OpenMP offload version of OpenMC based on a user-defined metric such as the figure of merit (FoM) (particles/s) or energy efficiency energy-delay product (EDP) on Crusher at Oak Ridge Leadership Computing Facility. The experimental results show that we achieve the improvement up to 29.49% in FoM and up to 30.44% in EDP.

Point containment algorithms for constructive solid geometry with unbounded primitives

We present several algorithms for evaluating point containment in constructive solid geometry (CSG) trees with unbounded primitives. Three algorithms are presented based on postfix, prefix, and infix notations of the CSG binary expression tree. We show that prefix and infix notations enable short-circuiting logic, which reduces the number of primitives that must be checked during point containment. To evaluate the performance of the algorithms, each algorithm was implemented in the OpenMC Monte Carlo particle transport code, which relies on CSG to represent solid bodies through which subatomic particles travel. Two sets of tests were carried out. First, the execution time to generate a rasterized image of a 2D slice of three CSG models of varying complexity was measured. Use of both prefix and infix notations offered significant speedup over the postfix notation that has traditionally been used in particle transport codes, with infix resulting in a 6× reduction in execution time relative to postfix for a model of a tokamak fusion device. We then measured the execution time of neutron transport simulations of the same three models using each of the algorithms. The results and performance improvements reveal the same trends as for the rasterization test, with a 5.52× overall speedup using the infix notation relative to the original postfix notation in OpenMC for the tokamak model.

Visualization of spatial dose distribution for effective radiation protection education in interventional radiology: obtaining high-accuracy spatial doses.

In recent years, eye lens exposure among radiation workers has become a serious concern in medical X-ray fluoroscopy and interventional radiology (IVR), highlighting the need for radiation protection education and training. This study presents a method that can maintain high accuracy when calculating spatial dose distributions obtained via Monte Carlo simulation and establishes another method to three-dimensionally visualize radiation using the obtained calculation results for contributing to effective radiation-protection education in X-ray fluoroscopy and IVR. The Monte Carlo particle and heavy ion transport code system (PHITS, Ver. 3.24) was used for calculating the spatial dose distribution generated by an angiography device. We determined the peak X-ray tube voltage and half value layer using Raysafe X2 to define the X-ray spectrum from the source and calculated the X-ray spectrum from the measured results using an approximation formula developed by Tucker et al. Further, we performed measurements using the "jungle-gym" method under the same conditions as the Monte Carlo calculations for verifying the accuracy of the latter. An optically stimulated luminescence dosimeter (nanoDot dosimeter) was used as the measuring instrument. In addition, we attempted to visualize radiation using ParaView (version 5.12.0-RC2) using the spatial dose distribution confirmed by the above calculations. A comparison of the measured and Monte Carlo calculated spatial dose distributions revealed that some areas showed large errors (12.3 and 24.2%) between the two values. These errors could be attributed to the scattering and absorption of X-rays caused by the jungle gym method, which led to uncertain measurements, and (2) the angular and energy dependencies of the nanoDot dosimetry. These two causes explain the errors in the actual values, and thus, the Monte Carlo calculations proposed in this study can be considered to have high-quality X-ray spectra and high accuracy. We successfully visualized the three-dimensional spatial dose distribution for direct and scattered X-rays separately using the obtained spatial dose distribution. We established a method to verify the accuracy of Monte Carlo calculations performed through the procedures considered in this study. Various three-dimensional spatial dose distributions were obtained with assured accuracy by applying the Monte Carlo calculation (e.g., changing the irradiation angle and adding a protective plate). Effective radiation-protection education can be realized by combining the present method with highly reliable software to visualize dose distributions.

