Monte Carlo Particle Transport Code Research Articles

Optimizations of the energy grid search algorithm in continuous-energy Monte Carlo particle transport codes

In this work we propose, implement, and test various optimizations of the typical energy grid-cross section pair lookup algorithm in Monte Carlo particle transport codes. The key feature common to all of the optimizations is a reduction in the length of the vector of energies that must be searched when locating the index of a particle’s current energy. Other factors held constant, a reduction in energy vector length yields a reduction in CPU time. The computational methods we present here are physics-informed. That is, they are designed to utilize the physical information embedded in a simulation in order to reduce the length of the vector to be searched. More specifically, the optimizations take advantage of information about scattering kinematics, neutron cross section structure and data representation, and also the expected characteristics of a system’s spatial flux distribution and energy spectrum. The methods that we present are implemented in the OpenMC Monte Carlo neutron transport code as part of this work. The gains in computational efficiency, as measured by overall code speedup, associated with each of the optimizations are demonstrated in both serial and multithreaded simulations of realistic systems. Depending on the system, simulation parameters, and optimization method employed, overall code speedup factors of 1.2–1.5, relative to the typical single-nuclide binary search algorithm, are routinely observed.

Improved calibration of mass stopping power in low density tissue for a proton pencil beam algorithm

Dose distributions for proton therapy treatments are almost exclusively calculated using pencil beam algorithms. An essential input to these algorithms is the patient model, derived from x-ray computed tomography (CT), which is used to estimate proton stopping power along the pencil beam paths. This study highlights a potential inaccuracy in the mapping between mass density and proton stopping power used by a clinical pencil beam algorithm in materials less dense than water. It proposes an alternative physically-motivated function (the mass average, or MA, formula) for use in this region. Comparisons are made between dose-depth curves calculated by the pencil beam method and those calculated by the Monte Carlo particle transport code MCNPX in a one-dimensional lung model. Proton range differences of up to 3% are observed between the methods, reduced to <1% when using the MA function. The impact of these range errors on clinical dose distributions is demonstrated using treatment plans for a non-small cell lung cancer patient. The change in stopping power calculation methodology results in relatively minor differences in dose when plans use three fields, but differences are observed at the 2%–2 mm level when a single field uniform dose technique is adopted. It is therefore suggested that the MA formula is adopted by users of the pencil beam algorithm for optimal dose calculation in lung, and that a similar approach is considered when beams traverse other low density regions such as the paranasal sinuses and mastoid process.

Simulations of a vacuum window and in-beam ion detection for use in radiobiological experiments with 50 MeV particle beams

The vacuum window and in-beam ion detection of a radiobiological endstation have been studied using the Monte Carlo particle transport code MCNPX 2.5.0, investigating the effects of different elements on an incident beam of 50 MeV protons. The scattering and energy spread as well as secondary particle generation are analysed. A 200 nm thick silicon nitride vacuum window was found to have the least effect on the incident beam, giving an energy spread of 5.01(8)× 10−4 MeV. Placing a silicon pixel detector before the sample caused scattering of the beam of around 15 μm in radius; too much to be used for experiments targeting individual nuclei.

Antiproton annihilation physics in the Monte Carlo particle transport code SHIELD-HIT12A

The Monte Carlo particle transport code SHIELD-HIT12A is designed to simulate therapeutic beams for cancer radiotherapy with fast ions. SHIELD-HIT12A allows creation of antiproton beam kernels for the treatment planning system TRiP98, but first it must be benchmarked against experimental data. An experimental depth dose curve obtained by the AD-4/ACE collaboration was compared with an earlier version of SHIELD-HIT, but since then inelastic annihilation cross sections for antiprotons have been updated and a more detailed geometric model of the AD-4/ACE experiment was applied.Furthermore, the Fermi–Teller Z-law, which is implemented by default in SHIELD-HIT12A has been shown not to be a good approximation for the capture probability of negative projectiles by nuclei. We investigate other theories which have been developed, and give a better agreement with experimental findings. The consequence of these updates is tested by comparing simulated data with the antiproton depth dose curve in water.It is found that the implementation of these new capture probabilities results in an overestimation of the depth dose curve in the Bragg peak. This can be mitigated by scaling the antiproton collision cross sections, which restores the agreement, but some small deviations still remain.Best agreement is achieved by using the most recent antiproton collision cross sections and the Fermi–Teller Z-law, even if experimental data conclude that the Z-law is inadequately describing annihilation on compounds. We conclude that more experimental cross section data are needed in the lower energy range in order to resolve this contradiction, ideally combined with more rigorous models for annihilation on compounds.

