Monte Carlo Particle Transport Research Articles

Ab initio electronic stopping power for protons in Ga0.5In0.5P/GaAs/Ge triple-junction solar cells for space applications.

Motivated by the radiation damage of solar panels in space, firstly, the results of Monte Carlo particle transport simulations are presented for proton impact on triple-junction Ga0.5In0.5P/GaAs/Ge solar cells, showing the proton projectile penetration in the cells as a function of energy. It is followed by a systematic ab initio investigation of the electronic stopping power (ESP) for protons in different layers of the cell at the relevant velocities via real-time time-dependent density functional theory calculations. The ESP is found to depend significantly on different channelling conditions, which should affect the low-velocity damage predictions, and which are understood in terms of impact parameter and electron density along the path. Additionally, we explore the effect of the interface between the layers of the multilayer structure on the energy loss of a proton, along with the effect of strain in the lattice-matched solar cell. Both effects are found to be small compared with the main bulk effect. The interface energy loss has been found to increase with decreasing proton velocity, and in one case, there is an effective interface energy gain.

Modernization of the DCHAIN-PHITS activation code with new features and updated data libraries

Induced radioactivity is critical in a variety of fields including maintenance and decommissioning of accelerators and reactors, nuclear waste management, and radionuclide production for medical and industrial applications. Serving this purpose, DCHAIN-PHITS is the time-dependent radionuclide production, buildup, burnup, and decay code which is coupled to and distributed with the PHITS general purpose Monte Carlo particle transport code. While the core of the DCHAIN code was first written over forty years ago, the DCHAIN-PHITS incarnation of the code represents its most substantial update in nearly twenty years. This work provides an overview of the function of DCHAIN-PHITS and the new additions to the code. These additions include updated decay and cross section nuclear data libraries, statistical uncertainty propagation, support for tetrahedral and three-dimensional grid mesh geometries, reaction tracking capabilities, and general performance improvements.

Measurement of thick target neutron yield at 180°for a mercury target induced by 3-GeV protons

A thick target neutron yield (TTNY) for a mercury target at an angle of 180°from the incident beam direction was measured with the time-of-flight method using a 3-GeV proton beam at the Japan Proton Accelerator Research Complex (J-PARC). Comparison of the experimental results with a Monte Carlo particle transport simulation by the Particle and Heavy Ion Transport code System (PHITS) shows that the default spallation model of PHITS, the Liège intranuclear cascade model version 4.6 coupled with the generalized evaporation model (INCL4.6/GEM) reproduces the measured TTNY fairly well, whereas apparent overestimation is observed for the former default model, the Bertini intranuclear cascade model coupled with GEM (Bertini/GEM). These trends are consistent with the experimental results of neutron-induced reaction rates obtained using indium and niobium activation foils.

Integral molten salt reactor temperature sensitivities using Serpent target motion sampling

The Molten Salt Reactor (MSR) concept has attracted several start-up companies due to the potential for reducing waste generation, realizing passive safety features, and seizing on the opportunity for cost effective economics. Terrestrial Energy, Inc (TEI), for example, has designed and is seeking regulator licensing for an Integral MSR (IMSR) design that contains all primary system components within a single reactor vessel containment. This study uses sensitivity analysis capabilities unique to Monte-Carlo particle transport code Serpent to determine and evaluate fuel salt Doppler and density temperature coefficients of reactivity for an IMSR over a range of normal or casualty temperature ranges. This study contrasts the temperature coefficients of three candidate molten salt fuels (FLiBe, FLiNaK, and NaF-RbF). The unique advantages and access to insightful neutron physics using Serpent code sensitivity methods are highlighted. An evaluation of the total reactor temperature coefficient and in particular the fuel salt temperature coefficients will show that an IMSR design can be self-limiting from a design basis accident perspective.

