Monte Carlo Particle Transport Code Research Articles

Simulating Charge Deposition by Cosmic Rays Inside Astronomical Imaging Detectors

In the context of space astronomy missions, accurate and fast modeling of spurious images generated by solar and galactic cosmic rays (GCRs) in imaging detectors is critical in order to assess and mitigate their detrimental effects. Currently, it is only possible to accurately reproduce the CR images including energy loss straggling, with the use of complex Monte Carlo particle transport codes. However, these codes are difficult to use and cannot be easily included in instrument simulation pipelines. Presented here is CosmiX, a novel, open-source, and Python-based CR model, overcoming the above-mentioned limitations, yet still capable of accurately reproducing the CR images, paving the way for semianalytical Monte Carlo modeling of the CR tracks. These are demonstrated by using CosmiX to reproduce irradiation test data of a PLATO charge-coupled device (CCD) and in-orbit CR data of Gaia CCD detectors.

Simulation of the response of a segmented high-purity germanium detector for gamma emission tomography of nuclear fuel

Irradiation testing of nuclear fuel is routinely performed in nuclear test reactors. For qualification and licensing of accident-tolerant fuels or generation IV reactor fuels, an extensive increase in irradiation testing is foreseen in order to fill the gaps of existing validation data, both in normal operational conditions and in order to identify operational limits. Gamma emission tomography (GET) has been demonstrated as a viable technique for studies of the behavior of irradiated nuclear fuel, e.g., measurement of fission gas release and inspection of fuel behavior under loss-of-coolant accident conditions. In this work, the aim is to improve the technique of GET for irradiated nuclear fuel, by developing a detector concept that allows for a higher spatial resolution and/or faster interrogation. We present the working principles of a novel concept for gamma emission tomography using a segmented high-purity germanium (HPGe) detector. The performance of this concept was investigated using the Monte Carlo particle transport code MCNP. In particular, the data analysis of the proposed detector was evaluated, and the performance, in terms of full energy efficiency and misidentification rate (i.e., localization failure), was assessed. We concluded that the segmented HPGe detector has an advantageous performance as compared to the traditional single-channel detector systems. Due to the scattering nature of gamma rays, a trade-off is presented between efficiency and cross-talk; however, the performance is nevertheless a substantial improvement over the currently used single-channel HPGe detector systems.

Photon transport physics in Serpent 2 Monte Carlo code

This paper presents the photon physics models used in Serpent 2 Monte Carlo particle transport code. The four main photon interactions – photoelectric effect, Rayleigh scattering, Compton scattering and electron–positron pair production – are included in Serpent, accompanied by three secondary photon production mechanisms (electron–positron annihilation, atomic relaxation and bremsstrahlung emitted by electrons and positrons). Models, algorithms and data used for these interaction mechanisms are described, including new efficient sampling algorithms for photoelectron direction and electron–positron pair production.

MCS – A Monte Carlo particle transport code for large-scale power reactor analysis

A new Monte Carlo (MC) neutron/photon transport code, called MCS, has been developed at Ulsan National Institute of Science and Technology (UNIST) with the aim of performing the high-fidelity multi-physics simulation of large-scale power reactors, especially pressurized water reactors (PWR). The high-fidelity multi-physics analysis of large-scale PWR is a challenging problem due to two aspects, the first being the difficulty of implementing various state of the art techniques into a single code system, and the other making it feasible to run such simulations on practical computing machines within reasonable amount of memory usage and computing time. In this paper, features implemented into MCS for large-scale PWR simulations are described including but not limited to depletion, thermal/hydraulics coupling, fuel performance coupling, equilibrium xenon, on-the-fly neutron cross-section Doppler broadening, and critical boron search. The efficient memory usage for burnup simulation and the high performance of MCS through various algorithms and optimizations (parallel fission bank, hash indexing) are illustrated on Monte Carlo performance benchmarks. Finally, the large-scale PWR analysis capability is fully demonstrated with BEAVRS Cycles 1 & 2 calculations.

