Monte Carlo Particle Transport Code Research Articles

Comprehensive approach to determination of space proton-induced displacement defects in silica optical fiber

Silica-based optical fibers presents a variety of applications used in radiation environments such as space, fusion facilities, accelerators and nuclear power plants. The radiation-induced displacement damage in optical fibers resulting in point defects may lead to attenuation signals that is a major concern for these applications. The present study proposes a computational approach to the calculation of the proton-induced displacement damage in vitreous silica. Therefore, Geant4 as a Monte Carlo particle transport code has been used to obtain the knock-on atom distributions caused by the interaction of space trapped proton with vitreous silica during an ISS mission. Moreover, molecular dynamics simulations using ReaxFF potential have been performed to produce the initial vitreous silica structure to evaluate the displacement damage cascades by LAMMPS package. The results show that ReaxFF has an appropriate potential to produce and evaluate the vitreous silica structure that provides better agreement with experimental data at both short-range and medium-range order. Furthermore, ODC(Si3) and NBOHC(O1) are dominant defect species created in the vitreous silica after trapped proton irradiation, where the total number of defects have increased on average by 94 for each keV increasing in PKA energy approximately.

An overview of the shielding optimization studies for the TRIUMF-ARIEL facility

The Advanced Rare IsotopE Laboratory (ARIEL) facility, dedicated to Radioactive Ion Beam (RIB) production via the Isotope Separation On-Line (ISOL) technique, is currently under construction at TRIUMF. The facility is to host dual target stations that will operate simultaneously, including a 500 MeV proton beam from the main TRIUMF cyclotron and a 50 MeV electron beam from an Electron Linear Accelerator (e-linac). The proton target station will operate a symbiotic medical target in series with the ISOL target.The ARIEL facility features a compact shielding envelope and as a result its optimization relied extensively on simulations generated using the Monte Carlo particle transport and interaction code FLUKA. Dose rate maps were produced of prompt radiation associated with beam losses and expected normal operations. Maps of transient fields were also generated considering radiation sources requiring transport within the facility which include radioactive waste generated in the course of normal operations. Activation studies were also conducted to estimate the amount of residual radioactivity contained in shielding materials at the time of decommissioning. Simulation results are primarily used to inform the ongoing design and construction of ARIEL; at present, the entirety of the proposed facility shielding has been validated by FLUKA simulation data. The specification of shielding geometry and the selection of materials relied extensively on engineering support through verification of CAD models and material sampling to optimize solutions for design simplicity and economy. Normal operation and failure scenarios were both evaluated, with optimized shielding materials and thicknesses selected considering the nature of the radiation fields present and their impact on room occupancy. Finally, a decommissioning plan has been defined based on the anticipated ARIEL lifetime, the characterization of structural activation, and a general disposal strategy.While ARIEL is still under construction and significant progress has been made, the ongoing radiation modelling has enabled validation of the current shielding design for nominal operation. Moreover, the systematic studies performed were paramount in defining the optimal strategies to adopt from target storage to that for decommissioning of the facility.

Energy deposition calculation by Monte Carlo simulation in irradiation of electric cables by electron beam

The reason for crosslinking the insulation polymer in wire and cable by radiation is to improve their performance at an elevated temperature, the chemical resistance, toughness and abrasion resistance in use. The quality of products made by radiation processing depends not only on the absorbed dose but also on homogenuity of dose distribution especially in circular objects like cables which have different density in conductor and insulation part, respectively.It is important to optimize crosslinking processes for reducing the operation cost while sustaining suitable quality of the final product. The computer simulation method becomes a very useful tool for optimizing process, providing essential information in a short time and reducing cost comparing with the conventional method based on experimental dosimetry. The application of several Monte Carlo codes for electron and photon transport and the advances in computing systems, make possible to accomplish dosimetry tasks by numerical simulation. With increased computing speed and decreasing hardware cost, it is expected that Monte Carlo based electron dose calculation algorithms will become available for routine electron beam dosimetry. We have developed the EBSCANmc (simulator program) in order to simulate and study the crosslinking of electric cables. EBSCANmc employs the MCNP (Monte Carlo N-Particle) 6.2 code, which is a general purpose Monte Carlo particle transport code. The computer simulation of the dose distributions in electric cables was performed for various process parameters, namely the level of electron beam energy, irradiation positions, the density of cable insulator, and backscattering. The results of computer simulation are compared with experimental data using the CTA film dosimeter.

The Calculation of Electric Field Caused by Distributed Charge

The potential and electric field responses in cylinder cavity with ceramics slice caused by distributed charge in X-ray environment were simulated with finite difference method. The numerical results indicate that the distributed charge in ceramics’ middle section is negative or positive, and positive of both sides. For the photon energies from 30 keV to 1 MeV with same fluence, distributed charge and electric field were decreased at first and then increased. Electrode has great effect to the potential and electric field responses. For different plating thickness, there exists minimum to the electric field amplitude, and for the plating, there exist optimal diameter and thickness. Distributed charge is calculated using Monte-Carlo particle transport code.