Observation of morphological abnormalities in silkworm pupae after feeding 137CsCl-supplemented diet to evaluate the effects of low dose-rate exposure

Since the Fukushima Dai-ichi Nuclear Power Plant (FDNPP) accident, morphological abnormalities in lepidopteran insects, such as shrinkage and/or aberration of wings, have been reported. Butterflies experimentally exposed to radiocesium also show such abnormalities. However, because of a lack of data on absorbed dose and dose–effect relationship, it is unclear whether these abnormalities are caused directly by radiation. We conducted a low dose-rate exposure experiment in silkworms reared from egg to fully developed larvae on a 137CsCl-supplemented artificial diet and estimated the absorbed dose to evaluate morphological abnormalities in pupal wings. We used 137CsCl at 1.3 × 103 Bq/g fresh weight to simulate 137Cs contamination around the FDNPP. Absorbed doses were estimated using a glass rod dosimeter and Monte Carlo particle transport simulation code PHITS. Average external absorbed doses were approximately 0.24 (on diet) and 0.016 mGy/day (near diet); the average internal absorbed dose was approximately 0.82 mGy/day. Pupal wing structure is sensitive to radiation exposure. However, no significant differences were observed in the wing-to-whole body ratio of pupae between the 137CsCl-exposure and control groups. These results suggest that silkworms are insensitive to low dose-rate exposure due to chronic ingestion of high 137Cs at a high concentration.

Computation of Fission Product Condensation on Chainlike Aerosols and Agglomerates—II: Role of Energy and Mass Dependence of Molecules

Although aerosols in some postaccident nuclear environments can be nonspherical, chainlike, or agglomerates, there have been limited investigations of the rate processes (such as coagulation, evaporation, condensation, and deposition) involving such particles. In a previous investigation, the understandings of condensation and evaporation on such particles were expanded through use of a one-speed approximation for modeling vapor (or fission product) molecular transport, and the present paper extends that work to energy- and mass-dependent transport of vapor molecules within the context of the linear Boltzmann equation via the Monte Carlo particle transport method for rigid sphere molecules. The results are benchmarked against available numerical results and experimental data for a single sphere, and it is found again that the normalized condensation rate has only a weak dependence on the molecular mass ratio (vapor to background) and that the one-speed approximation is quite good. Results are reported for a range of chainlike and agglomerate aerosols.

Impact of uranium oxide (UO2) fuel with molybdenum (Mo) inserts on pressurized water reactor performance and safety

This work investigates nuclear reactor performance and safety characteristics of UO2 with high thermal conductivity Mo insert structures by using multiphysics modeling techniques. The purpose of this study is to use scoping analyses to quantify the impact of using Mo inserts from neutronic and heat transfer standpoints. Attention is given to reactor performance parameters, such as cycle length, maximum fuel temperature, temperature gradients in the fuel, and stored energy in the fuel. The finite-element code BISON and the Monte Carlo particle transport code Serpent were used to perform sensitivity analyses on the Mo insert geometry to optimize the insert design and inform larger scale modeling that required the homogenization of the UO2 and Mo. Although BISON is often used as a fuel performance analysis tool, it is used in this context for heat transfer analysis only. Fuel performance optimization is outside the scope of the current study, but would be important for future work focused on this concept. The results showed that the insert had little impact on neutronic performance and that homogenizing the UO2 and Mo was acceptable for reactor physics calculations. Reactivity temperature coefficients calculated using homogeneous UO2-Mo were shown to be relatively similar to UO2, but higher Mo content and 235U enrichment can reduce the worth of soluble boron and control rods. The effect of insert geometry on heat transfer was much greater, and an approximately 15–20% difference in maximum fuel temperature was predicted between the best and worst performing heat transfer geometries. Thermal conductivity calibration based on the finite element analysis results was performed to improve the accuracy of temperature predictions in reactor analysis models that homogenized the UO2-Mo fuel. Compared with UO2 in a pressurized water reactor (PWR), the optimized UO2-Mo design increased the margin to fuel melt by 13–32% across the fuel cycle, but it requires the 235U enrichment to exceed 5% to match the cycle length of conventional UO2.

Shielding assessment of the IAMI facility

The Institute for Advanced Medical Isotopes (IAMI) is TRIUMF's new facility dedicated to the research and production of medical isotopes. This work details the shielding requirements and radiological studies performed during the facility design. The initial design was based on the existing TR30-2 machine at TRIUMF, and then refined using Monte Carlo (MC) simulations. These calculations were performed with the MC particle transport and interaction code FLUKA, in several iterations, until occupancy and dosimetry requirements were met. The simulation results here documented, besides validating the shielding design, were also used to reduce materials' cost, optimize services' location as well as in the decommission study.