Experimental, simulation, and theoretical investigations of gamma and neutron shielding characteristics for reinforced with boron carbide and titanium oxide composites

This study investigates the radiation shielding properties of composites containing polyester resin, boron carbide and titanium oxide in varying proportions. The radiation transmittance of the produced composites was measured using radioactive sources 241Am, 137Cs, 133Ba, 60Co, and 22Na across a photon energy range of 59.5 keV–1332.5 keV. To analyze radiation permeability, experimental geometry was simulated using Monte Carlo-based computer codes Monte Carlo N-Particle (MCNP), Particle and Heavy Ion Transport code System (PHITS) and FLUktuierende KAskade or Fluctuating Cascade (FLUKA), and the experimental data were compared with simulation results and theoretical data obtained from WINXCOM. The findings demonstrated a high degree of consistency between experimental, simulation, and theoretical results. Notably, the BCTiO50 sample, which included a 50% addition of TiO2, emerged as the most effective in photon radiation shielding. In addition to gamma shielding, the neutron shielding properties of the produced composites were evaluated using the FLUKA code, revealing that the BCTiO0 sample, with the highest percentage of boron, provided the best neutron shielding. This analysis included composites with decreasing boron carbide content by 10%, 20%, 30%, 40%, and 50% by weight, each replaced with titanium oxide, showcasing the potential applications of these polymer composites in radiation shielding.

Novel radiation quality metrics accounting for proton energy spectra for RBE proton models.

For proton therapy, a relative biological effectiveness (RBE) of 1.1 is widely applied clinically. However, due to abundant evidence of variable RBE in vitro, and as suggested in studies of patient outcomes, RBE might increase by the end of the proton tracks, as described by several proposed variable RBE models. Typically, the dose averaged linear energy transfer ( ) has been used as a radiation quality metric (RQM) for these models. However, the optimal choice of RQM has not been fullyexplored. This study aims to propose novel RQMs that effectively weight protons of different energies, and assess their predictive power for variable RBE in proton therapy. The overall objective is to identify an RQM that better describes the contribution of individual particles to the RBE of protonbeams. High-throughput experimental set-ups of in vitro cell survival studies for proton RBE determination are simulated utilizing the SHIELD-HIT12A Monte Carlo particle transport code. For every data point, the proton energy spectra are simulated, allowing the calculation of novel RQMs by applying different power levels to the spectra of LET or effective ( ) values. A phenomenological linear-quadratic-based RBE model is then applied to the in vitro data, using various RQMs as input variables, and the model performance is evaluated by root-mean-square-error (RMSE) for the logarithm of cell surviving fractions of each datapoint. Increasing the power level, that is, putting an even higher weight on higher LET particles when constructing the RQM is generally associated with an increased model performance, with dose averaged (i.e., dose averaged cubed LET, ) resulting in a RMSE value 0.31, compared to 0.45 for a model based on (linearly weighted) , with similar trends also observed for track averaged and -basedRQMs. The results indicate that improved proton variable RBE models can be constructed assuming a non-linear RBE(LET) relationship for individual protons. If similar trends hold also for an in vitro-environment, variable RBE effects are likely better described by or tracked averaged cubed LET ( ), or corresponding -based RQM, rather than linearly weighted or which is conventionally applied today.

Material attractiveness of irradiated fuel salts from the Seaborg Compact Molten Salt Reactor

Over the years, numerous evaluations of material attractiveness have been performed for conventional light water reactors to better understand the nature of the spent fuel material and its desirability for misuse at different points in the nuclear fuel cycle. However, availability of such assessments for newer, Generation IV reactors such as Molten Salt Reactors is rather limited. In the present study we address the gap in knowledge of material attractiveness for molten salt reactor systems and describe the nature of irradiated fuel salts which the nuclear safeguards community might be faced with in the near future as more and more such reactors enter commission and operation. Within the scope of the paper, we use a large database of simulated irradiated fuel salt isotopics (and other derived quantities such as gamma activity, decay heat, and neutron emission rates) developed specifically for a molten salt reactor concept in order to shed some light on possible weapons usability of uranium and plutonium present in the irradiated fuel salts. This has been achieved by proposing a new attractiveness metric that is better suited for quantifying attractiveness of irradiated salts from a model molten salt concept. The said metric has been computed using a database that has been created by simulating the irradiation of molten fuel salt in a concept core over a wide range of operational parameters (burnup, initial enrichment, and cooling time) using the Monte-Carlo particle transport code, Serpent. With the help of this attractiveness metric, the findings from this study have shown that in relative terms, molten salt spent fuel is more attractive than spent fuel produced by a conventional light water reactor. The findings also underscore the need for strengthened safeguards measures for such spent fuel. These results are expected to be useful in the future for regulatory authorities as well as for nuclear safeguards inspectors for designing a functional safeguards verification routine for irradiated fuel of such unique nature.