Monte Carlo simulation technology based on characteristic γ-ray spectrum analysis

<sec>Monte Carlo method is an ideal way to simulate criticality, shielding and nuclear detection. JMCT is a multipurpose 3D Mont Carlo (MC) neutron-photon-electron and coupled neutron /photon /electron transport code which is developed by IAPCM. The program is developed based on the combinatorial geometry parallel infrastructure JCOGIN and has the most functions of general Monte Carlo particle transport code, including the various variance reduction techniques. In addition, some new algorithms, such as Doppler broadening on-the-fly (OTF), uniform tally density (UTD), consistent adjoint driven importance sampling (CADIS), fast criticality search of boron concentration (FCSBC), the domain decomposition (DD), the two-level parallel computation of MPI and OpenMP, etc. have been developed, where the number of geometry zones, materials, tallies, depletion zones, memories and period of random number are big enough to simulate various extremely complicated problems. Also the JMCT is hybrid the discrete ordinate SN program JSNT to generate source biasing factors and weight window parameters for deep-penetration shielding problems. The input is based on the CAD modeling, and the result is a visualized output. The JMCT can provide technology support for radiation shielding design, reactor physics and criticality safe analysis. Especially, the JMCT is coupled depletion and thermal-hydraulic code for simulating the reactor feedback effect, including depletion, thermal feedback. In recent years, new function of γ-ray spectrum analysis has been developed. </sec><sec>In this paper, the working principles of timing measure are introduced. The advanced calibration count is developed for distinguishing between inelastic γ-ray and capture γ-ray based on time bin tally. On the other hand, when neutron collides with nuclide, the secondary photon is labeled into the primary line photon and primary continuous photon, where energy of primary line photon does not change with the incident neutron energy, such as carbon spectral-line at 4.43 MeV and oxygen spectral-line at 6.13 MeV. The element components of detected object can be determined by the primary line photon. On the other hand, expect value estimator (EVE) is used to produce the secondary photons. The advantage of EVE does not leak any event even with a small probability which is important for detecting the hide exploder. However the shortage of the EVE results in producing a great number of photons with small weight. If all of these small weight photons are simulated one by one, a great amount of computation time and memory will be consumed. For avoiding this case, a new algorithm is design by coupling EVE and DE (direct estimator). The all of secondary photons from EVE only make the direct tally take a little computing time, then end the photon history and return to the DE production photon model (one photon production at most). Final, the total tally is a summation of EVE direct tally and DE scattering tally. The use of new algorithm to realize the analysis of γ-ray spectrum will increase only a little computing time. The numerical tests are done by using own Monte Carlo code JMCT. The correctness and validity of the algorithm are shown preliminarily.</sec>

Radiation hardness studies and detector characterisation at the JSI TRIGA reactor

The JSI TRIGA reactor features several in-core and ex-core irradiation facilities, each having different properties, such as neutron/gamma flux intensity, spectra and irradiation volume. A series of experiments and calculations was performed in order to characterise radiation fields in irradiation channel thus allowing users to perform irradiations in a well characterised environment. Since 2001 the reactor has been heavily used for radiation hardness studies for components used at accelerators such as the Large Hadron Collider (LHC) at CERN. Since 2010 it has been extensively used for testing of new detectors and innovative data acquisition systems and methods developed and used by the CEA. Recently, several campaigns were initiated to characterise the gamma field in the reactor and use the experimental data for improvement of the treatment of delayed gammas in Monte Carlo particle transport codes. In the future it is planned to extend the testing options by employing pulse mode operation, installation of a high energy gamma ray irradiation facility and allow irradiation of larger samples at elevated temperature.

