Monte Carlo Code Research Articles

A Generic Implementation of the D1S Methodology for Dose Rate Assessment by Monte Carlo Codes in Fusion Applications

Dose rate assessment is of utmost importance in fusion reactors because of the maintenance operations that these facilities will require. The standard way to perform this assessment in ITER is use of the Direct 1-Step (D1S) methodology, which consists of performing neutron transport simulation, material activation, and decay photon transport simulation in one step and thus in one simulation only. Usually, implementation of the D1S methodology requires changing the source files in a reference Monte Carlo code. The purpose of the present work is to develop an easy-to-implement method for Monte Carlo codes to help calculate the dose rate in fusion reactors. To do so, the proposal is to act on nuclear data only and not on source files of the simulation codes. This is done by replacing prompt photon production in evaluation files by suitable decay photon production, taking into account radioactive decays of radionuclides and irradiation history. This study was to be applied first on the TRIPOLI-4® Monte Carlo code for the sake of simplicity. TRIPOLI-4 is the reference code for particle transport at CEA. The verification and validation process relies first on a comparison to the reference Rigorous 2-Step (R2S) methodology and then on an experiment, the so-called Frascati Neutron Generator (FNG) dose experiment. The nuclear database to be changed is of the ENDF-6 format, a recurrent format in neutronics studies. The analysis studied both neutron and photon responses to check if the simulation was performing normally in a physical way and to compare the results with references provided by simulations based on the R2S methodology or by the FNG dose experiment. The simulations have proven to be in good agreement with the experimental results.

Comparison of the dosimetry of scandium-43 and scandium-44 patient organ doses in relation to commonly used gallium-68 for imaging neuroendocrine tumours

BackgroundSeveral research groups have explored the potential of scandium radionuclides for theragnostic applications due to their longer half-lives and equal or similar coordination chemistry between their diagnostic and therapeutic counterparts, as well as lutetium-177 and terbium-161, respectively. Unlike the gallium-68/lutetium-177 pair, which may show different in-vivo uptake patterns, the use of scandium radioisotopes promises consistent behaviour between diagnostic and therapeutic radiopeptides. An advantage of scandium’s longer half-life over gallium-68 is the ability to study radiopeptide uptake over extended periods and its suitability for centralized production and distribution. However, concerns arise from scandium-44’s decay characteristics and scandium-43’s high production costs. This study aimed to evaluate the dosimetric implications of using scandium radioisotopes with somatostatin analogues against gallium-68 for PET imaging of neuroendocrine tumours.MethodsAbsorbed dose per injected activity (AD/IA) from the generated time-integrated activity curve (TIAC) were estimated using the radiopeptides [43/44/44mSc]Sc- and [68Ga]Ga-DOTATATE. The kidneys, liver, spleen, and red bone marrow (RBM) were selected for dose estimation studies. The EGSnrc and MCNP6.1 Monte Carlo (MC) codes were used with female (AF) and male (AM) ICRP phantoms. The results were compared to Olinda/EXM software, and the effective dose concentrations assessed, varying composition between the scandium radioisotopes.ResultsOur findings showed good agreement between the MC codes, with − 3 ± 8% mean difference. Kidneys, liver, and spleen showed differences between the MC codes (min and max) in a range of − 4% to 8%. This was observed for both phantoms for all radiopeptides used in the study. Compared to Olinda/EXM the largest observed difference was for the RBM, of 21% for the AF and 16% for the AM for scandium- and gallium-based radiopeptides. Despite the differences, our findings showed a higher absorbed dose on [43/44Sc]Sc-DOTATATE compared to its 68Ga-based counterpart.ConclusionThis study found that [43/44Sc]Sc-DOTATATE delivers a higher absorbed dose to organs at risk compared to [68Ga]Ga-DOTATATE, assuming equal distribution. This is due to the longer half-life of scandium radioisotopes compared to gallium-68. However, calculated doses are within acceptable ranges, making scandium radioisotopes a feasible replacement for gallium-68 in PET imaging, potentially offering enhanced diagnostic potential with later timepoint imaging.