ParaStell: parametric modeling and neutronics support for stellarator fusion power plants

The three-dimensional variation inherent to stellarator geometries and fusion sources motivates three-dimensional modeling to obtain accurate results from computational modeling in support of design and analysis of first wall, blanket, and shield (FWBS) systems. Manually constructing stellarator fusion power plant geometries in computer-aided design (CAD) and defining the corresponding fusion source can be cumbersome and challenging. The open-source parametric modeling toolset ParaStell has been developed to automate construction of such geometries in low-fidelity. Low-fidelity modeling is useful during the conceptual phase of engineering design as a means of rapidly exploring the design space of a given device. The modeling capability of ParaStell includes in-vessel components and magnets, for any given stellarator configuration, using a parametric definition and plasma equilibrium data. Furthermore, the toolset automates the generation of detailed, tetrahedral neutron source definitions and DAGMC geometries for use in neutronics modeling. ParaStell assists rapid design iteration, parametric study, and design optimization of stellarator fusion cores. As a demonstration of the design iteration capability, the effect of the three-dimensional parameter space on tritium breeding and magnet shielding is investigated, using the WISTELL-D configuration as a design basis. Blanket and shield thicknesses are varied in three dimensions, using the space available between the plasma edge and magnet coils as a constraint. The corresponding effects on tritium breeding ratio and magnet heating are tallied using the open-source Monte Carlo particle transport code OpenMC. The inclusion of additional and higher-fidelity modeling capabilities is planned for ParaStell’s future, as well as its implementation in machine-driven optimization.

Modeling gamma detectors in OpenMC: Validation of a newly implemented pulse-height tally

Gamma spectroscopy measurements can be simulated using a pulse-height tally functionality of Monte Carlo particle transport codes. Such a functionality must account for the complete simulation history of a particle’s energy deposited in a particular volume. OpenMC, an open-source application for neutron and photon particle transport which is widely used in the nuclear engineering community, previously lacked adequate simulation capabilities for gamma spectroscopy. In this paper, we outline the implementation of the photon pulse-height tally for OpenMC. Additionally, we validate the new function by comparing results to analytical calculations, other software, and experimental data.

Development and validation of fully open-source R2S shutdown dose rate capabilities in OpenMC *

We present the first fully open-source capabilities for shutdown dose rate (SDR) calculations of fusion energy facilities based on the Rigorous 2-Step (R2S) methodology. These capabilities have been implemented in the OpenMC Monte Carlo particle transport code, building on its existing capabilities while also leveraging new features that have been added to the code to support SDR calculations, such as decay photon source generation. Each of the individual physics components in the R2S workflow—neutron transport, activation, decay photon source generation, and photon transport—have been verified through code-to-code comparisons with MCNP6.2 and FISPACT-II 4.0. These comparisons generally demonstrate excellent agreement between codes for each of the physics components. The full cell-based R2S workflow was validated by performing a simulation of the first experimental campaign from the Frascati Neutron Generator (FNG) ITER dose rate benchmark problem from the Shielding INtegral Benchmark Archive and Database (SINBAD). For short cooling times, the dose calculated by OpenMC agrees with the experimental measurements within the stated experimental uncertainties. For longer cooling times, an overprediction of the shutdown dose was observed relative to experiment, which is consistent with previous studies in the literature. Altogether, these features constitute a combination of capabilities in a single, open-source codebase to provide the fusion community with a readily-accessible option for SDR calculations and a platform for rapidly analyzing the performance of fusion technology.

