Radiation Transport Code Research Articles

The voxelized photon Monte Carlo method for hypersonic radiation modeling

A novel method of solving for radiative energy transfer at hypersonic speeds, the voxelized photon Monte Carlo (voxPMC) method, is presented. The voxPMC method is implemented in the Discrete Adaptive Radiative Transport (DART) code and results are given for radiative heat transfer of Mach 9.6 flow over a cylinder. Simulation parameters for this method were tuned to determine requirements for accurate solutions, using benchmark results from NASA’s Nonequilibrium Radiative Transport and Spectra Program (NEQAIR). It was found that of the tuning parameters, the mesh resolution and number of simulated photon packets had the largest effect on the resulting solution accuracy. It was found that 108 photon packets, or 5 photon packets per smallest mesh cell dimension, produced accurate results. Voxel resolution refinement and optical boundary conditions were also tested and 2 voxels per smallest mesh cell dimension and an optical symmetry boundary condition produced results that aligned with those given by NEQAIR. Future work will determine if recommendations for these parameters hold for a wider range of applications. Loose coupling between flowfield results and radiation analysis resulted in radiative cooling, which smeared the shock gradients and led to a slight increase in convective heating at the stagnation point.

Validation of the single-event method for low-energy electron transport via stopping power calculations with MCNP

Monte Carlo simulations of low-energy (<50 keV) electron transport in matter are essential for a broad range of application fields. Several Monte Carlo codes have developed specialized treatments for this case, but a comprehensive validation of low-energy electron transport for general-purpose simulations remains lacking in the literature. One approach to accomplish this validation is calculation of stopping power using low-energy electron transport physics, as stopping power is a fundamental radiation transport quantity which must be simulated accurately for nearly any application. In this work, we use the Monte Carlo N-Particle (MCNP) radiation transport code with the single-event method for electron transport to calculate stopping powers of low-energy electrons (50 eV to 30 keV) in 41 elemental solids, 14 compound solids, and five rare gas solids, comparing simulation results to published semi-empirical stopping power calculations from optical measurements. In general, the simulations give good agreement (typically within ±10%) with semi-empirical stopping power values at higher energies: 300 eV and above for most elemental solids, 1 keV and above for compound solids, and 400 eV and above for rare gas solids. Agreement between MCNP and semi-empirical values is worse below these energies. The most significant source of error is the EPRDATA14 cross section data, which does not account for changes in electronic structure due to solid-state bonding, particularly in compound materials. The simplistic model of atomic excitation used to generate the EPRDATA14 cross sections is another key source of error. Additionally, the breakdown of the continuous slowing-down approximation introduces significant uncertainty at low energies, although this is a limitation of the calculation method and not of the simulation procedure. Accounting for these and other uncertainty sources, the single event method in MCNP is robust and able to give good accuracy for a variety of low-energy electron transport problems through diverse kinds of materials.

The influence of Monte Carlo radiation transport codes and hadron physics models on stray neutron dose estimations in proton therapy: An extended intercomparison

PurposeSelecting a Monte Carlo (MC) radiation transport code and nuclear physics model affects the simulation of secondary neutron fluence energy spectra and the estimation of the neutron dose equivalent in proton therapy. This study compares the MC codes FLUKA, MCNPX and GEANT4 from previous work with two additional MC codes, TOPAS (with either the Binary Intra-nuclear Cascade (BIC) or the Bertini intra-nuclear cascade model (Bert)) and PHITS (with either the Intra-Nuclear Cascade of Liege (INCL) or the Bert model), to facilitate a more comprehensive comparison. Material and methodsSecondary neutron fluence energy spectra were scored at different locations inside a water phantom, exposed to a 10 × 10 cm2 parallel mono-energetic proton beam with energies of 110 MeV, 150 MeV, 180 MeV and 210 MeV. The neutron dose equivalent was estimated by applying fluence-to-dose equivalent conversion coefficients. ResultsDiscrepancies between MC codes and physics models reach 90 % for the total neutron fluence. Differences are most pronounced in the forward direction of the beam due to a notable presence of high-energy neutrons and the higher uncertainty on the neutron cross-section data for these neutrons. Nuclear model selection within PHITS (INCL or Bert) and TOPAS (BIC or Bert) shows a worse agreement for the Bert model than for the other MC codes and nuclear models. The discrepancies in the neutron spectrum simulations propagate into differences in the estimated neutron dose equivalents. They are more prominent for distal positions (up to a factor of 3) than for lateral positions (up to a factor of 2) and decrease at higher proton energies. ConclusionsThis study shows considerable differences in the out-of-field neutron fluence energy spectra calculated by different MC codes and nuclear physics models. Differences in the neutron dose equivalent reach a factor of 3. The discrepancies are most apparent in distal positions and at lower proton energies.