Investigating polyurethane foam loaded with high-z nanoparticles for gamma radiation shielding compared to Monte Carlo simulations

Since the beginning of research into radiation and protection against it, the importance of searching for proper materials against radiation hazards has been recognized. Gamma radiation protection materials usually deal with heavy elements with high prices, which are hard to maintain. Polyurethane-based (PU) materials are popular in sound and thermal insulation due to, their low-weight properties and, most importantly, fast and convenient construction ingredients. PU foams (PUF) can be used as radiation shield and noise and heat resistance due to their approachability, light-weight, high resistance, and comfortable construction. This study involved simulation and an experiment to construct and investigate the properties of Polyurethane material doped with lead oxide as a gamma shield. The shield was considered in several weight fractions of lead, yielding several samples. The MCNPX 2.6 Monte Carlo code has been utilized for simulation procedure, and 137Cs was employed as the gamma source in both simulation and experiment. The results offer a promising response against the gamma radiation and are suitable for attenuating gamma rays.

Inferring the iron K emissivity profiles of accretion discs irradiated by extended coronae

ABSTRACT One of the most promising methods to measure the spin of an accreting black hole is fitting the broad iron K$\alpha$ line in the X-ray spectrum. The line profile also depends on the geometry of the hard X-ray emitting corona. To put constraints on the black hole spin and corona geometry, it is essential to understand how do they affect the iron K$\alpha$ line emissivity profile. In this work, we present calculations of the illumination and the iron K$\alpha$ emissivity profiles performed with the Monte Carlo GR radiative transfer code monk. We focus on distinction between the illumination and emissivity profiles, which is in most previous studies neglected. We show that especially for the case of black hole X-ray binaries (BHXRBs), the difference is very large. For active galactic nuclei (AGNs), the emissivity profile has a more similar shape as the illumination profile, but it is notably steeper in the innermost region within a few gravitational radii. We find out that the different behaviour between AGN and BHXRB discs is due to the different energy spectra of the illuminating radiation. This suggests that the emissivity profile of the iron K$\alpha$ line cannot be determined by black hole spin and corona geometry alone and the energy spectrum of the illuminating radiation has to be taken into account. We also examined the effect of including the self-irradiation, and find it to be more important than the corona emission in BHXRBs.

Development and Verification of Sampling Code NECP-SOUL for Evaluated Nuclear Data Files in ENDF-6 Format

A code called NECP-SOUL is developed to generate samples of evaluated nuclear data files in the ENDF-6 format. The motivation for the development is to obtain more accurate uncertainty quantification results based on the covariance data given in evaluated nuclear data files than the sampling method based on multigroup covariances. NECP-SOUL can handle all covariance data in ENDF-6 files, including average fission neutron multiplicities, resonance parameters, cross sections, angular distributions, energy distributions, and cross sections for the production of radioactive nuclides. In this work, samples of ENDF-6 files are input to the nuclear data processing code to generate A Compact ENDF (ACE)-formatted libraries, and these ACE-formatted libraries are used by the Monte Carlo code. Through statistical analysis, the uncertainty of the neutronics results can be obtained. NECP-SOUL is verified against another similar code, SANDY. In addition, the advantage of sampling ENDF-6 files is analyzed by comparing the uncertainty quantification results with those obtained by sampling multigroup covariance matrices. Moreover, the covariances of cross sections and angular distributions newly provided in ENDF/B-VIII.0 are used to analyze their effect on the uncertainty qualification results.

Comparative Evaluation of Two Analytical Functions for the Microdosimetry of Ions from 1H to 238U

The analytical microdosimetric function (AMF) implemented in the Monte Carlo code PHITS is a unique tool that bridges the gap between macro- and microscopic scales of radiation interactions, enabling accurate microdosimetric calculations over macroscopic bodies. The original AMF was published in 2006, based on the results of track structure calculations. Recently, a newer version of the AMF was proposed, incorporating an improved description of the energy loss at the microscopic scale. This study compares the older and the newer AMFs in computing microdosimetric probability distributions, mean values, and the relative biological effectiveness (RBE). To this end, 16000 microdosimetric lineal energy probability density distributions were simulated with PHITS for ions from 1H to 238U over a broad energy range (1–1000 MeV/n). The newer AMF was found to offer superior performance, particularly for very heavy ions, producing results that align more closely with published in vitro clonogenic survival experiments. These findings suggest that the updated AMF provides a more reliable tool for microdosimetric calculations and RBE modeling, essential for ion radiation therapy and space radiation protection.

Application limits of the scaling relations for Monte Carlo simulations in diffuse optics. Part 2: results.