Combining Wave and Particle Effects in the Simulation of X-ray Phase Contrast—A Review

X-ray phase-contrast imaging (XPCI) is a family of imaging techniques that makes contrast visible due to phase shifts in the sample. Phase-sensitive techniques can potentially be several orders of magnitude more sensitive than attenuation-based techniques, finding applications in a wide range of fields, from biomedicine to materials science. The accurate simulation of XPCI allows for the planning of imaging experiments, potentially reducing the need for costly synchrotron beam access to find suitable imaging parameters. It can also provide training data for recently proposed machine learning-based phase retrieval algorithms. The simulation of XPCI has classically been carried out using wave optics or ray optics approaches. However, these approaches have not been capable of simulating all the artifacts present in experimental images. The increased interest in dark-field imaging has also prompted the inclusion of scattering in XPCI simulation codes. Scattering is classically simulated using Monte Carlo particle transport codes. The combination of the two perspectives has proven not to be straightforward, and several methods have been proposed. We review the available literature on the simulation of XPCI with attention given to particular methods, including the scattering component, and discuss the possible future directions for the simulation of both wave and particle effects in XPCI.

Language and design evolution of the OpenMC Monte Carlo particle transport code

The OpenMC Monte Carlo particle transport code has been continuously developed for 13 years by a large community of contributors. In that time span, the codebase has undergone significant changes that have redefined what OpenMC is and made it an enduring presence in the nuclear science and engineering community. In this paper, we discuss the evolution of programming language use in OpenMC, trends in the overall design of the programming interfaces, and implications for the future of the code.

DPA calculation in a stainless steel baffle of Chooz-A PWR using auto-dosimetry measurements

This paper presents a method dedicated to the calculation of DPA (displacements per atom) in a stainless steel element irradiated inside Chooz-A reactor core. This method is based on experimental results obtained in samples taken from the analyzed stainless steel element, and calculation results. Experimental results consist in the mass of 59Co and the activity of 60Co, both measured in 2018, i.e. 27 years after the final shutdown of Chooz-A reactor. Calculations are carried out with the TRIPOLI-4® 3D Monte Carlo particle transport code, and the DARWIN/PEPIN2 depletion code, with a simplified modelling of Chooz-A reactor. As explained in this paper, these simplifications are corrected by the use of measurement results. Due to a lack of information about the irradiated steel element (initial 59Co quantity in particular), an iterative process was developed and used. It allows to evaluate missing information and to calculate corrected DPA results in samples. Besides, stress tests were carried out to check the robustness of the process and the uniqueness of DPA result for each sample. This work was carried out by CEA/SERMA with the financial support of EDF.

Compendium on Monte Carlo simulation of photoneutrons in the Giant Dipole Resonance energy range: The first five elements

Neutrons generated by photonuclear reactions, “photoneutrons”, are encountered in various applications involving high-energy gamma sources, electron accelerators or nuclear reactors. Monte Carlo particle-transport codes are generally used to simulate the emission of photoneutrons, characterize their field or assess their impact on nuclear systems. The aim of this work is to create a compendium on the simulation of photoneutrons using several Monte Carlo codes, i.e., MCNP6, PHITS and TRIPOLI-4, each code being run successively with ENDF/B-VIII.0 and JENDL-5 nuclear data libraries. We study the photoneutron fields produced by 50 elements with their natural isotopic composition from the reaction energy threshold up to 30 MeV, i.e., in the regime of the Giant Dipole Resonance (GDR). The photoneutron fields are characterized according to three observables, i.e., photoneutron current, energy spectrum and angular distribution. This paper presents the results obtained for the first five elements in order of increasing atomic number, i.e., deuterium, beryllium, carbon, nitrogen and oxygen. The compendium could serve as a handbook for users to master the current strengths and limitations of the codes, for code developers to make progress in the sampling of neutron-emitting photonuclear reactions, and more broadly for all researchers working on photoneutrons, whether they are evaluators of nuclear data libraries or experimental nuclear physicists.

