Monte Carlo Particle Transport Code Research Articles

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.

ParaStell: parametric modeling and neutronics support for stellarator fusion power plants

The three-dimensional variation inherent to stellarator geometries and fusion sources motivates three-dimensional modeling to obtain accurate results from computational modeling in support of design and analysis of first wall, blanket, and shield (FWBS) systems. Manually constructing stellarator fusion power plant geometries in computer-aided design (CAD) and defining the corresponding fusion source can be cumbersome and challenging. The open-source parametric modeling toolset ParaStell has been developed to automate construction of such geometries in low-fidelity. Low-fidelity modeling is useful during the conceptual phase of engineering design as a means of rapidly exploring the design space of a given device. The modeling capability of ParaStell includes in-vessel components and magnets, for any given stellarator configuration, using a parametric definition and plasma equilibrium data. Furthermore, the toolset automates the generation of detailed, tetrahedral neutron source definitions and DAGMC geometries for use in neutronics modeling. ParaStell assists rapid design iteration, parametric study, and design optimization of stellarator fusion cores. As a demonstration of the design iteration capability, the effect of the three-dimensional parameter space on tritium breeding and magnet shielding is investigated, using the WISTELL-D configuration as a design basis. Blanket and shield thicknesses are varied in three dimensions, using the space available between the plasma edge and magnet coils as a constraint. The corresponding effects on tritium breeding ratio and magnet heating are tallied using the open-source Monte Carlo particle transport code OpenMC. The inclusion of additional and higher-fidelity modeling capabilities is planned for ParaStell’s future, as well as its implementation in machine-driven optimization.

Modeling gamma detectors in OpenMC: Validation of a newly implemented pulse-height tally

Gamma spectroscopy measurements can be simulated using a pulse-height tally functionality of Monte Carlo particle transport codes. Such a functionality must account for the complete simulation history of a particle’s energy deposited in a particular volume. OpenMC, an open-source application for neutron and photon particle transport which is widely used in the nuclear engineering community, previously lacked adequate simulation capabilities for gamma spectroscopy. In this paper, we outline the implementation of the photon pulse-height tally for OpenMC. Additionally, we validate the new function by comparing results to analytical calculations, other software, and experimental data.

Development and validation of fully open-source R2S shutdown dose rate capabilities in OpenMC *

We present the first fully open-source capabilities for shutdown dose rate (SDR) calculations of fusion energy facilities based on the Rigorous 2-Step (R2S) methodology. These capabilities have been implemented in the OpenMC Monte Carlo particle transport code, building on its existing capabilities while also leveraging new features that have been added to the code to support SDR calculations, such as decay photon source generation. Each of the individual physics components in the R2S workflow—neutron transport, activation, decay photon source generation, and photon transport—have been verified through code-to-code comparisons with MCNP6.2 and FISPACT-II 4.0. These comparisons generally demonstrate excellent agreement between codes for each of the physics components. The full cell-based R2S workflow was validated by performing a simulation of the first experimental campaign from the Frascati Neutron Generator (FNG) ITER dose rate benchmark problem from the Shielding INtegral Benchmark Archive and Database (SINBAD). For short cooling times, the dose calculated by OpenMC agrees with the experimental measurements within the stated experimental uncertainties. For longer cooling times, an overprediction of the shutdown dose was observed relative to experiment, which is consistent with previous studies in the literature. Altogether, these features constitute a combination of capabilities in a single, open-source codebase to provide the fusion community with a readily-accessible option for SDR calculations and a platform for rapidly analyzing the performance of fusion technology.

Combining Wave and Particle Effects in the Simulation of X-ray Phase Contrast—A Review

X-ray phase-contrast imaging (XPCI) is a family of imaging techniques that makes contrast visible due to phase shifts in the sample. Phase-sensitive techniques can potentially be several orders of magnitude more sensitive than attenuation-based techniques, finding applications in a wide range of fields, from biomedicine to materials science. The accurate simulation of XPCI allows for the planning of imaging experiments, potentially reducing the need for costly synchrotron beam access to find suitable imaging parameters. It can also provide training data for recently proposed machine learning-based phase retrieval algorithms. The simulation of XPCI has classically been carried out using wave optics or ray optics approaches. However, these approaches have not been capable of simulating all the artifacts present in experimental images. The increased interest in dark-field imaging has also prompted the inclusion of scattering in XPCI simulation codes. Scattering is classically simulated using Monte Carlo particle transport codes. The combination of the two perspectives has proven not to be straightforward, and several methods have been proposed. We review the available literature on the simulation of XPCI with attention given to particular methods, including the scattering component, and discuss the possible future directions for the simulation of both wave and particle effects in XPCI.

