Monte Carlo Code Research Articles

An extended Monte Carlo simulation code for modeling gas chromatography experiments with superheavy elements and their homologs

AbstractMonte Carlo simulations are commonly used to model the behavior of chemical species of the heaviest elements and their homologs in gas chromatography experiments. In this paper, we present an extension of the fundamental Monte Carlo simulation proposed by Zvara in 1985. While preserving the core functionality, our code features two enhancements: first, it allows simulating experiments in which a primary radioisotope decays into a daughter isotope belonging to a different element, hence exhibiting different chemical properties. Second, it allows modeling scenarios where conversion of an initial chemical species to a different one can occur at temperatures high enough to overcome an activation barrier, facilitating simulations of related physisorption and chemisorption processes. This Monte Carlo code is applicable to open tubular and rectangular chromatography columns.

Source of the Fe Kalpha emission in T Tauri stars. Radiation induced by relativistic electrons during flares. An application to RY Tau

T Tauri stars (TTSs) are magnetically active stars that accrete matter from the inner border of the surrounding accretion disk; plasma becomes trapped into the large-scale magnetic structures and falls onto the star, heating the surface through the so-called accretion shocks. The X-ray spectra of the TTSs show prominent Fe Kalpha fluorescence emission at 6.4 keV (hereafter, Fe Kalpha emission) that cannot be explained in a pure accretion scenario because its excitation requires significantly more energy than the maximum available through the well-constrained free-fall velocity. Neither can it be produced by the hot coronal plasma. TTSs display all signs of magnetic activity, and magnetic reconnection events are expected to occur frequently. In these events, electrons may become accelerated to relativistic speeds, and their interaction with the environmental matter may result in Fe Kalpha emission. It is the aim of this work to evaluate the expected Fe Kalpha emission in the context of the TTS research and compare it with the actual Fe Kalpha measurements obtained during the flare detected while monitoring RY Tau with the XMM-Newton satellite. The propagation of high-energy electrons in dense gas generates a cascade of secondary particles that results in an electron shower of random nature, whose evolution and radiative throughput was simulated in this work using the Monte Carlo code PENELOPE. A set of conditions representing the environment of the TTSs where these showers may impinge was taken into account to generate a grid of models that can aid the interpretation of the data. The simulations show that the electron beams produce a hot spot at the point of impact; strong Fe Kalpha emission and X-ray continuum radiation are produced by the spot. This emission is compatible with RY Tau observations. The Fe Kalpha emission observed in TTSs could be produced by beams of relativistic electrons accelerated in magnetic reconnection events during flares.

Gap in drop collision rate between diffusive and inertial regimes explains the stability of fogs and non-precipitating clouds

Rain drops form in clouds by collision of submillimetric droplets falling under gravity: larger drops fall faster than smaller ones and collect them on their path. The puzzling stability of fogs and non-precipitating warm clouds with respect to this avalanche mechanism has been a longstanding problem. How can droplets of diameter around 10 $\mathrm {\mu }$ m have a low collision probability, inhibiting the cascade towards larger and larger drops? Here we review the dynamical mechanisms that have been proposed in the literature and quantitatively investigate the frequency of drop collisions induced by Brownian diffusion, electrostatics and gravity, using an open-source Monte Carlo code taking all of them into account. Inertia dominates over aerodynamic forces for large drops, when the Stokes number is larger than $1$ . Thermal diffusion dominates over aerodynamic forces for small drops, when the Péclet number is smaller than $1$ . We show that there exists a range of size (typically 3–30 $\mathrm {\mu }$ m for water drops in air) where neither inertia nor Brownian diffusion are significant, leading to a gap in the collision rate. The effect is particularly important, due to the lubrication film forming between the drops immediately before collision, and secondarily to the long-range aerodynamic interaction. Two different mechanisms regularise the divergence of the lubrication force at vanishing separation: the transition to a non-continuum regime in the lubrication film, when the separation is comparable to the mean free path of air, and the induction of a flow inside the drops due to shear at their surfaces. In the gap between inertia-dominated and diffusion-dominated regimes, dipole–dipole electrostatic interactions becomes the major effect controlling the efficiency of drop collisions.

DNA-damage RBE assessment for combined boron and gadolinium neutron capture therapy.