Target depth dependence of damage rate in metals by 150 MeV proton irradiation

A series of irradiation experiments with 150MeV protons was performed. The relationship between target depth (or shield thickness) and displacement damage during proton irradiation was obtained by in situ electrical resistance measurements at 20K. Positron annihilation lifetime measurements were also performed at room temperature after irradiation, as a function of the target thickness. The displacement damage was found to be high close to the beam incident surface area, and decreased with increasing target depth. The experimental results were compared with damage production calculated with an advanced Monte Carlo particle transport code system (PHITS).

Overview of particle and heavy ion transport code system PHITS

A general purpose Monte Carlo Particle and Heavy Ion Transport code System, PHITS, is being developed through the collaboration of several institutes in Japan and Europe. The Japan Atomic Energy Agency is responsible for managing the entire project. PHITS can deal with the transport of nearly all particles, including neutrons, protons, heavy ions, photons, and electrons, over wide energy ranges using various nuclear reaction models and data libraries. It is written in Fortran language and can be executed on almost all computers. All components of PHITS such as its source, executable and data-library files are assembled in one package and then distributed to many countries via the Research Organization for Information Science and Technology, the Data Bank of the Organization for Economic Co-operation and Development’s Nuclear Energy Agency, and the Radiation Safety Information Computational Center. More than 1500 researchers have been registered as PHITS users, and they apply the code to various research and development fields such as nuclear technology, accelerator design, medical physics, and cosmic-ray research. This paper briefly summarizes the physics models implemented in PHITS, and introduces some important functions useful for specific applications, such as an event generator mode and beam transport functions.

MC21 v.6.0 – A continuous-energy Monte Carlo particle transport code with integrated reactor feedback capabilities

MC21 is a continuous-energy Monte Carlo radiation transport code for the calculation of the steady-state spatial distributions of reaction rates in three-dimensional models. The code supports neutron and photon transport in fixed source problems, as well as iterated-fission-source (eigenvalue) neutron transport problems. MC21 has been designed and optimized to support large-scale problems in reactor physics, shielding, and criticality analysis applications. The code also supports many in-line reactor feedback effects, including depletion, thermal feedback, xenon feedback, eigenvalue search, and neutron and photon heating. MC21 uses continuous-energy neutron/nucleus interaction physics over the range from 10−5eV to 20MeV. The code treats all common neutron scattering mechanisms, including fast-range elastic and non-elastic scattering, and thermal- and epithermal-range scattering from molecules and crystalline materials. For photon transport, MC21 uses continuous-energy interaction physics over the energy range from 1keV to 100GeV. The code treats all common photon interaction mechanisms, including Compton scattering, pair production, and photoelectric interactions. All of the nuclear data required by MC21 is provided by the NDEX system of codes, which extracts and processes data from EPDL-, ENDF-, and ACE-formatted source files. For geometry representation, MC21 employs a flexible constructive solid geometry system that allows users to create spatial cells from first- and second-order surfaces. The system also allows models to be built up as hierarchical collections of previously defined spatial cells, with interior detail provided by grids and template overlays. Results are collected by a generalized tally capability which allows users to edit integral flux and reaction rate information. Results can be collected over the entire problem or within specific regions of interest through the use of phase filters that control which particles are allowed to score each tally. The tally system has been optimized to maintain a high level of efficiency, even as the number of edit regions becomes very large.

OpenMC: A state-of-the-art Monte Carlo code for research and development

This paper gives an overview of OpenMC, an open source Monte Carlo particle transport code recently developed at the Massachusetts Institute of Technology. OpenMC uses continuous-energy cross sections and a constructive solid geometry representation, enabling high-fidelity modeling of nuclear reactors and other systems. Modern, portable input/output file formats are used in OpenMC: XML for input, and HDF5 for output. High performance parallel algorithms in OpenMC have demonstrated near-linear scaling to over 100,000 processors on modern supercomputers. Other topics discussed in this paper include plotting, CMFD acceleration, variance reduction, eigenvalue calculations, and software development processes.