Analysis of fluctuations in ex-core neutron detector signal in Krško NPP during an earthquake

During an earthquake on December 29th 2020, the Krško NPP automatically shutdown due to the trigger of the negative neutron flux rate signal on the power range nuclear instrumentation. From the time course of the detector signal, it can be concluded that the fluctuation in the detector signal may have been caused by the mechanical movement of the ex-core neutron detectors or the pressure vessel components rather than the actual change in reactor power. The objective of the analysis was to evaluate the sensitivity of the neutron flux at the ex-core detector position, if the detector is moved in the radial or axial direction. In addition, the effect of the core barrel movement and core inside the baffle movement in the radial direction were analysed. The analysis is complemented by the calculation of the thermal and total neutron flux gradient in radial, axial and azimuthal directions. The Monte Carlo particle transport code MCNP was used to study the changes in the response of the ex-core detector for the above-mentioned scenarios. Power and intermediate-range detectors were analysed separately, because they are designed differently, positioned at different locations, and have different response characteristics. It was found that the movement of the power range ex-core detector has a negligible effect on the value of the thermal neutron flux in the active part of the detector. However, the radial movement of the intermediate-range detector by 5 cm results in 7%–8% change in the thermal neutron flux in the active part of the intermediate-range detector. The analysis continued with an evaluation of the effects of moving the entire core barrel on the ex-core detector response. It was estimated that the 2 mm core barrel radial oscillation results in ∼4% deviation in the power and intermediate-range detector signal. The movement of the reactor core inside baffle can contribute ∼6% deviation in the ex-core neutron detector signal. The analysis showed that the mechanical movement of ex-core neutron detectors cannot explain the fluctuations in the ex-core detector signal. However, combined core barrel and reactor core inside baffle oscillations could be a probable reason for the observed fluctuations in the ex-core detector signal during an earthquake.

Radiation attenuation and elemental composition of locally available ceramic tiles as potential radiation shielding materials for diagnostic X-ray rooms

Ceramic materials are being explored as alternatives to toxic lead sheets for radiation shielding due to their favorable properties like durability, thermal stability, and aesthetic appeal. However, crafting effective ceramics for radiation shielding entails complex processes, raising production costs. To investigate local viability, this study evaluated Malaysian ceramic tiles for shielding in diagnostic X-ray rooms. Different ceramics in terms of density and thickness were selected from local manufacturers. Energy Dispersive X-ray Fluorescence (EDXRF) and X-ray Fluorescence (XRF) characterized ceramic compositions, while Monte Carlo Particle and Heavy Ion Transport code System (MC PHITS) simulations determined Linear Attenuation Coefficient (LAC), Half-value Layer (HVL), Mass Attenuation Coefficient (MAC), and Mean Free Path (MFP) within the 40–150 kV energy range. Comparative analysis between MC PHITS simulations and real setups was conducted. The C3–S9 ceramic sample, known for homogeneous full-color structure, showcased superior shielding attributes, attributed to its high density and iron content. Notably, energy levels considerably impacted radiation penetration. Overall, C3–S9 demonstrated strong shielding performance, underlining Malaysia's potential ceramic tile resources for X-ray room radiation shielding.

Spatial-temporal tracking of molecular species in Geant4-DNA track structure simulations

Monte Carlo particle transport codes have been used to simulate the radiolysis of water and subsequent evolution of molecular species during the chemical stage. In Geant4, accounting of molecules is performed using a molecule counter which provides only a global molecule count. We developed a modular counter implementation that was used to perform spatial-temporal tracking of molecule counts. This custom counter was validated in homogeneous and simplistic split geometries for 1 MeV electrons and 50 MeV protons.Using a spherical cell with one nucleus and one hundred mitochondria as an example, the radiolytic yields in the cytoplasm, nucleus, and mitochondria were compared between targeted (electrons originating from within mitochondria) and unspecific (electrons originating from within cytoplasm) for (very) low energy electrons with kinetic energies of 100 eV to 1 keV.At short times radiolytic yields in source volumes were consistent with reference yields obtained in a large homogeneous volume. Past 1 ns diffusion out of the source volumes led to yield reductions. For volumes abutting the source volumes radiolytic yields at short times strongly depended on track end effects, most pronounced for 100 eV electrons. Similarly, diffusion out of the source volumes greatly increased yields in abutting volumes past 1 ns with magnitude decreasing and time of increase increasing with increasing electron energy. For higher energy electron beams (100 keV and 200 keV) radiolytic yields in all volumes were consistent with reference yields for large statistics.Thus, in cases where energy deposit sites are uniformly distributed in the volume of interest and energies of the charged particles are high to avoid track end effects reference yields may be used in place of resource intensive chemical stage simulations. Spatially heterogeneous energy deposits, as observed in targeted ɑ or β radiation therapy or with targeted nanoparticles, using a spatially sensitive molecule counter is suggested when estimating radiolytic yields outside of the source volumes or at distances that are within the diffusion range of the molecules of interest.