Neutron capture therapy (NCT) by 10B and 157Gd agents is a unique irradiation-based method which can be used to treat brain tumors. Current study aims to quantitatively evaluate the relative biological effectiveness (RBE) and dose distributions during the combined BNCT and GdNCT modalities through a hybrid Monte Carlo (MC) simulation approach. Snyder head phantom as well as a cubic hypothetical tumor was at first modeled by Geant4 MC Code. Then, the energy spectra and dose distribution relevant to the released secondary particles during the combined Gd/BNCT were scored for different concentrations of 157Gd and 10B inside tumor volume. Finally, the scored energy spectra were imported to the MCDS code to estimate both RBESSB and RBEDSB values for different 157Gd concentrations. The results showed that combined Gd/BNCT increases the fluence-averaged RBESSB values by about 1.7 times when 157Gd concentration increments from 0 to 2000µg/g for both considered cell oxygen levels (pO2=10% and 100%). Besides, a reduction of about 26% was found for fluence-averaged RBEDSB values with an increment of 157Gd concentration in tumor volume. From the results, it can be concluded that combined Gd/BNCT technique can improve tumor coverage with higher dose levels but in the expense of RBEDSB reduction which can affect the clinical efficacy of the NCT technique.

Enhancing gamma radiation shielding properties of iron metal and natural rubber composites

AbstractA comparative study was conducted to investigate the gamma and neutron radiation shielding properties of composites made from iron metal (micro and nano) dispersed within natural rubber (NR). The aim was to attenuate neutron and γ‐rays. The study utilized the gamma spectroscopy technique and the Monte Carlo FLUKA code. To assess the gamma ray shielding, a 3×3 Nal (Tl) detector and radioactive point sources were employed. Two comparative studies were carried out, one using Fe loaded NR (Fe/NR) composites in microsize and the other using Fe/NR composites in nanosize. The objective was to determine the impact of iron nanoparticles on the shielding ability of the composites. The findings revealed that the Fe/NR composites in nanosize exhibited superior gamma radiation shielding ability compared with the Fe/NR composites in microsize. These results were supported by the FLUKA Monte Carlo code, which demonstrated good agreement with both the experimental and theoretical XCOM data.

Measurement of effective 98Mo(n,γ)99 Mo reaction cross-section in the epi-cadmium neutron flux produced by 10 MeV electron LINAC with Ta-BeO assembly

AbstractAn experiment was carried out to study the feasibility of 99Mo production using an electron accelerator based epi-cadmium neutron source. The neutrons were produced by the interaction of bremsstrahlung end-point energy of 10 MeV with BeO blocks and the neutrons thus produced were moderated using High Density Polyethylene blocks. Effective cross-section of 98Mo(n,γ)99Mo reaction was measured using foil activation method. An enhancement in the effective cross-section due to resonances in the epi-cadmium region was observed. The photons and neutron production/transport in the experimental set-up were simulated using the FLUKA Monte Carlo code. The simulations were found to be in good agreement with the experimental findings.

Performance evaluation and validation of OpenMC code for criticality analysis of an MTR-type research reactor

This paper assesses the performance of OpenMC, an open-source Monte Carlo code for particle transport, through criticality analysis of an MTR-type research reactor. The evaluation involves validation against experimental data from the NUR research reactor's first fresh fuel core and a code-to-code comparison with MCNP5, using a consistent methodology ensuring identical input conditions. Key parameters, including excess reactivity, control rod worths, shutdown margin, critical configurations, kinetics parameters, neutron flux, power distributions, and peaking factors, were examined. Results indicate good agreement between OpenMC simulations, experimental data, and MCNP5 results, with marginal differences well within acceptable ranges. Discrepancies with experimental data are below 5 %, considering measurement accuracy, while code-to-code comparisons reveal relative differences below 1 % in most cases. This study supports OpenMC's reliability for MTR reactor criticality analysis, affirming the validity of the developed NUR reactor model for future applications and highlighting the code's credibility and applicability in the nuclear reactor simulations.