The limits of applicability of scaling relations to generate new simulations of photon migration in scattering media by re-scaling an existing Monte Carlo simulation are investigated both for the continuous wave and the time domain case. We analyzed the convergence properties in various scenarios by numerical methods, trying to derive practical guidelines for the judicious use of this approach, as well as a deeper understanding of the physics behind such relations. In the case of scaling of the absorption coefficient, the convergence is always rigorous both for the forward and inverse problems, relying on the derivatives with respect to the absorption coefficient. Also, the regenerated simulation inherits the very same noise of the original Monte Carlo simulation. In the case of scaling of the scattering coefficient, the situation is more critical. For forward problems, even for just a 10% uniform increase in scattering, appreciable deviations are observed whenever a high number of scattering interactions is involved. We tested a practical criterion based on the number of scattering events in the original simulation to judge the convergence of the scaling factors. For inverse problems, the scaling relations provide accurate regenerated simulations apart from the noise level that is increased with respect to the initial simulation, although anyway lower than the noise level obtained by implementing the direct calculation. The results of this study are important whenever an increase of Monte Carlo code throughput is mandatory, e.g., for fast data analysis of diffuse data, or in machine-learning scenarios, when generating huge datasets is needed.

Modelling Gd-diamond and Gd-SiC neutron detectors

A custom Monte Carlo (MC) computer model was developed to simulate thermal neutron absorption in, and subsequent photon and electron emission from, natural Gd with a view to using the material as a neutron conversion layer for neutron detectors. The MC code also modelled photon and electron detection with two dissimilar detectors: a thick (500 μm) single crystal diamond detector; and a thin (5.15 μm) commercial off the shelf (COTS) 4H–SiC photodiode detector. The detectors’ quantum detection efficiencies (QE) for hard X-rays and γ-rays were relatively low in comparison to their QE for electrons, thus making it possible to collect electron spectra from the Gd layer neutron conversion products which were not overwhelmed by photon emissions from the Gd. The MC code was utilised to determine the optimal thickness of Gd for the efficient detection of a thermal neutron flux. These radiation hard and spectroscopic detectors paired with natural Gd could find utility as robust and compact thermal neutron detectors for nuclear science and engineering, space science, and other applications.

Modeling the impact of coincidence loss on count rate statistics and noise performance in counting detectors for imaging applications

Coincidence loss can have detrimental effects on the image quality provided by pixelated counting detectors, especially in dose-sensitive applications like cryoEM where the information extracted from the recorded signal needs to be maximized. In this work, we investigate the impact of coincidence loss phenomena on the recorded statistics in counting detectors producing sparse binary images. First, we derive exact analytical expressions for the mean and the variance of the recorded counts as a function of the incoming event rate. Second, we address the problem of the mean and variance of the recorded events (i.e., pixel clusters identified as individual incoming events), which also acts as a function of the incoming event rate. In this frame, we review previous studies from different disciplines on approximated two-dimensional models, and we critically reinterpret them in our context and evaluate the suitability of their adoption in the present case. The knowledge of the first two momenta of the recorded statistics allows inferring about the signal-to-noise ratio (SNR) and the detective quantum efficiency at zero frequency (DQE0). Analytical results are validated through comparison with numerical data obtained with a custom-made Monte Carlo code. We chose a realistic case study for cryoEM application consisting of a 25-µm-thick MAPS detector featuring a pixel size of 10 µm and illuminated with electrons of 300 keV energy over a wide range of incoming rate.

Monte Carlo Source Convergence Diagnosis Method and Thermal-Physics Coupling Method Based on Functional Expansion Tallies

In iterative Monte Carlo calculations for nuclear reactors, the inactive cycles should be calculated first to ensure that the source distribution is converged, and then the tallies of various parameters in the active cycles can begin. In order to acquire the mesh-free distribution of the fission source, this research proposes the functional expansion tallies (FET) source convergence diagnosis method in the Reactor Monte Carlo code, which is a self-developed stochastic simulation code maintained by the Reactor Engineering Analysis Laboratory of Tsinghua University. Due to the randomness in Monte Carlo calculations and the difficulty in determining the precise source convergence, this paper proposes a diagnostic tool based on function curve similarity and moving average, and proposes an online real-time source convergence diagnosis method. The FET online source convergence method can terminate the calculation of the inactive cycle in real time according to the convergence diagnostic tool; thus it can greatly decrease the calculation time. The precise and effective transfer of data between different meshes is a difficult issue of thermal and physical coupling. Converting two separate meshes and transferring the data are exceptionally difficult and complex tasks within the conventional nuclear thermal-physics coupling approach. By applying the FET method to nuclear thermal-physics coupling, the mesh-free continuous-space fission source distribution can be obtained, which is suitable for more complex meshes. Additionally, computational memory can be minimized by replacing (transforming) the data from numerous mesh power distribution data points with the coefficients of the function.