Radiation protection studies for the INRNE-BAS cyclotron facility using Monte Carlo FLUKA code

The Bulgarian National Cyclotron Centre at the Institute for Nuclear Research and Nuclear Energy, Bulgarian Academy of Sciences (INRNE-BAS), is going to be a new facility dedicated to production of medical radioisotopes and research in the corresponding fields. This work details the shielding requirements of the facility and the preliminary radiological studies that we have carried out. The activity concentration of the long-lived radioisotopes is also evaluated which provides a useful guideline for the decommissioning plan of the facility. For these purposes we have performed simulations with the Monte Carlo particle transport and interaction code FLUKA.

A better hash method for high-fidelity Monte Carlo simulations on nuclear reactors

With the increasing demand for high-fidelity nuclear reactor simulations, the acceleration of Monte Carlo particle transport codes is becoming a core problem. One of the bottlenecks is locating millions or even billions of cells and fetching their associated parameters in the repeated geometry structure. Typically, Monte Carlo codes utilize a hash function to accelerate the cell locating and parameter indexing process. Specifically, they use the “cell vector → hash → parameter” method to accelerate the direct “cell vector → parameter” method. In this work, we propose a better hash method based on the Cyclic Redundancy Check (CRC) mechanism, which has been mathematically proven to be efficient and produce fewer hash collisions. Experimentally, this new hash method has been compared with some other hash functions and showed its superiority in terms of the calculation speed and collision probabilities. This hash method has been integrated into the Reactor Monte Carlo code RMC and worked well in practical applications.

Neutronic response of the neutral beam after EAST NBI upgraded

In order to improve the heating effect of neutral beam injector, the neutral beam injector (NBI) of the Experimental Advanced Superconducting Tokamak (EAST) is planned to be upgraded. After the EAST NBI is upgraded, the neutron emission intensity will increase. Neutron leaking from the neutral beam port penetrate the beamline, resulting in the material radiation damage. The neutronic response has been investigated by using the Monte Carlo particle transport code MCNP with the two dimensional RZ model of the EAST NBI. The results demonstrate that a 40% increase of the neutron emission rate has been attained with the upgraded EAST NBI. However, the simulations for the activation of different materials, displacement damage and nuclear heating of the neutral beam still satisfy the requirement of material safety performance. The calculations of nuclear response show that the materials can meet the requirements of the NBI steady state operation after the EAST NBI updated.

Effect of ion-induced nuclear reactions on structure modification and radiolysis in LiF irradiated by 410 MeV 36S ions

The objective of this research is to study the structural damage in LiF crystals irradiated with 410 MeV 36S ions. We report on the appearance of a remarkable structural damage far (hundreds of μm) beyond the penetration depth (136 μm) of sulfur ions. Selective chemical etching on samples irradiated at room temperature revealed the formation of extended defects within the ion-irradiated zone and also beyond the range of sulfur ions. Based on AFM image analysis, the defects were identified as nano-size colloids (10 nm). In the irradiated zone, nanoindentation tests indicated a significant hardening effect. Annealing experiments showed manifestations of intense radiolysis both in the ion-irradiated and beyond the range areas with a maximum of around 620 K. Comparison of experimental results with those of performed model simulation using Monte-Carlo particle transport code FLUKA support the assumption that the cause of the observed long-range damage is irradiation by secondary particles created in inelastic nuclear collisions between incident sulfur ions and lithium and fluorine nuclei of the target atoms. Neutrons, γ-rays and protons are rated as the dominating components of the secondary particle flux. The results can be helpful for assessing the risks for damage delocalization and for reasonable selection of ions in the applications of ion technologies to optical materials like LiF.