Modelling nuclear fuel assembly with thermal-hydraulic feedback and burnup using WIMS-PANTHER-Serpent

State-of-the-art deterministic code sequence WIMS-PANTHER and Monte Carlo particle transport code Serpent are employed in this work for computationally modelling the UK EPR nuclear reactor core’s fuel assembly with and without thermal-hydraulic feedback, the results of which targets the government energy policymakers and wider academic community.

Assessment of CASL VERA for BWR analysis and application to accident tolerant SiC/SiC channel box

Application of the Virtual Environment for Reactor Applications (VERA) to BWR analysis is assessed in this paper by comparing results to those calculated using other widely-used modeling tools, namely the U.S. Nuclear Regulatory Commission’s PARCS/PATHS and the Serpent Monte Carlo particle transport code. Additionally, VERA is used to calculate 3-D temperature and fast neutron flux distributions in silicon carbide (SiC) fiber-reinforced, SiC matrix composite (SiC/SiC) BWR channel boxes, which are being studied as an Accident Tolerant Fuel core structural material concept. The code-to-code comparisons were favorable, and the SiC/SiC channel box evaluation demonstrates the many advanced modeling features VERA provides while also highlighting the non-uniformity in fast neutron flux distributions that can play a role in potential SiC/SiC channel box deformation. Traditional BWR analysis tools do not have the calculation fidelity necessary for coupled assessment of flux and temperature gradients in a SiC/SiC channel box. VERA is a state-of-the-art modeling environment that was developed to increase the safety and economic competitiveness of nuclear power through improved modeling accuracy. While VERA has already been deployed in the nuclear industry for PWR applications, the current study is a vital initial step in the extensive development, validation, and verification that VERA must go through to be useful for BWR applications.

Random model for radiation shielding calculation of particle reinforced metal matrix composites and its application

The work aimed to calculate the radiation biological shielding performance of particle reinforced metal matrix composite (PRMMCs) using more reasonable model instead of conventional Uniform Filling Model, also attempted to provide a basis for the radiation shielding optimal design of such materials. Firstly, RSA (Random Sequential Adsorption) Model and GRM (Grid Random Model) were established based on MATLAB and Monte Carlo Particle transport program MCNP, and then advantages and disadvantages of them were compared. Later, the influences of metal matrix type, particle (B4C) content, particle shape and particle shape parameters on the biological shielding performance of materials were calculated under different energy neutrons and different thickness shield using random models. Finally, the optimal aspect ratio of regular hexahedral B4C was calculated by Genetic Algorithm combined with MATLAB and MCNP. It indicated that GRM could be applied to radiation shielding calculation of PRMMCs.

Characterization of the neutron spectra in three irradiation channels of the JSI TRIGA reactor using the GRUPINT spectrum adjustment code

Reference neutron fields in irradiation facilities are crucial for experimental research work in nuclear science and technology in general, and at the fundamental level for the measurement of neutron cross sections and validation of evaluated cross-section data. Monte Carlo particle transport techniques in conjunction with state-of-the-art nuclear data libraries are becoming the reference in the broad context of nuclear reactor analysis, and are extensively used for detailed computational characterization of irradiation facilities, including the calculation of neutron spectra. However, adjustment of the neutron spectra based on experimental data is still required.In this work we present the characterization of the neutron spectra in the three most important irradiation channels of the JSI TRIGA reactor with different spectral characteristics, as an important step in developing the experimental capabilities of the reactor for the purpose of nuclear data validation through activation dosimetry measurements. In the first part of this work we performed detailed Monte Carlo calculations of the neutron spectra, comparing the results obtained using different nuclear data libraries. In the main part of this work we employed the JSI developed GRUPINT spectrum adjustment code for the characterization of the neutron spectra, on the basis of the calculated neutron spectra and measured reaction rate ratios. A comprehensive overview of the methodology employed in the code and its abilities is given.We observe that a positive compensation of up to 10 % in the thermal part of the spectrum is needed, while no compensation is needed in the epithermal and fast parts. Absolute neutron flux levels were computed from the adjusted neutron spectra and absolute reaction rate measurements, and compared to the results of a previous analysis using only the Monte Carlo calculations. The work presented in this paper represents the basis for future experimental and analytical work, in particular extending the scope of the measured capture and threshold dosimetry nuclear reactions, to support a thorough characterization of the neutron spectra in the irradiation channels of the reactor. This effort will enable the JSI TRIGA reactor to support a broad scope of experiments for fundamental research as well as development or testing activities at the highest level.