CAD and constructive solid geometry modeling of the Molten Salt Reactor Experiment with OpenMC

In this study, we present a detailed comparison of two independently developed models of the Molten Salt Reactor Experiment (MSRE) for Monte Carlo particle transport simulations: the constructive solid geometry (CSG) model that was developed in support of the MSRE benchmark in the International Handbook of Evaluated Reactor Physics Benchmark Experiments, and a CAD model that was developed by Copenhagen Atomics. The original Serpent reference CSG model was first converted to OpenMC’s input format so that it could be systematically compared to the CAD model, which was already available as an OpenMC model, using the same Monte Carlo code. Results from simulations using the Serpent and OpenMC CSG models showed that keff agreed within 10 pcm while the flux distribution in space and energy generally agreed within 0.1%. Larger differences were observed between the OpenMC CAD and CSG models; notably, the keff computed for the CAD model was 1.00872, which is more than 1% lower than the value for the CSG model and much closer to experiment. Several areas of the reactor that were modeled differently in the CSG and CAD models were discussed and, in several cases, their impact on keff was quantified. Lastly, we compared the computational performance and memory usage between the CAD and CSG models. Simulation of the CSG model was found to be 1.4–2.3× faster than simulation of the CAD model based on the Embree ray tracer while using 4× less memory, highlighting the need for continued improvements in the CAD-based particle transport ecosystem. Finally, major performance degradation was observed for CAD-based simulations when using the MOAB ray tracer.

A [formula omitted]-radioisotope battery based on betavoltaic and beta-photovoltaic dual effects

Betavoltaic batteries with strontium sources are expected to achieve higher output power because of their high energy density and low self-absorption rate with the development of MEMS. However, the high energy of beta particles emitted from 90Sr/90Y can prone to radiation damage of the semiconductor. In order to exploit the high energy of beta particles emitted from 90Sr/90Y and to avoid radiation damage to the semiconductor, this paper presents a 90Sr/90Y-radioisotope battery based on betavoltaic (BV) and beta-photovoltaic (BPV) dual effects. In the work, the energy deposition of beta particles in LYSO:Ce and GaAs was simulated by Monte Carlo code, and thickness of the scintillator was determined. And the doping concentrations and junction depth of semiconductor were optimized based on the theoretical calculations to obtain the best output performance of device. When the thickness of LYSO:Ce is 0.158 cm, the output power density Pm of the optimized dual-effect battery is 0.61 μW/cm2. And the conversion efficiency of the device is 0.92%. At this time, the doping concentrations are Na=1.58×1017cm−3 and Nd=3.16×1018cm−3, and the junction depth xj=0.05μm. All calculated parameter values are considered as theoretical limit values. In addition, the contribution of BV effect and BPV effect to the output performance of the dual-effect radioisotope battery was investigated. Different scintillator thicknesses lead to different percentages of the two mechanics. In addition, the BV effect and BPV effect output proportion is also affected by the average energy of the radiation source. In the case that the average electron energy on the semiconductor surface is 0.27 MeV, the higher the radioactive source energy, the thicker the scintillator is required, resulting in more BPV effect and less BV effect.

Neutronic design analysis of CEA Cadarache URANIE subcritical assembly

This paper aims to study neutronic modeling using decommissioned fuel pins of the URANIE subcritical assembly at the Cadarache center of CEA for the future Tunisian Subcritical Assembly. Indeed, the CNSTN intends to exploit the fuel to implementing its subcritical assembly. Therefore, a comprehensive computational study of the URANIE fuel was carried out to analyze the neutronic parameters for future facilities including criticality safety, the neutron flux, and neutron spectrum, for future facilities. Indeed, MCNP Monte Carlo code was used to develop a three-dimensional sketch for URANIE core. Two versions were used to estimate an effective multiplication factor (keff): MCNP5 and MCNP6. The keff was varied from 0.85842 to 0.86463 with a maximal difference of 623 pcm and depended on ENDF/B data library releases. The nominal value of keff is acquired at the optimum lattice pitch of 4.6 cm, 30 cm in reflector thickness, and 185 cm in moderator level. Moreover, the reactivity of the facility is always negative. In addition, the energy spectrum was depicted at the three planes using 283 energy groups. Furthermore, the radial neutron flux was evaluated at eight positions. The axial neutron flux was estimated at various spatial planes of the assembly core. Inside the reactor core, the findings revealed indicate that the fast flux is higher and can be reach 7 × 10−4 n/cm2.Q and decreased steadily moving away from the center to the reflector part about 8 × 10−4 n/cm2.Q.

Characterizing the voxel-based approaches in radioembolization dosimetry with reDoseMC.