Geometrical optimisation of a segmented HPGe detector for spectroscopic gamma emission tomography—A simulation study

Segmented coaxial HPGe (High Purity Germanium) detectors have recently been shown to be feasible for Gamma Emission Tomography (GET). This type of detector allows for a combination of high efficiency and high energy resolution in gamma spectrometry of irradiated nuclear fuel. The ultimate aim of developing segmented HPGe for GET measurements is to achieve a high spatial resolution to facilitate imaging of rod-internal features and interrogation of smaller irradiated fuel samples.In this work, we present the optimisation of a segmented HPGe detector through a simulation study using the Monte Carlo particle transport code MCNP. Constraints to each dimension of the detector were identified, from manufacturing limitations and requirements arising from the use of a finite-sized collimator slit. In particular, a relationship between the minimum inner radius of the coaxial detector and the segments azimuthal dimension was derived based on the identified constraints. Segment azimuthal and radial dimensions have been varied (based on the derived relationship between the azimuthal and radial dimension) and the full energy efficiency and misidentification rate were evaluated to obtain the optimal dimensions. The optimal ranges of the segment dimensions were determined.

Individual dosimetry system for targeted alpha therapy based on PHITS coupled with microdosimetric kinetic model

BackgroundAn individual dosimetry system is essential for the evaluation of precise doses in nuclear medicine. The purpose of this study was to develop a system for calculating not only absorbed doses but also EQDX(α/β) from the PET-CT images of patients for targeted alpha therapy (TAT), considering the dose dependence of the relative biological effectiveness, the dose-rate effect, and the dose heterogeneity.MethodsA general-purpose Monte Carlo particle transport code PHITS was employed as the dose calculation engine in the system, while the microdosimetric kinetic model was used for converting the absorbed dose to EQDX(α/β). PHITS input files for describing the geometry and source distribution of a patient are automatically created from PET-CT images, using newly developed modules of the radiotherapy package based on PHITS (RT-PHITS). We examined the performance of the system by calculating several organ doses using the PET-CT images of four healthy volunteers after injecting 18F-NKO-035.ResultsThe deposition energy map obtained from our system seems to be a blurred image of the corresponding PET data because annihilation γ-rays deposit their energies rather far from the source location. The calculated organ doses agree with the corresponding data obtained from OLINDA 2.0 within 20%, indicating the reliability of our developed system. Test calculations by replacing the labeled radionuclide from 18F to 211At suggest that large dose heterogeneity in a target volume is expected in TAT, resulting in a significant decrease of EQDX(α/β) for higher-activity injection.ConclusionsAs an extension of RT-PHITS, an individual dosimetry system for nuclear medicine was developed based on PHITS coupled with the microdosimetric kinetic model. It enables us to predict the therapeutic and side effects of TAT based on the clinical data largely available from conventional external radiotherapy.

VALIDATION OF SERPENT FOR FUSION NEUTRONICS ANALYSIS AT JET

Fusion neutronics analysis before and after experiments at JET is traditionally performed using Monte Carlo particle transport code Monte Carlo N-Particle. For redundancy and diversity reasons there is a need of an additional Monte Carlo code, such as Serpent 2, capable of fusion neutronics analysis. In order to validate the Serpent code for fusion applications a detailed model of JET was used. Neutron fluxes and reaction rates were calculated and compared for positions outside the tokamak vacuum vessel, in the vacuum vessel above the plasma and next to a limiter inside the vacuum vessel. For all detector positions with DD and DT neutron sources the difference between neutron fluxes calculated with both Monte Carlo codes were within 2σ statistical uncertainty and for most positions (more than 90 % of all studied positions) even within 1σ uncertainty. Fusion neutronics analysis in the JET tokamak with Serpent took on average 10 % longer but this can be improved by changing the threshold value for determination of the transport method used. With the work presented in this paper the Serpent Monte Carlo code was validated to be a viable alternative to MCNP for fusion neutronics analysis for the JET tokamak.

Estimating gamma and neutron radiation fluxes around BWR quivers for nuclear safeguards verification purposes

Non-destructive assay (NDA) methods are at the core of nuclear safeguards verification of spent nuclear fuel (SNF). In Sweden, the spent nuclear fuel from all the reactor sites is moved to the Swedish central interim storage facility for spent nuclear fuel (for which the Swedish acronym is Clab). A new facility, Clink, is planned at the site where the SNF will undergo a safeguards verification prior to encapsulation for long-term storage. The fuel to be encapsulated includes both regular fuel assemblies as well as “non-regular” fuel assemblies including fuel objects called quivers, which are specially designed containers to house damaged or failed and leaking spent fuel rods in a way to isolate the rods from the environment and prevent contamination. The quiver concept was recently introduced in the Swedish nuclear market by Westinghouse Electric Sweden AB and it has led to some unique challenges from a safeguards verification standpoint which stem from their construction. Their overall stainless steel build, while providing robustness to the structure, also greatly diminishes the possibility of detecting gamma or neutron radiation using traditional safeguards measurement devices. The current investigation looks into the practicalities of safeguards verification of boiling water reactor (BWR) quiver objects in the spent fuel pool from above, and also assesses the possibility of their verification from the side using the widely used Fork detector. The Fork instrument has been routinely employed by both operators and inspectors around the world to verify spent fuel for routine safeguards inspections. In the present work, we model the BWR quiver and the Fork instrument in the Monte Carlo particle transport code, Serpent2 to estimate the radiation flux around the quiver objects. We have shown that the gamma and neutron radiation from the BWR quiver were heavily attenuated by the stainless steel lid and could not be relied on to make a safeguards verification from above. Furthermore, it was established that while gamma radiation from the quiver remains measurable on the sides of the quiver by the Fork instrument, the neutron counts were low compared to a typical BWR fuel assembly of similar fuel content albeit within the limits of detectability of the Fork.

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.

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.

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.

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.

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.

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.

Simulating Charge Deposition by Cosmic Rays Inside Astronomical Imaging Detectors

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

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

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

Photon transport physics in Serpent 2 Monte Carlo code

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