The contributions of source regions to organ doses from incorporated radioactive iodine

The separate contributions of all source regions to organ doses from 131 I and 123 I administered to the body were calculated using voxelized reference phantoms. The photon and electron components of organ doses were also evaluated for each source region. The MCNPX Monte Carlo particle transport code was utilized for dose calculations. All organs and tissues of male and female phantoms were taken into account as source regions with their corresponding cumulated activities. The results showed that cumulated activities assigned to source regions and inter-organ distances were two factors that strongly affected the contribution of each source to the organ dose. The major contribution of the dose to the main source regions arose from self-irradiation of electrons, while for nearby organs it was due to photons emitted by the main source organs. In addition, self-irradiation plays an important role in the dose delivered to most target organs for lower thyroid uptakes.

Current status of the “Hybrid Kurotama model” for total reaction cross sections

To be able to calculate the nucleon+nucleus and nucleus+nucleus total reaction cross sections with precision is of great importance for studies of fundamental nuclear properties, e.g., the nuclear structure. This is also very important for particle and heavy ion transport calculations since in all particle and heavy ion transport codes, the probability function according to which a projectile particle will collide within a certain distance in a matter depends on the total reaction cross sections. This will also scale the calculated partial fragmentation cross sections. It is therefore crucial that accurate total reaction cross section models are used in the transport calculations. In this paper, a new general purpose total reaction cross section model/subroutine called “Hybrid Kurotama” is presented. The model has been tested against available p+He, p+nucleus, and nucleus+nucleus total reaction cross sections and an overall better agreement has been found than for earlier published models. This model is therefore very suitable to be used in any deterministic or Monte Carlo particle and heavy ion transport code.

Conversion of ICRP male reference phantom to polygon-surface phantom

The International Commission on Radiological Protection (ICRP) reference phantoms, developed based on computed tomography images of human bodies, provide much more realism of human anatomy than the previously used MIRD5 (Medical Internal Radiation Dose) mathematical phantoms. It has been, however, realized that the ICRP reference phantoms have some critical limitations showing a considerable amount of holes for the skin and wall organs mainly due to the nature of voxels of which the phantoms are made, especially due to their low voxel resolutions. To address this problem, we are planning to develop the polygon-surface version of ICRP reference phantoms by directly converting the ICRP reference phantoms (voxel phantoms) to polygon-surface phantoms. The objective of this preliminary study is to see if it is indeed possible to construct the high-quality polygon-surface phantoms based on the ICRP reference phantoms maintaining identical organ morphology and also to identify any potential issues, and technologies to address these issues, in advance. For this purpose, in the present study, the ICRP reference male phantom was roughly converted to a polygon-surface phantom. Then, the constructed phantom was implemented in Geant4, Monte Carlo particle transport code, for dose calculations, and the calculated dose values were compared with those of the original ICRP reference phantom to see how much the calculated dose values are sensitive to the accuracy of the conversion process. The results of the present study show that it is certainly possible to convert the ICRP reference phantoms to surface phantoms with enough accuracy. In spite of using relatively less resources (<2 man-months), we were able to construct the polygon-surface phantom with the organ masses perfectly matching the ICRP reference values. The analysis of the calculated dose values also implies that the dose values are indeed not very sensitive to the detailed morphology of the organ models in the phantom for highly penetrating radiations such as photons and neutrons. The results of the electron beams, on the other hand, show that the dose values of the polygon-surface phantom are higher by a factor of 2–5 times than those of the ICRP reference phantom for the skin and wall organs which have large holes due to low voxel resolution. The results demonstrate that the ICRP reference phantom could provide significantly unreasonable dose values to thin or wall organs especially for weakly penetrating radiations. Therefore, when compared to the original ICRP reference phantoms, it is believed that the polygon-surface version of ICRP reference phantoms properly developed will not only provide the same or similar dose values (say, difference <5 or 10%) for highly penetrating radiations, but also provide correct dose values for the weakly penetrating radiations such as electrons and other charged particles.