Implementation and verification of PyNE R2S with DAG-OpenMC

The mesh-based Rigorous-Two-Step (R2S) method has been widely used in the accurate estimation of the shutdown dose rate (SDR) of fusion systems. Several mesh-based R2S code has been implemented based on Monte Carlo particle transport code MCNP5 or DAG-MCNP5 and inventory calculation code such as FISPACT-II, ALARA or ACAB. Recently, the OpenMC, a community-developed Monte Carlo particle transport code, implemented the photon particle transport capability, which enables OpenMC to be used in R2S. In this paper, PyNE R2S is extended with the support of DAG-OpenMC, i.e., OpenMC with DAGMC. Modifications that allow PyNE to read the flux result in the state point file of OpenMC has been implemented to PyNE R2S workflow. A custom source routine of OpenMC that reads the photon source file from PyNE R2S and sampling photon particle in OpenMC photon transport has been implemented in OpenMC. With these modifications, PyNE R2S is now capable of performing R2S calculations with both DAG-MCNP5 and DAG-OpenMC. The ITER FNG dose rate benchmark problem has been used to validate the code. The FNG neutron source has been modified for neutron transport with DAG-OpenMC. The shutdown dose rate of 19 cooling times has been calculated and compared with the experimental data and the computational results of other R2S codes. The results of PyNE R2S with DAG-OpenMC show satisfactory agreement with the experimental data and other computational results. Therefore, we consider PyNE R2S with DAG-OpenMC is a reliable code to calculate the SDR of fusion systems.

A general-purpose Monte Carlo particle transport code based on inverse transform sampling for radiotherapy dose calculation

The Monte Carlo (MC) method is widely used to solve various problems in radiotherapy. There has been an impetus to accelerate MC simulation on GPUs whereas thread divergence remains a major issue for MC codes based on acceptance-rejection sampling. Inverse transform sampling has the potential to eliminate thread divergence but it is only implemented for photon transport. Here, we report a MC package Particle Transport in Media (PTM) to demonstrate the implementation of coupled photon-electron transport simulation using inverse transform sampling. Rayleigh scattering, Compton scattering, photo-electric effect and pair production are considered in an analogous manner for photon transport. Electron transport is simulated in a class II condensed history scheme, i.e., catastrophic inelastic scattering and Bremsstrahlung events are simulated explicitly while subthreshold interactions are subject to grouping. A random-hinge electron step correction algorithm and a modified PRESTA boundary crossing algorithm are employed to improve simulation accuracy. Benchmark studies against both EGSnrc simulations and experimental measurements are performed for various beams, phantoms and geometries. Gamma indices of the dose distributions are better than 99.6% for all the tested scenarios under the 2%/2 mm criteria. These results demonstrate the successful implementation of inverse transform sampling in coupled photon-electron transport simulation.

Effect of Al2O3np on the Properties and Microstructure of B4Cp/Al Composites

Aluminum-matrix boron carbide (B4Cp/Al) is a kind of neutron absorbing material widely used in nuclear spent fuel storage. In order to improve the tensile property of B4Cp/Al composites, a new type of nano-Al2O3 particle (Al2O3np) reinforced B4Cp/Al + Al2O3np composites were prepared by powder metallurgy method. The Monte Carlo particle transport program (MCNP) was used to determine the influence of Al2O3np on the thermal neutron absorptivity of composites. The universal material testing machine and scanning electron microscope (SEM) were used to study the mechanical properties, microstructure and fracture morphology of B4Cp/Al composites. The results indicated that the neutron absorption properties of B4Cp/Al composites were not affected by the addition of nano-Al2O3 particles in the range of 1 wt%–15 wt%. The addition of Al2O3np can obviously reduce the grain size of B4Cp/Al matrix metals thus improve the tensile strength of the composites. The addition threshold of Al2O3np is about 2.5 wt%. Both B4Cp and Al2O3np change the fracture characteristics of the composites from toughness to brittleness, and the latter is more important.