Analysis of fluctuations in ex-core neutron detector signal in Krško NPP during an earthquake

During an earthquake on December 29th 2020, the Krško NPP automatically shutdown due to the trigger of the negative neutron flux rate signal on the power range nuclear instrumentation. From the time course of the detector signal, it can be concluded that the fluctuation in the detector signal may have been caused by the mechanical movement of the ex-core neutron detectors or the pressure vessel components rather than the actual change in reactor power. The objective of the analysis was to evaluate the sensitivity of the neutron flux at the ex-core detector position, if the detector is moved in the radial or axial direction. In addition, the effect of the core barrel movement and core inside the baffle movement in the radial direction were analysed. The analysis is complemented by the calculation of the thermal and total neutron flux gradient in radial, axial and azimuthal directions. The Monte Carlo particle transport code MCNP was used to study the changes in the response of the ex-core detector for the above-mentioned scenarios. Power and intermediate-range detectors were analysed separately, because they are designed differently, positioned at different locations, and have different response characteristics. It was found that the movement of the power range ex-core detector has a negligible effect on the value of the thermal neutron flux in the active part of the detector. However, the radial movement of the intermediate-range detector by 5 cm results in 7%–8% change in the thermal neutron flux in the active part of the intermediate-range detector. The analysis continued with an evaluation of the effects of moving the entire core barrel on the ex-core detector response. It was estimated that the 2 mm core barrel radial oscillation results in ∼4% deviation in the power and intermediate-range detector signal. The movement of the reactor core inside baffle can contribute ∼6% deviation in the ex-core neutron detector signal. The analysis showed that the mechanical movement of ex-core neutron detectors cannot explain the fluctuations in the ex-core detector signal. However, combined core barrel and reactor core inside baffle oscillations could be a probable reason for the observed fluctuations in the ex-core detector signal during an earthquake.

Radiation attenuation and elemental composition of locally available ceramic tiles as potential radiation shielding materials for diagnostic X-ray rooms

Ceramic materials are being explored as alternatives to toxic lead sheets for radiation shielding due to their favorable properties like durability, thermal stability, and aesthetic appeal. However, crafting effective ceramics for radiation shielding entails complex processes, raising production costs. To investigate local viability, this study evaluated Malaysian ceramic tiles for shielding in diagnostic X-ray rooms. Different ceramics in terms of density and thickness were selected from local manufacturers. Energy Dispersive X-ray Fluorescence (EDXRF) and X-ray Fluorescence (XRF) characterized ceramic compositions, while Monte Carlo Particle and Heavy Ion Transport code System (MC PHITS) simulations determined Linear Attenuation Coefficient (LAC), Half-value Layer (HVL), Mass Attenuation Coefficient (MAC), and Mean Free Path (MFP) within the 40–150 kV energy range. Comparative analysis between MC PHITS simulations and real setups was conducted. The C3–S9 ceramic sample, known for homogeneous full-color structure, showcased superior shielding attributes, attributed to its high density and iron content. Notably, energy levels considerably impacted radiation penetration. Overall, C3–S9 demonstrated strong shielding performance, underlining Malaysia's potential ceramic tile resources for X-ray room radiation shielding.