Yttrium-90 ( ) represents the primary radioisotope used in radioembolization procedures, while holmium-166 ( ) is hypothesized to serve as a viable substitute for due to its comparable therapeutic potential and improved quantitative imaging. Voxel-based dosimetry for these radioisotopes relies on activity images obtained through PET or SPECT and dosimetry methods, including the voxel S-value (VSV) and the local deposition method (LDM). However, the evaluation of the accuracy of absorbed dose calculations has been limited by the use of non-ideal reference standards and investigations restricted to the liver. The objective of this study was to expand upon these dosimetry characterizations by investigating the impact of image resolutions, voxel sizes, target volumes, and tissue materials on the accuracy of and dosimetry techniques. A specialized radiopharmaceutical dosimetry software called reDoseMC was developed using the Geant4 Monte Carlo toolkit and validated by benchmarking the generated kernels with published data. The decay spectra of both and were also compared. Multiple VSV kernels were generated for the liver, lungs, soft tissue, and bone for isotropic voxel sizes of 1 mm, 2 mm, and 4 mm. Three theoretical phantom setups were created with 20 or 40mm activity and mass density inserts for the same three voxel sizes. To replicate the limited spatial resolutions present in PET and SPECT images, image resolutions were modeled using a 3D Gaussian kernel with a Full Width at Half Maximum (FWHM) ranging from 0 to 16mm and with no added noise. The VSV and LDM dosimetry methods were evaluated by characterizing their respective kernels and analyzing their absorbed dose estimates calculated on theoretical phantoms. The ground truth for these estimations was calculated usingreDoseMC. The decay spectra obtained through reDoseMC showed less than a 1% difference when compared to previously published experimental data for energies below 1.9MeV in the case of and less than 1% for energies below 1.5MeV for . Additionally, the validation kernels for VSV exhibited results similar to those found in published Monte Carlo codes, with source dose depositions having less than a 3% error margin. Resolution thresholds ( ), defined as resolutions that resulted in similar dose estimates between the LDM and VSV methods, were observed for . They were 1.5mm for bone, 2.5 mm for soft tissue and liver, and 8.5 mm for lungs. For , the accuracy of absorbed dose deposition was found to be dependent on the contributions of absorbed dose from photons. Volume errors due to variations in voxel size impacted the final dose estimates. Larger target volumes yielded more accurate mean doses than smaller volumes. For both radioisotopes, the radial dose profiles for the VSV and LDM approximated but never matched the referencestandard. reDoseMC was developed and validated for radiopharmaceutical dosimetry. The accuracy of voxel-based dosimetry was found to vary widely with changes in image resolutions, voxel sizes, chosen target volumes, and tissue material; hence, the standardization of dosimetry protocols was found to be of great importance for comparable dosimetryanalysis.

Dosimetry studies for an optimized reloading of a panoramic Cobalt-60 source using FLUKA Monte Carlo code

The Cobalt-60 source used at the Tunisian Gamma Irradiation Facility is nearing the end of its useful life. Due to the current low dose rates, the irradiation process takes a long time. To address exposure time and cost issues, source racks need to be reloaded to improve the quality of the exposure process. This work aims to find the optimal new configuration for future reloading of the Cobalt-60 source rack and to evaluate the dose rate distribution in the vicinity of the Irradiator and the Dose Uniformity Ratio (DUR) in a dummy product. When keeping old pencils, the optimal configuration will need to be adjusted to accommodate the new Cobalt-60 pencils. The replenishment method developed using FLUKA-based Monte Carlo code involved the addition of six new pencils: three in the upper top (outer) holder of the source rack and three in the lower bottom (inner) source holder. The spatial distribution of dose rates in the vicinity of the source rack before and after reloading was evaluated. For the radiation processing, the maximum and minimum absorbed dose in the dummy product were estimated to calculate the DUR value. The DUR value for a single-use medical product is about 2.5. The results obtained before and after reloading show that the spatial distribution of the dose rate is uniform for the proposed configurations, and has a value of approximately 0.4 kGy/h after reloading at the conveyor belt location.