Operation and shutdown dose rate analysis of CFETR ECRH system

Electron Cyclotron Resonance Heating (ECRH) is one of auxiliary heating systems to be installed in CFETR. For heating purpose of ECRH, a penetration space has been designed to transmit the wave to plasma. This will significantly increase the neutrons streaming through the wave channel and result in heavy irradiation of ECRH port structures. For nuclear safety analysis, the irradiation dose during operation of CFETR should be limited according to the regulation of radiological zone. In case of failure, the ECRH is planned to be maintained by workers after shutdown and waiting time of one month until the dose rate decreases to low level of 100 μSv/h. The shielding capabilities of ECRH port has been evaluated in view of dose effect both during operation and after shutdown of CFETR. For that purpose, a 3-D neutronics model has been built to launch the nuclear analysis. The operation dose rate was directly transferred from the neutron and photon flux by using of ICRP-74 flux to dose conversion factors. The shutdown dose rate was obtained with NASCA code system by coupling the Monte Carlo particle transport and FISPACT activation calculation. The obtained operation and shutdown dose data can be provided for the shielding optimization of ECRH wave system.

Energy-reduced beta radiation fields from 90Sr/90Y for the BSS 2

For decades, the irradiation facility for beta radiation, the Beta Secondary Standard BSS 2 developed at PTB, has been in worldwide use to irradiate dosemeters, detectors and other devices with calibrated beta sources. In this work two energy-reduced radiation fields based on the 90Sr/90Y source are described. Their electron and photon particle spectra 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 EGSnrc/BEAMnrc and are provided in the supplementary data.

Development of a predictive capability of short-pulse laser-driven broadband x-ray radiography

High intensity, short-pulse laser interaction with a solid metal target produces broadband hard x-rays potentially for various applications of x-ray radiography. Here experimental benchmarking of numerical modelling for short-pulse laser-driven broadband x-ray radiography is presented. Angular dependent x-ray spectra are first calculated with a hybrid particle-in-cell code, Large Scale Plasma (LSP), using fast electron parameters inferred from an analysis of measured bremsstrahlung signals. Subsequently, a calculated x-ray spectrum in the direction of radiography is used in photon transport calculations using a Monte Carlo code, Particle and Heavy Ion Transport code System (PHITS), to simulate a radiographic image including a modelled 3D test object, an x-ray attenuation filter and an image plate detector. Simulated radiographic images are compared with measurements obtained in an experiment using a 50-TW Leopard short-pulse laser at the University of Nevada Reno. Results show that simulations reproduce the experimental images well for three different attenuation filters (plastic, aluminium, and brass), while 1D transmission profiles for the plastic and aluminium filters are quantitatively in good agreement. The modelling approach established in this work could be used as a predictive tool to simulate radiographic images of complex 3D solid objects at any arbitrary angular position or to optimize experimental components such as the source spectrum, x-ray attenuation filters and a detector type depending on a radiographic object without carrying out radiographic experiments.

Design, shielding mechanism and tensile property of a novel (B4C+6061Al)/Cf/6061Al laminar neutron-shielding composite

In this study, a novel (20 wt% B4C+6061Al)/Cf/6061Al neutron-shielding layer composite was successfully designed and fabricated by the Monte Carlo particle transport programme (MCNP) and spark plasma sintering, respectively. The MCNP simulation and thermal-neutron-shielding test results show that the shielding performance of the (20 wt% B4C+6061Al)/Cf/6061Al composite improved by 6% compared with the 20 wt% B4C/6061Al composite and is approximately equal to that of the 30 wt% B4C/6061Al composite. The B4C particles were homogeneously distributed in the B4C/6061Al neutron absorption layer, and the neutron absorption layer, reflection layer and moderation layer were tightly connected. The neutron reflection layer Cf increases the reflection times of the thermal neutron, which results in the decrease in neutron energy and promotes the neutron absorption. Thermal neutrons are mainly absorbed by the 10B (n, α)7Li transmutation reaction in the neutron absorption layer. The high modulus and axial strength of Cf and the good plasticity of the 6061 Al matrix contribute to the good mechanical properties of the (B4C+6061Al)/Cf/6061Al composite. The tensile strength and strain of the (B4C+6061Al)/Cf/6061Al composite are 245 MPa and 14%, respectively.

Simulating Charge Deposition by Cosmic Rays Inside Astronomical Imaging Detectors

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