Improving tally efficiency and accuracy of multi-group scattering matrix calculations in the Monte Carlo code NECP-MCX

Two issues arise in the calculation of the multi-group scattering matrix when employing a continuous-energy Monte Carlo code for generating homogenized multi-group cross-sections. Firstly, the analog estimator is used to evaluate group-to-group elements, which leads to large statistical uncertainty. Secondly, employing the scalar flux as the weighting function in generating the high-order scattering matrix introduces errors in fast reactor calculations. For the first issue, the repeated collision approach and pre-tabulated cross-section approach are adopted to improve the tally efficiency. For the second issue, the average scattering cosine is calculated based on the conservation of the mean square displacement of neutrons, which is then used to correct the first-order self-scattering cross-section. To evaluate the effectiveness of the above approaches, a PWR pin-cell problem and fast reactor core problems are tested. The results demonstrate that: 1) The figure of merit for multi-group scattering matrix calculations was improved by 8–12 times with the pre-tabulated cross-section approach. 2) Biases of keff were reduced from over 500 pcm to less than 300 pcm when using the corrected self-scattering cross-section. 3) The corrected self-scattering cross-section also yielded higher accuracy for the assembly power calculations, where the maximum biases are reduced from 5 % to 1 %.

Machine Learning for Predicting Neutron Effective Dose

The calculation of effective doses is crucial in many medical and radiation fields in order to ensure safety and compliance with regulatory limits. Traditionally, Monte Carlo codes using detailed human body computational phantoms have been used for such calculations. Monte Carlo dose calculations can be time-consuming and require expertise in different processes when building the computational phantom and dose calculations. This study employs various machine learning (ML) algorithms to predict the organ doses and effective dose conversion coefficients (DCCs) from different anthropomorphic phantoms. A comprehensive data set comprising neutron energy bins, organ labels, masses, and densities is compiled from Monte Carlo studies, and it is used to train and evaluate the supervised ML models. This study includes a broad range of phantoms, including those from the International Commission on Radiation Protection (ICRP-110, ICRP-116 phantom), the Visible-Human Project (VIP-man phantom), and the Medical Internal Radiation Dose Committee (MIRD-Phantom), with row data prepared using numerical data and organ categorical labeled data. Extreme gradient boosting (XGB), gradient boosting (GB), and the random forest-based Extra Trees regressor are employed to assess the performance of the ML models against published ICRP neutron DCC values using the mean square error, mean absolute error, and R2 metrics. The results demonstrate that the ML predictions significantly vary in lower energy ranges and vary less in higher neutron energy ranges while showing good agreement with ICRP values at mid-range energies. Moreover, the categorical data models align closely with the reference doses, suggesting the potential of ML in predicting effective doses for custom phantoms based on regional populations, such as the Saudi voxel-based model. This study paves the way for efficient dose prediction using ML, particularly in scenarios requiring rapid results without extensive computational resources or expertise. The findings also indicate potential improvements in data representation and the inclusion of larger data sets to refine model accuracy and prevent overfitting. Thus, ML methods can serve as valuable techniques for the continued development of personalized dosimetry.

Mini-ridge filter designs for conformal FLASH proton therapy

The present study designed a mini-ridge filter capable of generating a 30 mm spread-out Bragg peak (SOBP) specific to the proton FLASH beamline for use in proton FLASH mouse studies. Proton FLASH therapy delivers ultra-high dose rates exceeding 40 Gy/s, necessitating instantaneous SOBP generation via static method. The TOol for PArticle Simulation (TOPAS) was used to model the FLASH beamline, which enabled accurate replication of beam characteristics identified from measurements, and ensured accurate dose calculations. Comparisons between simulated and measured peak widths and R80d values, accurate to within 1 mm, verified the accuracy and reliability of Monte Carlo code. Two mini-ridge filters of different materials were designed and optimized using MATLAB, simulated with TOPAS for SOBP verification, and fabricated using 3D-printing or computer numerical control manufacturing. The results indicated strong agreement between the simulated and measured SOBP widths for the polylactic acid mini-ridge filter but with discrepancies observed in the range and flatness measurements, which may be due to the use of 3D printing. By contrast, the aluminium alloy mini-ridge filter achieved excellent results, generating a 31.6 mm SOBP width with 0.89% flatness. Therefore, the present study validated the accuracy of our Monte Carlo code in simulating the FLASH beamline and the feasibility of our mini-ridge filters.