PHITS simulations of the Protective curtain experiment onboard the Service module of ISS: Comparison with absorbed doses measured with TLDs

“Protective curtain” was the physical experiment onboard the International Space Station (ISS) aimed on radiation measurement of the dose – reducing effect of the additional shielding made of hygienic water-soaked wipes and towels placed on the wall in the crew cabin of the Service module Zvezda. The measurements were performed with 12 detector packages composed of thermoluminescent detectors (TLDs) and plastic nuclear track detectors (PNTDs) placed at the Protective curtain, so that they created pairs of shielded and unshielded detectors.We simulated the experiment by the general purpose 3D Monte Carlo Particle and Heavy Ion Transport code System (PHITS), as 10cm-thick water-filled panels housed in a model of the Zvezda module. External radiation environment was modeled using the AP8MIN and ISO-15390 standard models for the trapped proton (TP) and galactic cosmic ray (GCR) spectra, respectively. The absorbed doses were calculated for all detector packages used in the experiment. Comparison of calculated results with experimental data (TLDs) showed good agreement for the total (TP+GCR) absorbed doses. Further, we analyzed the systematic uncertainty introduced by differences in the detector thicknesses used in the simulations from the ones used in the measurements.The reducing effect of the Protective curtain was studied by comparing the calculated absorbed doses in shielded and unshielded detectors separately for the TPs and GCRs. In case of TPs, the reducing effect was larger than 60% and 40% for pairs of detectors located at aluminum wall and at crew cabin window, respectively. In case of GCRs, small shielding effect was observed for detectors located behind the window but for those located behind the aluminum wall, the effect was even opposite: the absorbed doses in the unshielded detectors were about 10% lower than in the shielded ones. This result was confirmed by the depth-dose analysis using rectangular source emitting broad parallel incident particles impinging on the simple geometry composed of aluminum/glass box and water box of variable thickness simulating the spacecraft wall/window and Protective curtain, respectively. The additional dose in the shielded detectors is related to the secondary fragments known as the “wall effect”. However, since GCR contributes by about 30% and 15% only to the total dose in water in shielded and unshielded detectors, respectively, the total shielding effect is high and the Protective curtain is very efficient when it is applied on a spacecraft at low-Earth orbits.

Determination of transit dose profile for a 192Ir HDR source

Several studies have reported methodologies to calculate and correct the transit dose component of the moving radiation source for high dose rate (HDR) brachytherapy planning systems. However, most of these works employ the average source speed, which varies significantly with the measurement technique used, and does not represent a realistic speed profile, therefore, providing an inaccurate dose determination. In this work, the authors quantified the transit dose component of a HDR unit based on the measurement of the instantaneous source speed to produce more accurate dose values. The Nucletron microSelectron-HDR Ir-192 source was characterized considering the Task Group 43 (TG-43U1) specifications. The transit dose component was considered through the calculation of the dose distribution using a Monte Carlo particle transport code, MCNP5, for each source position and correcting it by the source speed. The instantaneous source speed measurements were performed in a previous work using two optical fibers connected to a photomultiplier and an oscilloscope. Calculated doses were validated by comparing relative dose profiles with those obtained experimentally using radiochromic films. TG-43U1 source parameters were calculated to validate the Monte Carlo simulations. These agreed with the literature, with differences below 1% for the majority of the points. Calculated dose profiles without transit dose were also validated by comparison with ONCENTRA(®) Brachy v. 3.3 dose values, yielding differences within 1.5%. Dose profiles obtained with MCNP5 corrected using the instantaneous source speed profile showed differences near dwell positions of up to 800% in comparison to values corrected using the average source speed, but they are in good agreement with the experimental data, showing a maximum discrepancy of approximately 3% of the maximum dose. Near a dwell position the transit dose is about 22% of the dwell dose delivered by the source dwelling 1 s and reached 104.0 cGy per irradiation in a hypothetical clinical case studied in this work. The present work demonstrated that the transit dose correction based on average source speed fails to accurately correct the dose, indicating that the correct speed profile should be considered. The impact on total dose due to the transit dose correction near the dwell positions is significant and should be considered more carefully in treatments with high dose rate, several catheters, multiple dwell positions, small dwell times, and several fractions.

High accuracy tritium measurement for the verification of the tritium production rate calculations with MCNPX

This paper presents high accuracy tritium production rate measurement results compared with calculations using the MCNPX Monte Carlo particle transport code. The experimental results are regarded as reference values for a new passive technique based on the secondary charged particle activation method developed for measuring the tritium production rate in the test blanket modules of the ITER Tokamak. The 16O(t,n)18F reaction, which is one of the possible tritium monitor reactions, was also extensively investigated, and the experimentally determined reaction rates were compared with simulations.Li2CO3 solution was filled and sealed into quartz ampoules which were irradiated in the Training Reactor of the Budapest University of Technology and Economics. The amount of 18F was determined using γ-spectroscopy. Then the precise tritium measurements were carried out in the Hertelendi Laboratory of Environmental Studies using the 3H–3He ingrowth method, where the 3He produced during the storage time is measured by a static noble gas mass spectrometer (VG-5400).The HT/HTO ratio in the irradiated aqueous solutions was found to be 0.1323±0.0034. Based on the comparison of the measurements and the simulations it was pointed out that the model calculations underestimate the reaction rate of both the Li6(n,t)α and the 16O(t,n)18F reactions by 5–10% and 15%, respectively.