Mitigating space radiation using magnesium(-lithium) and boron carbide composites

The health effects of galactic cosmic radiation are a serious impediment to crewed exploration of the solar system. OLTARIS, an interface for the 3DHZETRN deterministic radiation transport code, was used to assess the response of aerospace materials to this constant radiation exposure. Traditional aerospace structural materials like aluminum can, after a certain mass, increase the health effects of such radiation. However, materials with lower atomic mass may mitigate this build-up in secondary radiation with increasing areal density. As such, lower atomic mass structural alloys of magnesium and magnesium–lithium are promising candidates. These alloys may reduce the mass of structures when substituted for aluminum alloys. Reinforcement with boron carbide could further reduce atomic mass while also improving the mechanical properties of such lightweight alloys. This study found that the lower atomic mass of these materials increased nuclear fragmentation upon cosmic radiation interactions, leading to a softening of the secondary (neutron) radiation spectra. This softened spectra reduced the effective dose equivalent, a measure of health effects, for magnesium(-lithium) alloys and their boron carbide-reinforced composites when compared to aluminum.

Radwaste analysis and package optimization with the F4E-radwaste tool and application to the ITER IVVS

A methodology to evaluate the neutron activation, categorize as radwaste, generate radiological checklists, and optimize the packaging for disposal of activated components according to French regulations is presented. The methodology obtains results like specific activity, dose rates and radwaste relevant parameters in a fine 3D mesh superimposed over the geometry of interest and can handle the typically complex models common in the nuclear fusion analyses like those of the ITER project. State-of-the-art radiation transport and activation codes are used along with a purposefully developed Python tool (F4E-radwaste), which performs the post-processing of the large amount of data generated and includes a graphical user interface. The 3D nature of the results allows the organization of packages of activated materials in a way that minimizes the amount of mass classified with higher radwaste levels and is subject to more stringent regulations and expensive disposals. An application case study of the methodology is presented for the In-Vessel Viewing System (IVVS) ITER system. The IVVS is especially suited as an example for this type of analysis as its elongated shape pointing towards the plasma results in a significant gradient of neutron flux and activation levels, and therefore radwaste categories. It is demonstrated how the methodology enables the reduction of the total higher radwaste level mass by optimizing the packaging of the IVVS radwaste.

Optimisation of the Tripathi model using a nuclear reaction cross-section database

Nuclear reaction cross-sections are an essential ingredient to reliable deterministic and stochastic radiation transport codes used for radiation protection in space and heavy-ion therapy applications. A recent study compared the existing literature data compiled within the open-access GSI-ESA-NASA cross-section database to the models implemented in the transport codes most commonly used for radiation protection in space and heavy-ion therapy applications. The outcome of the comparison was that none of the models fit well the experimental data for all projectile-target systems at all energy ranges. Therefore, the literature data were exploited to optimise the Tripathi–Cucinotta–Wilson model as reported in this work. This model is used as default in FLUKA, TRiP, and SpaceTRiP, it is part of the hybrid-Kurotama (HK) model used in particle and heavy ion transport code (PHITS), and it is implemented in Geant4. The consequences of using the proposed Tripathi–Cucinotta–Wilson optimisation in the HK model are also analysed.