Simulating Fuel Assembly Bowing: A Neutron-Diffusion Method Based on Arbitrary Quadrilateral Node and Conformal Mapping Technique

Fuel assembly bowing, widely observed in a pressurized water reactor (PWR), often results in an asymmetrical power distribution. This paper proposes a neutron-diffusion method that integrates the arbitrary quadrilateral node with the conformal mapping technique to characterize the impact of fuel assembly bowing on power distribution. The proposed method involves a nonlinear iteration process to solve the neutron-diffusion equation. The global coarse-mesh finite difference equation is established on the arbitrary quadrilateral nodes, which are redivided in response to fuel assembly bowing. The local two-node nodal expansion method equation is established on the rectangular nodes, which are mapped from the original arbitrary quadrilateral nodes using the conformal mapping technique. The proposed method has improved our self-developed core code, named SPARK, for PWRs. To verify this novel method, two distinct types of fuel assembly bowing are modeled based on the mini core. The reference results for these models were obtained using the Monte Carlo code NECP-MCX. The numerical results suggest a robust agreement between the biases of keff and power distributions and their corresponding reference results.

Monte Carlo simulation for photon energy response from the RADOS dosemeter.

Ionizing radiation interaction models are commonly included in Monte Carlo codes. However, as there are different models and different output quantities available, it is important to understand the physical phenomena used. So, the aim of this study is to analyze the photon interaction model called Simple Physical Treatment of the Monte Carlo code MCNPX to estimate the energy dependence of photons from the RADOS dosemeter and to validate this calculation method by comparing it with experimental results found in the literature. The energy deposition in the MTS-N detector and the air kerma were obtained accompanied by their uncertainties, which varied between 1.5% and 3.0%. The RADOS dosemeter simulation and the calculation methodology applied in this study have been validated. Therefore, when using the Simple Physical Treatment model, the effective dose can be corrected by a correction factor calculated from the simulated results.

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.

Gas physisorption impact on prolate dust in free-molecule flows: A static study

Gas–solid coupling systems operating at low pressure or the micro/nanoscale generally exist in nature and industrial manufacture. Although the gas-scattering model has been widely used to study this problem on the dust surface, the consideration of gas physisorption was often neglected in previous applications of gas–surface scattering models. Therefore, this study aims to investigate the distribution of gas physisorption on the dust surface and assess its impact on the static force experienced by nonspherical dust in free-molecule flows. In this study, the prolate dust spinning around its minor axis is considered and the in-house direct simulation Monte Carlo code is used. Results show that gas physisorption on prolate dust is influenced by changes in gas number densities, Mach number, and dust shape. Furthermore, the gas physisorption enhances the gas–dust coupling for dust with a smooth surface at low gas pressure, attributed to the increasing ratio of Maxwell diffuse scattering of gas molecules on the gas-adsorbed part of the surface. Hence, gas physisorption was suggested as a potential factor for gas–dust coupling at low gas pressure.

Conceptual design and simulated performance of the Barrel part of gamma array for the experimental terminal at HFRS of HIAF

The Barrel of γ array, consisting of 640 CsI(Tl) crystals and covering a polar angle range from 36.4° to 135.6°, has been designed for the experimental terminals at High energy FRagment Separator of High-Intensity heavy-ion Accelerator Facility in Huizhou. Together with the existing Endcap, consisting of 1024 CsI(Tl) crystals and covering a polar angle from 15.6° to 36.4°, at the External Target Facility in Lanzhou to form a new γ array, it is expected to meet the experimental requirements for in-beam γ-ray spectroscopy technique. A Monte Carlo code based on Geant4 was implemented to study the performance of the new γ array. Taking into account the Doppler correction and the intrinsic energy resolution of the detector, the energy resolution is 5.4% (FWHM) for the Endcap and 9.8% (FWHM) for the Barrel when the energy of the emitted γ-ray is 1 MeV in the Center-of-Mass frame with the beam energy of 500 MeV/u (beam velocity β = 0.759). The full-space photopeak efficiency is greater than 40% when the energy of the emitted γ-ray is less than 5 MeV in the Center-of-Mass frame and the beam energy is less than 700 MeV/u (β = 0.821). Such performance can satisfy most of the physics requirements for in-beam γ-ray spectroscopy experiments.