Analytical and fine Monte Carlo Kerma coefficients evaluation for a wide range photon energy relevant to some chemotherapy agents

Simultaneously combination of chemotherapy drugs and ionizing radiations has become a common treatment method for advanced cancers. Tumor irradiation in presence of chemotherapy drugs can affect deposited energy within tissue. Hence, to evaluate how chemotherapy drugs affect the absorbed dose within tissues during chemoradiotherapy, Kerma coefficient values for 26 anti-cancer agents were estimated for photon energies ranging from 1 keV to 20 MeV through analytical analysis and Monte Carlo (MC) simulation approaches.Estimation of Kerma coefficients through MC simulation relevant to different chemotherapy drugs was performed with MCNPX V.2.6.0 MC code. Besides, Kerma coefficient values were calculated using the reported mass energy absorption coefficients, as the theoretical method. Significant statistical test was also employed to evaluate the accordance of obtained Kerma coefficients by two considered methods.Simulation results were consistent with evaluated Kerma coefficients by analytical method which minimum and maximum mean relative differences are found for Bevacizumab and Methotrexate, respectively. Furthermore, a significant correlation is found with chemical composition of selected chemotherapy drugs at considered photon energy range. Fluorouracil showed the closest Kerma ratio to unity among the studied anti-cancer drugs. Kerma coefficient ratios for most of selected drugs were close to unity at higher energy regions (100 keV–20 MeV) while greater values were found at lower photon energies (below 100 keV).From the results, presence of these anti-cancer agents can impact Kerma coefficient and subsequently alter the absorbed dose within tissue during the cancer treatment with chemoradiotherapy technique.

Analysis of the atomic ratio of H and Zr effect on the neutronics parameters of ZrH moderated space nuclear reactor

Zirconium hydride (ZrH) is an ideal moderator for space nuclear reactors due to its exceptional properties, including high hydrogen content, small neutron absorption cross section, and high operating temperature. The primary goal of this study is to examine how the atomic ratio of H to Zr (H/Zr ratio) influences the neutronics parameters of a ZrH moderated space nuclear reactor, with a focus on establishing a reliable reference for ensuring the optimal safety and minimization of such reactors. The neutronics calculation based on the ZrH moderated space nuclear reactor named Topaz-II is performed using the Reactor Monte Carlo code (RMC code) with the ENDF/VII cross-section database. The effects of the H/Zr ratio were studied with a particular focus on the initial keff, the burnup, the temperature reactivity coefficient and the criticality safety. The results show that with the increase of the H/Zr ratio, the initial keff increases while the drums’ worth decreases. The Moderator Temperature Coefficient (MTC) is a positive value that rises as the H/Zr ratio increases. In the dropping accidents, the reactor full of voids with seawater is more serious, and the introduced reactivity decreases with the increase of the H/Zr ratio.

Artificial High- and Low-density Materials in Bone Mineral Densitometry Using Dual-energy X-ray Absorptiometry: A GATE Monte Carlo Simulation of "Black-hole" Artifact.

The objective of the study was to evaluate the effect of artificial high- and low-density materials on Bone mineral density (BMD)scans in dual-energy X-ray absorptiometry (DXA) method and emergence of black-hole artifact through GATE Monte Carlo simulation. GATE Monte Carlo code was utilized to simulate the artifact encountered in clinical scans acquired by HOLOGIC® bone densitometer. Two simplified phantoms were designed. The first one was a rectangular box with six smaller cubes inside and the second one was a body torso. Materials of cubes were spine bone, polymethyl methacrylate (PMMA), barium sulfate suspension in water, stainless steel, titanium alloy, and gold. The torso phantom contained objects of 5 vertebrae, bowel and 3 small spherical objects near the surface of the torso as piercing objects on the abdominal wall, each overlying the vertebrae. Using 100 and 140 kVp, spectral X-rays were generated to simulate DXA. For both phantoms, two simulations were carried out. The pair of projections acquired for each phantom were then subtracted and analyzed by curve fitting techniques. Except for spine bone, in which radio-opacity decreases with increasing spectral X-ray energy (from 100 to 140 kVp), other squares exhibit little changes over different energies. PMMA shows consistently very low radio-opacity. Four other materials (barium sulfate in water, stainless steel alloy, titanium alloy, and gold), all attenuate the X-ray photons substantially. Except for spine bone, other materials are barely noticeable in pairwise subtracted images. In torso phantom, piercing objects are visualized as "holes" in vertebrae. Both artificial high- and low-density materials, compared to bone, are eliminated during the subtraction of dual-energy X-ray profiles and therefore, can create black-hole artifact.