Transmutation of Transuranic Elements as Solid Coating in Molten Salt Reactor Fuel Channel

The accumulation of spent nuclear fuel (SNF) is presently considered as a hindrance of the massive deployment of nuclear power plant, especially regarding the transuranic (TRU) elements. Eliminating TRU through transmutation is one of the most feasible alternative as a technical solution to solve the issue. This study explores the possibility of TRU transmutation using molten salt reactor (MSR) in a heterogeneous configuration, where a solid TRU is coated inside the fuel channel filled with liquid salt fuel. Such configuration is proposed to allow higher TRU loading into fluoride salt mixture without compromising the safety of the reactor. TRU coating was applied in consecutively outward radial fuel channel layers with coating thicknesses of 2.5 mm and 5 mm. Calculation was performed using MCNP6.2 radiation transport code and ENDF/B-VII.0 neutron cross section library. From the results, TRU coating with smaller thickness and positioned closer to the centre of the core exhibit higher transmutation efficiency due to exposure to higher neutron flux. Highest transmutation efficiency was achieved at 67.93% after 160 days of burnup. This shows a potential of achieving highly efficient TRU using heterogeneous configuration in MSR core.

Comparison of Doses in Lunar Habitats Located at the Surface and in Crater

As humanity prepares for extended lunar exploration, understanding the radiation environment on the Moon is important for astronaut safety. This study utilized the Particle and Heavy-Ion Transport code System (PHITS), a stochastic Monte Carlo-based radiation transport code, to simulate the radiation environment inside a habitat, focusing on the impact of galactic cosmic rays (GCRs) interacting with local lunar and habitat material, and to calculate the effective dose equivalent. Placing a lunar base in a crater can provide additional shielding by reducing the GCR flux incident on the base. Furthermore, the secondary radiation field created by GCR interactions may be altered by the local topological features. GCR transport calculations were performed for a hypothetical base on a flat surface and in shallow and deep craters to determine the overall efficacy in dose reduction gained by placing a base in a 100 m diameter crater. Our findings indicate that the depth of lunar habitats significantly influences the effective dose equivalent, with deeper locations offering substantial protection. Specifically, alongside a crater wall at a deep depth (15 m), in solar minimum conditions, the total dose was reduced by approximately 44.9% compared to the dose at the surface. Similarly, at a shallow depth (5 m), a reduction of approximately 10.7% was observed. As the depth of the crater increased, the neutron contribution to the total dose also increased. Comparing the simulated doses to NASA’s lifetime exposure limits provides insights into mission planning and astronaut safety, emphasizing the importance of strategic habitat placement and design.

The influence of graphite moderator density on the neutronic performance and fuel economy of a Molten Salt Reactor

Typical thermal molten salt reactor (MSR) employs graphite as its neutron moderator, due to its chemical compatibility and low neutron absorption. It is not without issues, however, since graphite provides positive temperature feedback when the MSR is fueled with U-233. Reducing the graphite density can possibly ameliorate this issue, but at the cost of other neutronic parameters, such as fertile conversion into fissile nuclide, and fuel economy, such as fissile makeup cost. This study evaluates the influence of graphite moderator density on the neutronic performance and fuel economy of an MSR. Passive Compact Molten Salt Reactor (PCMSR) design was used as the MSR model. To evaluate the neutronic parameters, MCNP6.2 radiation transport code with ENDF/B-VII.0 neutron cross section library were used. The evaluated neutronic parameters were temperature coefficient of reactivity (TCR), kinetic parameters, and conversion ratio (CR). Meanwhile, the fuel economy evaluation comprised the initial fuel inventory cost and fissile makeup cost. From the calculation results, lowering graphite density remarkably improves TCR, but also decreased CR in a notable amount. While effective delayed neutron fraction (βeff) was not particularly affected, mean neutron generation time (Λ) was reduced in lower graphite density. Considering the fuel economy, initial fuel cost is not significantly affected, but the fissile makeup cost can be sensitive of graphite density.