Simulation of the radiation fields of the Beta Secondary Standard BSS 2

Since several years, the irradiation facility for beta radiation, the Beta Secondary Standard BSS 2 developed at PTB, is in worldwide use to irradiate devices with calibrated beta sources. In this work the electron and photon particle spectra of the BSS 2 radiation fields are made available as data files, in addition angular distributions and the depth dose profiles are given. The spectra were determined using the Monte Carlo particle transport code BEAMnrc and are provided as electronic files. In order to verify the simulations, from the same simulations the depth dose curves in a phantom were deduced and compared with corresponding measurements — the agreement is quite good — proving the correctness of the particle spectra.

Neutronic design of a very high temperature reactor core with low graphite content

A relevant issue concerning the deployment of the very high temperature reactors is the accumulation of radioactive graphite and its final disposal. One way to reduce the amount of graphite in this type of reactors is to replace the reflector with other material. In this study the Gas Turbine Modular Helium Reactor was investigated, and most of the reflector was replaced with Ferritic–Martensitic steel. The Monte Carlo particle transport code MCNPX 2.6 was used. A fully detailed three dimensional core model was set up in order to evaluate the neutronic response of the modified core design. Results show that the proposed core performs well in terms of the neutron multiplication factor and reactivity coefficients. Considering a reactor life time of 60years, and taking into account 6years period for the replacement of the removable reflector, more than one thousand cubic meters of irradiated graphite are saved, which represents estimated economic saving of USD $237,541,000 and 7.73×1015Bq in radioactivity.

Evaluation of the secondary radiation impact on personnel during the dismantling of contaminated nuclear equipment

The article contains a numerical analysis of the secondary radiation contribution to the total radiation affecting the operational personnel during the dismantling activities of the contaminated equipment at a nuclear power plant. This study considers a widely applicable Monte Carlo particle transport code MCNPX and real Ignalina nuclear power plant records. A simplified albedo method is investigated in order to analyze the selected geometrical design cases. Additionally, the impact of the secondary radiation on the personnel dose was analyzed. The numerical MCNPX simulation allowed ascertaining the optimal distance between the source and the wall for the working personnel in closed rooms with contaminated equipment. The developed dose rate maps of the secondary radiation showed cross-sectional distribution of the dose rate inside the enclosed area.

蒙特卡罗粒子输运程序JMCT研制

蒙特卡罗粒子输运程序JMCT研制

Passive measurements of mixed-oxide fuel for nuclear nonproliferation

We present new results on passive measurements and simulations of mixed-oxide fuel-pin assemblies. Potential tools for mixed-oxide fuel pin characterization are discussed for future nuclear-nonproliferation applications. Four EJ-309 liquid scintillation detectors coupled with an accurate pulse timing and digital, offline and optimized pulse-shape discrimination method were used. Measurement analysis included pulse-height distributions to distinguish between purely fission neutron sources and alpha-n plus fission neutrons sources. Time-dependent cross-correlation functions were analyzed to measure the fission neutron contribution to the measured sample's neutron source. The use of Monte Carlo particle transport code MCNPX-PoliMi is discussed in conjunction with the measurements.

Shielding study at the Fukui Prefectural Hospital Proton Therapy Center

At the Fukui Prefectural Hospital Proton Therapy Center, neutron doses behind the concrete shields and at the maze were measured with three types of radiation monitors (DARWIN, Wendi-2, and a rem meter) along with solid-state nuclear track detectors. The measured data were compared with estimations of analytical models and the Monte Carlo code, Particle and Heavy-Ion Transport code System (PHITS). The analytical model, using the parameters employed in the shielding design of the facility, gave considerably larger values than the measured data. This means that the facility was designed with a sufficient margin of safety. The results calculated by PHITS were less than those of the analytical model and were about three times larger than the measured data. From a perspective that seeks conservative estimation with less margin, the Monte Carlo simulation is a good tool for shielding design of accelerator-based proton treatment facilities.