Design of a compact shielding envelope and elements of radiological protection at the TRIUMF-ARIEL facility

The Advanced Rare Isotope Laboratory (ARIEL) is currently under construction at TRIUMF. ARIEL's mission is to supply existing experiments with more diverse and more intense radioactive ion beams (RIB) produced via the Isotope Separation On-Line (ISOL) method. The drivers inducing the nuclear reactions at ARIEL include a 500 MeV proton beam from the main cyclotron as well as a 35 MeV electron beam with the intent of simultaneous operation using two separate additional target stations. The generated RIB will be made available to experiments in condensed matter and subatomic physics as well as practical applications such as medical isotopes for life sciences research.This work details the various simulations performed using the Monte Carlo (MC) particle transport and interaction code FLUKA. The code was systematically employed to predict prompt and residual radiation levels to characterize the radiation fields generated via the irradiation of various types of targets. An iterative combined development process between CAD-based engineering of the shielding envelope, nuclear physics optimization of the target system and particle tracing simulations was established.Preliminary simulation studies resulted in the selection of shielding materials and respective thicknesses with the goal of maximizing attenuation while mitigating residual activation. The results provide clear operational constraints and aid in determination of expected dose rates outside of shielding in high occupancy areas. A series of optimization steps followed the preliminary work, testing various parameters including beam power, target material and shielding configurations. These results were evaluated considering technical feasibility, safety, and economic impacts on ARIEL construction and eventual operation.Although the project, in its infancy, is expected to proceed through stages of increasing complexity, the shielding and radiation transport modeling completed to date helps to define and validate nominal operation and failure scenario procedures for all future project stages.

A Serpent2-SUBCHANFLOW-TRANSURANUS coupling for pin-by-pin depletion calculations in Light Water Reactors

This work presents the development of a coupling scheme for Serpent2, a continuous-energy Monte Carlo particle transport code, SUBCHANFLOW, a subchannel thermalhydraulics code, and TRANSURANUS, a fuel-performance code, suitable for large-scale high-fidelity depletion calculations for Light Water Reactors. The calculation method is based on the standard neutronic-thermalhydraulic approach, replacing the simple fuel-rod solver in SUBCHANFLOW with the more complex thermomechanic model of TRANSURANUS. The depletion method is fully coupled and semi-implicit, and the implementation relies on an object-oriented design with mesh-based feedback exchange. The results of the three-code system for a 360-day depletion calculation of a VVER-1000 fuel assembly with a pin-by-pin modelling approach are presented and analyzed. The performance of this tool, as well as the bottlenecks for its application to full-core problems, are discussed.

Benchmarking FLUKA, PHITS, MCNPX, and MARS15 codes with product yields of 209Bi(p, x) reactions

The Monte Carlo particle transport codes, FLUKA, PHITS, MCNPX, and MARS15 were benchmarked by comparison to the measured product yields of proton-induced reactions on 209Bi from 43 to 100 MeV energy range. Two rounds of experiment were carried out to measure product yields of 209Bi(p, xn)207,206,205,204,203Po and 209Bi(p, pxn)207,206,205,204Bi reactions by using 100 MeV and 69 MeV proton beams. The targets were arranged in the stacks consisting of 209Bi activation foils, Al and Au monitor foils, and Pb plates as energy degraders. The predictive capability of the codes has been proven to be satisfactory from 55 to 100 MeV energy range. However, all codes are likely to underestimate the product yields of both investigated reactions over whole energy range and also the discrepancy between codes and measurements become larger at energy lower than 55 MeV.

Reactivity analysis of a direct nuclear heated gas turbine with nitrogen coolants

This paper investigates the effects on core reactivity of using a nitrogen coolant within a prismatic High Temperature Gas Reactors (HTGR). The intent is that the nitrogen coolant will be directly compatible with existing technologically mature gas turbine rotating machinery and therefore simplify the design path to a direct nuclear heated closed cycle gas turbine. The 14N isotope of nitrogen makes up the majority (99.6%) of natural nitrogen and has a reasonably significant absorption cross-section for thermal neutrons. An understanding is therefore required of any implications for reactivity insertions during coolant pressure and temperature changes. The analysis was undertaken using the Serpent continuous-energy Monte Carlo particle transport code modelling an infinite lattice core of fuel blocks containing tristructural-isotropic (TRISO) coated fuel particles. The results of the analysis show that for low pressure ratio turbo machinery (10:1) with atmospheric compressor inlet pressure the reactivity swings are small (<60pcm). Equally the effects of temperature transients were also shown to provide only a small positive reactivity coefficient (∼0.09pcm/K) which inversely diminishes with absolute temperature. The conclusion of this paper is that the effect of nitrogen on core reactivity does not prevent its utilisation as a potential reactor coolant for small prismatic HTGRs, potentially enabling direct coupling of existing air derived turbomachinery.