Neutronic assessment and optimization of ACP-100 reactor core models to achieve unit multiplication and radial power peaking factor

This study focuses on the neutronic and burnup assessment of the PWR-based Chinese ACP-100 SMR core. Four different core models with varying fuel compositions and assembly configurations were analyzed using the Monte Carlo code SERPENT 2. Three of these models shared uniformly enriched fuel assemblies of 3 wt.%, 4 wt.%, and 4.45 wt.% 235U respectively, while the fourth model (the proposed design for the ACP-100 in this study) featured a mix of differently enriched fuel assemblies along with the inclusion of burnable absorbers. The burnup analysis of the cores spanned 1570 EFPDs, evaluating several neutronic parameters (beta effective, reactivity coefficients) as well as fuel cycle characteristics (cycle burnup, discharge burnup and cycle length). Changes in the isotopic concentration of fissile, fertile, and poison materials were observed and the assembly-wise linear power distribution, neutron flux spectra, and radial power peaking factors of the models were determined. In the mixed core, a combination of borated water, IFBAs and control rods were employed to achieve exact criticality (unit keff) and uniform power distribution (unit radial power peaking factor). Insertion of control rods into 41 of the 57 assemblies, in conjunction with IFBAs placed strategically, attained a near unit radial power peaking factor (PPF) of 1.0586. Finally, the combination of IFBAs, sol-bor of 936 ppm concentration, and CRs with rod worth of 414.33 pcm per assembly brought down the keff to 1.000, but increased the PPF to 1.117. While the achievement of uniform power distribution came at the cost of fuel burnup, the proposed mixed core successfully remained critical for over 1.5 years and reached a discharge burnup of 23.37 GWd/t for a two-batch refueling scheme.

Comparison of experimental measurements and fast Monte Carlo simulations for typical set-ups in fluoroscopically-guided interventional procedures

PyMCGPU-IR is an application based on the Monte Carlo code MCGPU-IR that automatically retrieves procedure information from X-ray systems and medical worker positions from a tracking camera system. PyMCGPU-IR calculates personal dose equivalent values, doses in organs and effective dose values for both patient and medical staff in interventional radiology procedures and displays them visually. The code’s main advantage lies in its time efficiency, enabling simulations in under two minutes with statistical uncertainties below 5% (k = 2).This study involved testing in a hospital room using an interventional X-ray system. A RANDO phantom simulated the patient, with passive dosimeters affixed to the back for measuring skin dose values. A PMMA slab phantom represented the operator, with passive and active dosimeters affixed to its front surface to measure Hp(10). The irradiation conditions were simulated with PyMCGPU-IR using voxelized geometries to represent both phantoms. Results demonstrate good agreement between PyMCGPU-IR simulations and measured patient skin dose values, with differences of up to 5% for mean skin dose and up to 14% for the peak skin dose. Concerning Hp(10) values on the operator phantom, PyMCGPU-IR calculated values fall within the uncertainty ranges of dosimeter measurements for most points. The highest Hp(10) discrepancy is 42%, which is acceptable when compared with the typical variability observed between active and passive personal dosimeters measurements in interventional radiology. The results demonstrate PyMCGPU-IR’s satisfactory performance for patient and personal dosimetry, compared to existing solutions like commercial skin dose calculation software and personal physical dosimeters.

Neutronic simulation of Traveling Wave Reactor (TWR) core in multi-cycles using Monte Carlo method

Abstract An ingenious design, Traveling Wave Reactor (TWR) has been suggested for full exploitation of uranium resources in next generation of nuclear reactors. This design involves the slow propagation of a nuclear fission wave through a long initially subcritical core and the transmutation of nuclear fuel. It has been widely proven that the initiation and propagation of fission waves are feasible prompting the Terra Power Company to conduct preliminary commercialization studies on this project. In equilibrium state, the shapes of the neutron flux, nuclide densities and power density distribution remain constant but the burning region moves in axial (or radial) directions. In this case the in situ fissile material production and consumption drives the operation. Only natural or depleted uranium is required for fresh fuel region. However, in equilibrium state, the burning region contains a spectrum of fission products as well as higher actinides whose are not easily available for initial TWR core construction. One solution is based on the use of enriched uranium in the first cycle to ignite the fission wave and then employment of the composition in subsequent cycles up to the equilibrium state. The main objective of this work is the feasibility study of forming a fission wave as well as the neutronic analysis of a Gas-Cooled Traveling Wave Reactor in various cycles up to the equilibrium one. The MCNP Monte Carlo code was utilized for the analysis of criticality and burn-up calculation. In each cycle, the axial fresh fuel loading and spent fuel discharge, were subjected to analysis. Results showed that the equilibrium state can be obtained by using special arrays of absorbers, from the third cycle where the shape of neutron flux and the power density distribution in axial direction remain unchanged.