Subcritical neutron noise measurements for plutonium systems with varying geometry and mass

Recently, experiments were performed at the National Criticality Experiments Research Center (NCERC) with Zero Power Physics Reactor (ZPPR) plates in two very different geometries. The intent of this set of experiments was to provide the EUCLID (Experiments Underpinned by Computational Learning for Improvements in Nuclear Data) project team measured data of multiple radiation detector responses to be used in an adjustment of plutonium nuclear data. The work shown in this paper will focus on multiplicity results measured for subcritical arrangements of ZPPR plates in the two geometries. The Hage–Cifarelli formalism of the Feynman variance-to-mean method, a moments-based time-correlated neutron noise approach, was used. A validation of the methodology used to estimate uncertainties in the moments was performed and is presented. The effective neutron multiplication factor (keff) value were inferred from this analysis. The keffvalues were compared to simulated values using the radiation transport code Monte Carlo N-Particle® (MCNP). The results shown in this work provide evidence of larger bias in keffvalues as the geometric buckling of the system increases. This effect will be important for nuclear data validation and future adjustments of plutonium nuclear data.

Core nuclear design and heat transfer analysis of a megawatt-level SiC coated particle based vehicular micro reactor

As a versatile and diversified source of new energy, transportable micro reactors can offer high flexibility and the ability to provide power services even in remote areas. When utilizing coated fuel particles and supercritical carbon dioxide (S-CO2) coolant, the transportable micro reactor core achieves exceptional inherent safety and high power generation efficiency, rendering it incredibly competitive within the nuclear market. The Vehicular Micro Reactor (VMR) previously investigated is an integrated micro-reactor cooled by S-CO2, which employs TRistructural ISOtropic (TRISO) fuel particles dispersed in a conventional graphite matrix to generate 5 MW of power output. However, its drawback lies in the low fuel loading capacity. In recent year, the application of high-strength SiC materials in the field of nuclear energy has been developing continuously. In this paper, a new type of double-layered SiC coated fuel particle is designed, based on which a SiC-based Vehicular Micro Reactor (SVMR) is proposed and explored. It represents an improved design of VMR, with the main difference being its utilization of SiC as a matrix and filling by the two-layered SiC-coated fuel particle to achieve higher fuel loading than the VMR. In this paper, both the neutronic and heat transfer characteristics of the SVMR are discussed. The neutronic simulation is conducted using SCALE code based on Monte Carlo method, the module KENO IV was used to analyze the effective neutron multiplication factor (keff), and SCALE/TRITON (Transport Rigor Implemented with Time-Dependent Operation for Neutronic Depletion) coupled by 2-D radiation transport code NEWT and ORIGEN for burn-up calculation. By optimizing the assembly size and SiC-coated fuel particle size, the SVMR core achieves high fuel loading capacity and superior fuel utilization efficiency. The results of the heat transfer analysis also demonstrate that the SVMR core, which is entirely SiC-based, exhibits exceptional heat transfer capability. The excellent neutron economy and inherent safety make SVMR a more competitive micro reactor concept.

Development of nuclear de-excitation model EBITEM Ver.2

ABSTRACT The gamma de-excitation model of the general-purpose radiation transport code Particle and Heavy Ion Transport code System, called the Evaluated Nuclear Structure Data File (ENSDF)-Based Isomeric Transition and isomEr production Model (EBITEM) has been upgraded with a focus on precise neutron-capture reaction simulation. The first de-excitation subsequent to neutron capture of numerous nuclei, which was formerly simulated by a model based on the single particle model, is calculated using the Evaluated Gamma Activation File. The database used for further de-excitation, ENSDF, retrieved in 2013, was replaced with Reference Input Parameter Library 3 to consider internal conversion. The internal conversion process was interfaced with the atomic de-excitation model to assess the emission of Auger electrons and fluorescent X-rays. The spectra of gamma-rays from neutron-capture reactions calculated by the upgraded EBITEM correlate better with the evaluated cross-section data than those of the previous version.