High-fidelity neutronics and thermal-hydraulics analysis of helical cruciform fuel based on small modular FHR

Fluorine High-temperature Reactor (FHR) is a novel reactor type without water-cooling, but with a high outlet temperature. Small Modular FHR is convenient for modular assembly and transportation. Therefore, it is suitable for remote and water shortage areas to provide efficient and clean nuclear energy. The neutronics and heat transfer of the fuel are the key issues when the small modular FHR has high thermal power output. This research proposes the fuel – TRISO (TRi-ISOtropic) particles dispersed in Helical Cruciform Fuel (HCF). HCF can be self-supported in radial direction, no additional grid-spacers are required, which can decrease the pumping power and improve the efficiency of the reactor system. For small modular FHR, it can be disassembled, transported, and assembled better without grid-spacers and simplifying the core structure. The heat transfer capacity of the coolant also could be further improved because of the special structure of HCF. Based on the Monte-Carlo particle transport code MCNP, the neutronic characteristic from the assembly to the whole core of FHR is studied. By obtaining the fission power distribution in near-critical condition, the thermal-hydraulics of the single most heated HCF and 7 bundle HCF in the heated assembly are calculated. In order to preliminary analyze the thermal-hydraulics of the whole core scale, the thermal-hydraulics analysis of porous media is carried out for 1/6 core. The isotropic resistance tensor and thermal equilibrium porous model have been introduced with reasonable approximation, and porous resistance relation has been carried out. The radial power peak factor is 1.185 and the axial power peak factor is 1.435 in near-critical condition, which proves the arrangement of the HCF and the control rods can flatten the power distribution effectively. The average temperature of the coolant outlet in the active region of the core is 974.44 K, and the average temperature of the upper plenum outlet is 972.85 K. High coolant outlet temperature can ensure effective supply of high-temperature process heat. And the highest temperature of the fuel is 1271.49 K, which is lower than the limitation 1573 K. The parameters of this research can supply the neutronics and thermal-hydraulics references for further FHR design.

A simulation optimization design of the filter stack spectrometer for laser-plasma interaction experiment

Filter stack spectrometers are widely employed in laser facilities for the spectrum measurement of bremsstrahlung photons. However, this method suffers from large uncertainty of unfolding due to its intrinsic limit resolution. For this, an optimization study on filter stack spectrometer is conducted. This procedure is implemented by a hybrid particle swarm optimization and genetic algorithm (PSO-GA). Monte-Carlo particle transport code Fluka is used for the simulation of the response matrix. Gravel algorithm, based on the least-square method, is used for the unfolding. For mono-energetic photons, this optimized filter stack spectrometer design provides a better energy resolution. For continuous distribution, this optimized filter stack spectrometer design yields a narrower unfolding solution space in the presence of measurement error.

Optically Stimulated Luminescent Response of the LiMgPO4 Silicone Foils to Protons and Its Dependence on Proton Energy

Modern radiotherapy (RT) techniques, such as proton therapy, require more and more sophisticated dosimetry methods and materials. One of the newly developed technologies is based on flexible sheets made of a polymer, with the embedded optically stimulated luminescence (OSL) material in the form of powder (LiMgPO4, LMP) and a self-developed optical imaging setup. The detector properties were evaluated to study its potential application in the proton treatment plan verification for eyeball cancer. The data showed a well-known effect of lower luminescent efficiency of the LMP material response to proton energy. The efficiency parameter depends on a given material and radiation quality parameters. Therefore, the detailed knowledge of material efficiency is crucial in establishing a calibration method for detectors exposed to mixed radiation fields. Thus, in the present study, the prototype of the LMP-based silicone foil material was tested with monoenergetic uniform proton beams of various initial kinetic energies constituting the so-called spread-out Bragg peak (SOBP). The irradiation geometry was also modelled using the Monte Carlo particle transport codes. Several beam quality parameters, including dose and the kinetic energy spectrum, were scored. Finally, the obtained results were used to correct the relative luminescence efficiency response of the LMP foils for monoenergetic and spread-out proton beams.

Thermal moderator-reflector design of the 24 Hz target station for the High Brilliance Neutron Source

The High Brilliance Neutron Source (HBS) project develops a high current accelerator driven neutron source (HiCANS), aiming to substitute existing neutron research reactors, that reach the end of life. This study focuses on the thermal neutron moderator design for a low frequency, long pulse target station of HBS. We investigate the neutronic characteristics of the D2O+H2O mixture moderator material using the Monte Carlo particle transport code PHITS. Our findings show that, compared to D2O with lower neutron absorption, confinement of H2O is the dominant factor in achieving a high neutron flux within the moderator and thus a high neutron brightness. We find that surrounding the central thermal water moderator with a D2O+H2O mixture can achieve a better balance between confinement and absorption. However, because of multiple angles of reflecting neutron, this design will also increase the divergence of the neutron beam.