Radioactive waste inventory estimation of a research reactor for decommissioning scenario development

ABSTRACT Since radioactive waste inventory plays an important role in decommissioning scenario development, we developed an estimation method of the amount of radioactive wastes for a research reactor as a case study of the 100 KWth KARTINI reactor. In order to address geometrical complexity of the research reactor with experimental equipment such as beam ports, we calculated the distribution of neutron flux by using Monte Carlo particle transport code, and the radionuclide inventories generated by neutron capture in the reactor components and structures. The amount of radioactive wastes arising from the decommissioning was estimated by assigning radioactive parts according to the clearance criterion of each radionuclide. In addition, the size of a segment of the dismantled biological shielding was considered to estimate the amount of radioactive wastes addressing the actual dismantling activity. It is also found that the amount of radioactive wastes increases with the size of the segment because it includes partially clearance materials.

Computation Speeds and Memory Requirements of Mesh-Type ICRP Reference Computational Phantoms in Geant4, MCNP6, and PHITS.

Recently, Task Group 103 of the International Commission on Radiological Protection completed the development of new adult male and female mesh-type reference computational phantoms, which are planned for use in future International Commission on Radiological Protection dose coefficient calculations. In the present study, the performance of major Monte Carlo particle transport codes, i.e., Geant4, MCNP6, and PHITS, were investigated for the mesh-type reference computational phantoms by performing transport simulations of photons, electrons, neutrons, and helium ions for some external and internal exposures, and simultaneously measuring the memory usage, initialization time, and computation speed of the adult male mesh-type reference computational phantom in the codes. The measured results were then compared with the values measured with the current adult male voxel-type reference computational phantom in International Commission on Radiological Protection Publication 110 as well as five voxel phantoms produced from the adult male mesh-type reference computational phantom with different voxel resolutions, i.e., 0.1 × 0.1 × 0.1 mm, 0.6 × 0.6 × 0.6 mm, 1 × 1 × 1 mm, 2 × 2 × 2 mm, and 4 × 4 × 4 mm. From the results, it was found that in all of the codes, the memory usage of the mesh-type reference computational phantom is greater than that of the voxel-type reference computational phantom and the lowest resolution voxelized phantom, but it is sufficiently lower than the maximum memory, 64 GB, that can be installed in a personal computer. The required initialization time of the mesh-type reference computational phantom and of the voxel-type reference computational phantom and voxelized phantoms in resolutions lower than 0.6 × 0.6 × 0.6 mm was less than a few minutes in all of the codes. As for the computation speed among the codes, MCNP6 showed the worst performance for the mesh-type reference computational phantom, which was slower than that for the voxel-type reference computational phantom by up to ~50 times and slower than that for all of the voxelized phantoms by up to ~40 times. By contrast, PHITS showed the best performance for the mesh-type reference computational phantom, which was faster than that for the voxel-type reference computational phantom by up to ~3 times and faster than that for all of the voxelized phantoms by up to ~20 times. This high performance of PHITS is indeed encouraging considering that it is used nowadays by the International Commission on Radiological Protection for most dose coefficient calculations.

Topology optimization in the framework of the linear Boltzmann equation – a method for designing optimal nuclear equipment and particle optics

In this study, we describe a procedure of topology optimization in the framework of the linear Boltzmann equation, implemented using a reference Monte-Carlo particle transport code. This procedure can design complex structures that optimize the transport of particles, according to preset objectives and constraints. Although simple and perfectible, this procedure was successfully tested on a series of difficult problems, with results outperforming human capabilities. Improved, it could be used to assist or automate the design of particle optics or nuclear devices.