EJUSTCO: Monte Carlo radiation transport code hybrid with ANN model for gamma-ray shielding simulation

EJUSTCO: Monte Carlo radiation transport code hybrid with ANN model for gamma-ray shielding simulation

Activation Analysis in Preparation for a Tungsten Irradiation Experiment at LANSCE

To organize the safe handling of activated material, knowing the residual dose rates is crucial. In this work, we present the pre-experiment activation analysis for an experiment in which tungsten blocks are irradiated by 800-MeV protons. In this analysis, we use the Monte Carlo N-Particle (MCNP) code for radiation transport, Attila4MC for unstructured mesh generation, and Activation in Accelerator Radiation Environments (AARE), including CINDER2008, for activation analysis. If the tungsten blocks must be removed within a day after the experiment, then exposure to personnel entering the room must be reduced. One exposure-reduction strategy is to add carbon steel shielding around the tungsten blocks, efficiently reducing the dose from the activated tungsten. However, the shielding becomes activated itself during irradiation: 56Mn is the dominant contributor for short decay times. The actual schedule at the time of the experiment allowed sufficient cool-off time for the tungsten in the room so that additional shielding was not necessary. A less rigorous comparison of the calculated values with the post-experiment measurements showed reasonable agreement.

Monte Carlo Radiation Transport for Astrophysical Transients Powered by Circumstellar Interaction

In this paper, we introduce SuperLite, an open-source Monte Carlo radiation transport code designed to produce synthetic spectra for astrophysical transient phenomena affected by circumstellar interaction. SuperLite utilizes Monte Carlo methods for semi-implicit, semirelativistic radiation transport in high-velocity shocked outflows, employing multigroup structured opacity calculations. The code enables rapid post-processing of hydrodynamic profiles to generate high-quality spectra that can be compared with observations of transient events, including superluminous supernovae, pulsational pair-instability supernovae, and other peculiar transients. We present the methods employed in SuperLite and compare the code’s performance to that of other radiative transport codes, such as SuperNu and CMFGEN. We show that SuperLite has successfully passed standard Monte Carlo radiation transport tests and can reproduce spectra of typical supernovae of Type Ia, Type IIP, and Type IIn.

SN 2022acko: The First Early Far-ultraviolet Spectra of a Type IIP Supernova

We present five far- and near-ultraviolet spectra of the Type II plateau supernova, SN 2022acko, obtained 5, 6, 7, 19, and 21 days after explosion, all observed with the Hubble Space Telescope/Space Telescope Imaging Spectrograph. The first three epochs are earlier than any Type II plateau supernova has been observed in the far-ultraviolet revealing unprecedented characteristics. These three spectra are dominated by strong lines, primarily from metals, which contrasts with the featureless early optical spectra. The flux decreases over the initial time series as the ejecta cool and line blanketing takes effect. We model this unique data set with the non–local thermodynamic equilibrium radiation transport code CMFGEN, finding a good match to the explosion of a low-mass red supergiant with energy E kin = 6 × 1050 erg. With these models we identify, for the first time, the ions that dominate the early ultraviolet spectra. We present optical photometry and spectroscopy, showing that SN 2022acko has a peak absolute magnitude of V = − 15.4 mag and plateau length of ∼115 days. The spectra closely resemble those of SN 2005cs and SN 2012A. Using the combined optical and ultraviolet spectra, we report the fraction of flux as a function of bluest wavelength on days 5, 7, and 19. We create a spectral time-series of Type II supernovae in the ultraviolet, demonstrating the rapid decline of flux over the first few weeks of evolution. Future observations of Type II supernovae are required to map out the landscape of exploding red supergiants, with and without circumstellar material, which is best revealed in high-quality ultraviolet spectra.