Spatial-temporal tracking of molecular species in Geant4-DNA track structure simulations

Monte Carlo particle transport codes have been used to simulate the radiolysis of water and subsequent evolution of molecular species during the chemical stage. In Geant4, accounting of molecules is performed using a molecule counter which provides only a global molecule count. We developed a modular counter implementation that was used to perform spatial-temporal tracking of molecule counts. This custom counter was validated in homogeneous and simplistic split geometries for 1 MeV electrons and 50 MeV protons.Using a spherical cell with one nucleus and one hundred mitochondria as an example, the radiolytic yields in the cytoplasm, nucleus, and mitochondria were compared between targeted (electrons originating from within mitochondria) and unspecific (electrons originating from within cytoplasm) for (very) low energy electrons with kinetic energies of 100 eV to 1 keV.At short times radiolytic yields in source volumes were consistent with reference yields obtained in a large homogeneous volume. Past 1 ns diffusion out of the source volumes led to yield reductions. For volumes abutting the source volumes radiolytic yields at short times strongly depended on track end effects, most pronounced for 100 eV electrons. Similarly, diffusion out of the source volumes greatly increased yields in abutting volumes past 1 ns with magnitude decreasing and time of increase increasing with increasing electron energy. For higher energy electron beams (100 keV and 200 keV) radiolytic yields in all volumes were consistent with reference yields for large statistics.Thus, in cases where energy deposit sites are uniformly distributed in the volume of interest and energies of the charged particles are high to avoid track end effects reference yields may be used in place of resource intensive chemical stage simulations. Spatially heterogeneous energy deposits, as observed in targeted ɑ or β radiation therapy or with targeted nanoparticles, using a spatially sensitive molecule counter is suggested when estimating radiolytic yields outside of the source volumes or at distances that are within the diffusion range of the molecules of interest.

Radiation protection studies for the INRNE-BAS cyclotron facility using Monte Carlo FLUKA code

The Bulgarian National Cyclotron Centre at the Institute for Nuclear Research and Nuclear Energy, Bulgarian Academy of Sciences (INRNE-BAS), is going to be a new facility dedicated to production of medical radioisotopes and research in the corresponding fields. This work details the shielding requirements of the facility and the preliminary radiological studies that we have carried out. The activity concentration of the long-lived radioisotopes is also evaluated which provides a useful guideline for the decommissioning plan of the facility. For these purposes we have performed simulations with the Monte Carlo particle transport and interaction code FLUKA.

A better hash method for high-fidelity Monte Carlo simulations on nuclear reactors

With the increasing demand for high-fidelity nuclear reactor simulations, the acceleration of Monte Carlo particle transport codes is becoming a core problem. One of the bottlenecks is locating millions or even billions of cells and fetching their associated parameters in the repeated geometry structure. Typically, Monte Carlo codes utilize a hash function to accelerate the cell locating and parameter indexing process. Specifically, they use the “cell vector → hash → parameter” method to accelerate the direct “cell vector → parameter” method. In this work, we propose a better hash method based on the Cyclic Redundancy Check (CRC) mechanism, which has been mathematically proven to be efficient and produce fewer hash collisions. Experimentally, this new hash method has been compared with some other hash functions and showed its superiority in terms of the calculation speed and collision probabilities. This hash method has been integrated into the Reactor Monte Carlo code RMC and worked well in practical applications.

Neutronic response of the neutral beam after EAST NBI upgraded

In order to improve the heating effect of neutral beam injector, the neutral beam injector (NBI) of the Experimental Advanced Superconducting Tokamak (EAST) is planned to be upgraded. After the EAST NBI is upgraded, the neutron emission intensity will increase. Neutron leaking from the neutral beam port penetrate the beamline, resulting in the material radiation damage. The neutronic response has been investigated by using the Monte Carlo particle transport code MCNP with the two dimensional RZ model of the EAST NBI. The results demonstrate that a 40% increase of the neutron emission rate has been attained with the upgraded EAST NBI. However, the simulations for the activation of different materials, displacement damage and nuclear heating of the neutral beam still satisfy the requirement of material safety performance. The calculations of nuclear response show that the materials can meet the requirements of the NBI steady state operation after the EAST NBI updated.