Neutron balance approach for inline critical rod position search calculation in Monte Carlo reactor analysis

The critical rod position search has been one of the issues in the Monte Carlo (MC) reactor analysis. This paper proposes a practical and simple approach for the critical rod position search based on the neutron balance equation and the monotonic relation between the control rod absorption rate and the total rod insertion length. The proposed method was implemented in an inline manner within the McCARD, the MC reactor design code, so that the critical rod position is calculated and updated for each fission source iteration. The numerical results in a typical fast reactor problem demonstrate the promising performance of the proposed method in the critical rod position search calculation. Furthermore, the cycle depletion calculation was performed to show the capability of the McCARD to simulate the rodded operation condition. By addressing the critical rod position search problem through the neutron balance approach, the applicability of the MC code in the advanced reactor design will be expanded.

A comparative investigation on mechanical, gamma-ray and neutron shielding properties of some iron and boron containing concretes samples for nuclear safety applications

This study explores the gamma-ray and neutron shielding properties of fourteen different concrete samples, each tailored with varying percentages of Boron Carbide, Iron, and Iron Boride. Using the MCNP 6.3 Monte Carlo code, we calculated transmission factors for photon energies of 0.662 MeV, 1.1732 MeV, and 1.3325 MeV, and analyzed the impact of concrete thickness on shielding efficacy. Additionally, the Phy-X/PSD software was used to compute critical parameters such as linear and mass attenuation coefficients, half-value layer, tenth-value layer, mean free path, and fast neutron removal cross-section to gain a comprehensive understanding of each material's shielding capabilities. Our findings indicate that adding iron to the concrete matrix significantly enhances its attenuation properties, with the 20%Fe+80%Concrete sample outperforming all others. This composition demonstrated the lowest transmission factors across all tested energies and thicknesses, indicating superior photon attenuation. Moreover, the 20%Fe+80% Concrete exhibited the highest fast neutron removal cross-section, making it highly effective for environments requiring neutron shielding. In addition to the shielding properties, we analyzed the Elastic (Young's) Modulus of the concrete samples to understand their mechanical properties. Standard Concrete had an Elastic Modulus of 261.24 GPa, while the introduction of boron carbide significantly enhanced the Elastic Modulus, with pure boron carbide concrete exhibiting a value of 518.88 GPa. Concrete samples with varying percentages of boron carbide (5%, 10%, 15%, and 20%) showed a progressive increase in Elastic Modulus, indicating that higher proportions of boron carbide consistently enhance the material's stiffness. Conversely, concrete samples with iron boride and iron showed slight reductions in Elastic Modulus. It can be concluded that the boron carbide enhances stiffness, iron and iron boride provide a balance between stiffness and other properties. In conclusion, the 20%Fe+80%Concrete is a standout material that could greatly improve radiation shielding, offering major benefits.

Range Different of Mono-Energetic 290 MeV/u Carbon Ion Across Geant4 Version with Variation of I-values

The mean ionization potential (I-value) is a primary determinant of the position of the heavy ion Bragg peak. To minimize their impact on beam range errors and quantify their uncertainties, the currently used I-values in Geant4 material database are revisited. The study aims at comparing set of I-values in different Geant4 versions and cross validation with PHITS Monte Carlo (MC) code. The Bragg curves of mono-energetic 290 MeV/u carbon ion beams in a Polymethyl Methacrylate (PMMA) phantom were simulated using Geant4 versions 10.6.2 and 11.2.1. Similar beam energies were replicated using the PHITS code. The Bragg curves showed good agreement between the two versions of Geant4 MC code for an I-value of 74 eV, corresponding to G4_PLEXIGLASS in the material database. When the I-value was lowered to 65 eV and 48 eV, the Bragg curves from Geant4 version 11.2.1 shifted to shallower depths. This research provides insights for evaluating the Geant4 physics model.

Development of neutron-photon coupling transport code IMPC-NP and optimization of method for fabrication of neutron-photon multi-group cross section

This paper introduces a Monte Carlo code IMPC-NP which can realize the simulation of neutron-photon coupling transport. In IMPC-NP, Delta-tracking algorithm and Weight delta-tracking algorithm are used for neutron and photon transport. A new method of fabrication of photon multi-group cross sections named "Scattering merging" is proposed, which is implemented and on IMPC-NP. The results of multiple benchmarks show that the new method can greatly improve the accuracy of neutron and photon calculation in deterministic code.