The Progenitors of Superluminous Type Ia Supernovae

Recent observations of type Ia supernovae (SNe Ia) have discovered a subclass of “super-Chandrasekhar” SNe Ia (SC SNe Ia) whose high luminosities and low ejecta velocities suggest that they originate from the explosions of white dwarfs (WDs) with masses that exceed the Chandrasekhar mass limit. Different models have been proposed to explain the progenitors of these explosions, including a “magnetized WD” model and a “WD merger” model. To test the robustness of these models, we conduct a 1D numerical parameter survey of WD explosions using these models as initial conditions. We follow the explosions using the hydrodynamics code Castro and then use the radiation transport code SuperNu to create light curves and spectra for the models. We find that while both classes of models fall within the range of SC SNe Ia observations on the light-curve width–luminosity relation, only the WD merger models reproduce the observed low ejecta velocities. The light curves of our merger models are more similar photometrically to observations than our magnetized models. Given this, we discuss possible explanations for the brightest SC SNe Ia observations that cannot be reproduced with our WD merger models. This study provides the basis for future SC SNe Ia observations and higher-dimensional numerical models.

Utilization of ACE nuclear data file toolkit ACEtk to calculate relative sensitivity coefficients of point-kinetics parameters

Sensitivity and uncertainty methods are quintessential for nuclear criticality safety and experiment design. This type of analysis relies on calculations of sensitivity coefficients; sensitivity coefficients of the effective neutron multiplication factor with respect to some nuclear data are predominantly calculated and used. As a part of the Laboratory Directed Research & Development project EUCLID (Experiments Underpinned by Computational Learning for Improvements in nuclear Data) at Los Alamos National Laboratory, sensitivity coefficients of many radiation detector measurement responses with respect to nuclear data were investigated. Specifically, this paper outlines a method to calculate point-kinetics parameters relative sensitivity coefficients with respect to nuclear data. Point-kinetics parameters such as the prompt neutron decay constant, effective delayed neutron fraction, and neutron generation time are especially important to experimenters and reactor operators designing systems with dynamic neutron populations. This method couples capabilities of the ACE (A Compact ENDF) nuclear data file toolkit, ACEtk, with the ability to load cross sections into the radiation transport code Monte Carlo N-Particle (MCNP). Key aspects of optimizing this method for a particular application and sensitivity profiles of the Jezebel criticality experiment are examined and discussed.

Shutdown dose rate analysis with the Shift Monte Carlo radiation transport code and modular verification workflow

Calculation of the shutdown dose rate is crucial for safe fusion reactor operations. The Rigorous-two-step (R2S) method is a method that requires connected neutron transport, activation, and gamma transport. Shift has integrated variance reduction with a deterministic solver Denovo, supports multiple geometry formats, and is scalable. These features make it an attractive transport solver choice for an R2S workflow. An R2S workflow for the Shift Monte Carlo code is developed and compared to the existing Oak Ridge National Laboratory Shutdown Dose Rate Code Suite (ORCS) workflow. Also, a Python framework for integrating two R2S workflows is developed to mix and match each step in the R2S workflow for improved collaboration and verification experience. Results show that the Shift-Denovo R2S workflow and the ORCS workflow calculate the shutdown dose rate of the ITER Shutdown Dose Rate benchmark problem with an average relative error of 2.285%.

Radiological Characterization with a Fluence Conversion Coefficients–Based Method: A Practical Example of the Preparatory Studies to the Pilot Beam at the CERN Large Hadron Collider

At high-energy accelerator facilities like the ones that are part of the accelerator complex at the European Organization for Nuclear Research (CERN), Monte Carlo radiation transport codes are widely employed to face the challenges of estimating radionuclide production yields and activities with the aim of performing the radiological characterization of activated components. Indeed, it is of paramount importance to ensure adequate radiation protection during scheduled maintenance, transport, and handling of these components and to establish their proper disposal pathway once they ultimately reach the end of their useful life. This paper summarizes the principles of the fluence conversion coefficients method that was developed as a complementary approach for radiological characterization studies. Then, the Monte Carlo simulations in preparation to the pilot beam run at the Large Hadron Collider at CERN in 2021 are presented as a practical example of possible applications. Finally, the flexibility of the method and the most relevant operational radiation protection implications are discussed in relation to the provided example.