Accelerating Monte Carlo Shielding Calculations in TRIPOLI-4 with a Deterministic Adjoint Flux

In radiation protection studies, the goal is to estimate the response of a detector exposed to a strongly attenuated radiation field. Monte Carlo (MC) particle transport codes give the possibility to efficiently solve for such responses using several variance-reduction (VR) methods that help allocating more CPU time to the simulation of highly contributing histories. The TRIPOLI-4® MC particle transport code offers two main methods, the exponential transform and adaptive multilevel splitting (AMS), which rely on the definition of a suitable importance map. In this paper, we present an implementation of a generalized Consistent Adjoint Driven Importance Sampling (CADIS) methodology for TRIPOLI-4. The implementation relies on coupling with the IDT code, a deterministic solver for the Boltzmann adjoint transport equation, for the generation of importance maps. We study the performance of both VR methods present in TRIPOLI-4 in this setting. In particular, to our knowledge, this is the first time that a CADIS-like methodology has been applied to AMS.

Development and validation of an open source Monte Carlo dosimetry model for wide-beam CT scanners using Fluka.

PurposeDevelopment and validation of an open source Fluka‐based Monte Carlo source model for diagnostic patient dose calculations.MethodsA framework to simulate a computed tomography (CT) scanner using Fluka Monte Carlo particle transport code was developed. The General Electric (GE) Revolution scanner with the large body filter and 120 kV tube potential was characterized using measurements. The model was validated on benchmark CT test problems and on dose measurements in computed tomography dose index (CTDI) and anthropomorphic phantoms. Axial and helical operation modes with provision for tube current modulation (TCM) were implemented. The particle simulations in Fluka were accelerated by executing them on a high‐performance computing cluster.ResultsThe simulation results agreed to better than an average of 4% of the reference simulation results from the AAPM Report 195 test scenarios, namely: better than 2% for both test problems in case 4 using the PMMA phantom, and better than 5% of the reference result for 14 of 17 organs in case 5, and within 10% for the three remaining organs. The Fluka simulation results agreed to better than 2% of the air kerma measured in‐air at isocenter of the GE Revolution scanner. The simulated air kerma in the center of the CTDI phantom overestimated the measurement by 7.5% and a correction factor was introduced to account for this. The simulated mean absorbed doses for a chest scan of the pediatric anthropomorphic phantom was completed in ~9 min and agreed to within the 95% CI for bone, soft tissue, and lung measurements made using MOSFET detectors for fixed current axial and helical scans as well as helical scan with TCM.ConclusionA Fluka‐based Monte Carlo simulation model of axial and helical acquisition techniques using a wide‐beam collimation CT scanner demonstrated good agreement between measured and simulated results for both fixed current and TCM in complex and simple geometries. Code and dataset will be made available at https://github.com/chezhia/FLUKA_CT.

APPLICATION OF SERPENT CODE CALCULATIONS: TRITIUM PRODUCTION USING FAST NEUTRONS

This paper focuses on tritium production by fast neutrons (especially neutrons with energy 14.029 MeV). Research on tritium production is important and useful in designing future fusion reactors since tritium serves as the principal fuel in DT fusion. Many neutron reactions of different types of target elements have been explored and some observations have been made. The reaction rate is the basis for an evaluated summary in this paper. Numerous calculations were made using the relatively new Serpent code, the Monte Carlo code initially developed for reactor physics calculations, but continuously updated to become general purpose Monte Carlo particle transport code.

Experimental validation of the FLUKA Monte Carlo code for dose and -emitter predictions of radioactive ion beams