Methods for processing and use of thermal neutron scattering data in OpenMC

OpenMC, an open source Monte Carlo particle transport code, relies on its own nuclear data format that is based on the HDF5 file format. The Python API within OpenMC includes an openmc.data module, which handles generation of thermal scattering HDF5 libraries starting from either ACE or ENDF files. When starting from ENDF files, NJOY is automatically executed to produce intermediate ACE files. In most cases, the resulting HDF5 data is equivalent to that stored in an ACE file. For incoherent elastic scattering, however, a novel sampling method is derived from the double-differential cross section and can be used to preserve the true continuous angular distribution. In this case, the characteristic bound cross section and Debye–Waller integral are taken directly from an ENDF file. Testing on the H in ZrH evaluation from ENDF/B-VIII.0 has demonstrated how the new sampling method matches the true distribution whereas the traditional treatment based on discrete cosines can leaves significant gaps in the angular distribution.

Modeling RBE with other quantities than LET significantly improves prediction of in vitro cell survival for proton therapy.

For proton therapy, a relative biological effectiveness (RBE) of 1.1 has broadly been applied clinically. However, as unexpected toxicities have been observed by the end of the proton tracks, variable RBE models have been proposed. Typically, the dose-averaged linear energy transfer (LETd ) has been used as an input variable for these models but the way the LETd was defined, calculated, or determined was not always consistent, potentially impacting the corresponding RBE value. This study compares consistently calculated LETd with other quantities as input variables for a phenomenological RBE model and attempts to determine which quantity that can best predicts proton RBE. The comparison was performed within the frame of introducing a new model for the protonRBE. High-throughput experimental setups of in vitro cell survival studies for proton RBE determination are simulated using the SHIELD-HIT12A Monte Carlo particle transport code. Together with LET, , here called effective Q (Qeff ), and Q are scored. Each quantity is calculated using the dose and track averaging methods, because the scoring includes all hadronic particles, all protons or only primaries. A phenomenological linear-quadratic-based RBE model is subsequently applied to the in vitro data with the various beam quality descriptors used as input variables and the goodness of fit is determined and compared using a bootstrapping approach. Both linear and nonlinear fit functions weretested. Versions of Qeff and Q outperform LET with a statistically significant margin, with the best nonlinear and linear fit having a relative root mean square error (RMSE) for RBE2Gy ± one standard error of 1.55 ± 0.04 (Qeff, t, primary ) and 2.84 ± 0.07 (Qeff, d, primary ), respectively. For comparison, the corresponding best nonlinear and linear fits for LETd, all protons had a relative RMSE of 2.07 ± 0.06 and 3.39 ± 0.08, respectively. Applying Welch's t-test for comparing the calculated RMSE of RBE2Gy resulted in two-tailed p-values of <0.002 for all Q and Qeff quantities compared to LETd, all protons . The study shows that Q or Qeff could be better RBE descriptors that dose averagedLET.

The development of a novel particle transport and thermal-hydraulic calculation chain for the European DEMO

A versatile calculation chain has been under development at VTT and Aalto University, featuring the ASCOT Monte Carlo orbit-following suite of codes, the Serpent Monte Carlo particle transport code and the Apros® thermal-hydraulic system code. This project aims to establish a comprehensive analytical environment that can aid researchers at multiple stages during the maturation of DEMO. Plasma product source term profiles have been generated using the ASCOT code, providing input for subsequent neutron transport calculations with Serpent. These studies utilized the CAD-based geometry of the equatorial breeding unit of the Water-Cooled Lithium-Lead (WCLL) breeding blanket. Efforts towards optimization have been made, investigating leakages and scalability of these external source simulations. Preliminary particle transport results were reported along with the methodology of converting the tallied data into Apros-relevant input.

Impact of the Lorentz force on electron track structure and early DNA damage yields in magnetic resonance-guided radiotherapy

Magnetic resonance-guided radiotherapy (MRgRT) has been developed and installed in recent decades for external radiotherapy in several clinical facilities. Lorentz forces modulate dose distribution by charged particles in MRgRT; however, the impact of Lorentz forces on low-energy electron track structure and early DNA damage induction remain unclear. In this study, we estimated features of electron track structure and biological effects in a static magnetic field (SMF) using a general-purpose Monte Carlo code, particle and heavy ion transport code system (PHITS) that enables us to simulate low-energy electrons down to 1 meV by track-structure mode. The macroscopic dose distributions by electrons above approximately 300 keV initial energy in liquid water are changed by both perpendicular and parallel SMFs against the incident direction, indicating that the Lorentz force plays an important role in calculating dose within tumours. Meanwhile, DNA damage estimation based on the spatial patterns of atomic interactions indicates that the initial yield of DNA double-strand breaks (DSBs) is independent of the SMF intensity. The DSB induction is predominantly attributed to the secondary electrons below a few tens of eV, of which energy deposition patterns are not considerably affected by the Lorentz force. Our simulation study suggests that treatment planning for MRgRT can be made with consideration of only changed dose distribution.