In the context of hadrontherapy, whilst ions are capable of effectively destroying radio resistant, deep seated tumors, their treatment localization must be well assessed to ensure the sparing of surrounding healthy tissue and treatment effectiveness. Thus, range verification techniques, such as online positron-emission-tomography (PET) imaging, hold great potential in clinical practice, providing information on the in vivo beam range and consequent tumor targeting. Furthermore, emitting radioactive ions can be an asset in online PET imaging, depending on their half-life, compared to their stable counterparts. It is expected that using these radioactive ions the signal obtained by a PET apparatus during beam delivery will be greatly increased, and exhibit a better correlation to the Bragg Peak. To this end, FLUKA Monte Carlo particle transport and interaction code was used to evaluate, in terms of annihilation events at rest and dose, the figure of merit in using emitter, radioactive ion beams (RI). For this purpose, the simulation results were compared with experimental data obtained with an openPET prototype in various online PET acquisitions at the Heavy Ion Medical Accelerator in Chiba (HIMAC), in collaboration with colleagues from the National Institute of Radiological Sciences’ (NIRS) Imaging Physics Team. The dosimetry performance evaluation with FLUKA benefits from its recent developments in fragmentation production models. The present work estimated that irradiations with RI, produced via projectile fragmentation and their signal acquisition with state-of-the-art PET scanner, lead to nearly a factor of two more accurate definition of the signals’ peak position. In addition to its more advantageous distribution shape, it was observed at least an order magnitude higher signal acquired from 11C and 15O irradiations, with respect to their stable counterparts.

Demonstration of Multigroup Multiband Cross Section Generation in Monte Carlo Simulations

This is the investigation into the generation of high-fidelity multigroup multiband cross sections from Monte Carlo neutron transport simulations. Previous methods for generating multigroup multiband (MGMB) cross sections, and multigroup cross sections, assume an approximate shape for the scalar flux. This approximate flux shape is the product of an energy-dependent spectrum and a cross-section-dependent self-shielding factor for the material of interest. Since, these methods assume a scalar flux a priori, the cross sections generated may not be sufficiently consistent with the actual scalar flux of any particular problem. Instead, in our approach we tally both the reaction rate and the flux for each energy group and cross section magnitude band using the Mercury Monte Carlo particle transport code and compute the group-band cross sections from these tallies. This approach eliminates the need for assuming a flux by using a Monte Carlo simulation to calculate these cross sections. The increased physical fidelity of MGMB also requires more data storage, since more than one cross section value needs to be stored per energy group. However, this storage tradeoff could be mitigated by using less energy groups.

Advances in the Treatment of the Electromagnetic Cascade in the TRIPOLI-4 Monte Carlo Code

TRIPOLI-4 is a Monte Carlo particle transport code developed at CEA Saclay (France) that is employed in the domains of nuclear reactor physics, criticality safety, shielding/radiation protection, and nuclear instrumentation. The goal of this paper is to report on current developments, validation, and verification made in TRIPOLI-4 in the treatment of electron/positron/photon transport. The new capabilities and improvements concern refinements to the electron transport algorithm, the introduction of a charge-deposition score, the new thick-target bremsstrahlung option, the upgrade of the bremsstrahlung model, and the improvement of electron angular straggling at low energy. The importance of each of the developments mentioned above is illustrated by comparisons with calculations performed with other codes and with experimental data.

An overview of recent developments in FLUKA PET tools

The new developments of the FLUKA Positron-Emission-Tomography (PET) tools are detailed. FLUKA is a fully integrated Monte Carlo (MC) particle transport code, used for an extended range of applications, including Medical Physics. Recently, it provided the medical community with dedicated simulation tools for clinical applications, including the PET simulation package. PET is a well-established imaging technique in nuclear medicine, and a promising method for clinical in vivo treatment verification in hadrontherapy. The application of clinically established PET scanners to new irradiation environments such as hadrontherapy requires further experimental and theoretical research to which MC simulations could be applied. The FLUKA PET tools, besides featuring PET scanner models in its library, allow the configuration of new PET prototypes via the FLUKA Graphical User Interface (GUI) Flair. Both the beam time structure and scan time can be specified by the user, reproducing PET acquisitions in time, in a particle therapy scenario. Furthermore, different scoring routines allow the analysis of single and coincident events, and identification of parent isotopes generating annihilation events. Two reconstruction codes are currently supported: the Filtered Back–Projection (FBP) and Maximum–Likelihood Expectation Maximization (MLEM), the latter embedded in the tools. Compatibility with other reconstruction frameworks is also possible. The FLUKA PET tools package has been successfully tested for different detectors and scenarios, including conventional functional PET applications and in beam PET, either using radioactive sources, or simulating hadron beam irradiations. The results obtained so far confirm the FLUKA PET tools suitability to perform PET simulations in R&D environment.