Radiation Transport Code Research Articles

Radiation exposure of astronauts following an intense solar particle event: analysis and comparison of doses in male and female voxel phantoms.

According to NASA's plans, a human travel to the Moon is planned by the end of 2025 with the Artemis II mission, and humans should land on the Moon again in 2026. Exposure to space radiation is one of the main risks for the crew members; while for these short missions the doses from galactic cosmic rays would be relatively low, the possible occurrence of an intense solar particle event (SPE) represents a major concern, especially considering that in 2025 the Sun activity will be at its peak. Quantifying the amount and the effects of such exposure is therefore crucial, to identify shielding conditions that allow respecting the dose limits established by the various space agencies. By exploiting an interface between the BIANCA biophysical model and the FLUKA Monte Carlo radiation transport code, in this work we implemented a male and a female voxel phantom and we calculated absorbed doses and Gy-Eq doses in the various tissues/organs, as well as effective doses, following exposure to the August 1972 SPE, the most intense event of the modern era. The calculations were performed respect the organ dose limits for 30 d missions. A detailed comparison between male and female doses was then carried out, also considering that the Artemis II crew will include a woman. The results showed that female doses tend to be higher than male doses, especially with light shielding. This should be taken into account in mission design, also considering that, in a typical lunar mission, up to 15% of time may be spent in extra-vehicular activities, and thus with light shielding. More generally, this work outlines the importance of performing separate calculations for male and female astronauts when dealing with radiation doses and effects.

Liquid water radiolysis induced by secondary electrons generated from MeV-energy carbon ions.

Fast ion beams induce damage to deoxyribonucleic acid (DNA) by chemical products, including secondary electrons, produced from interaction with liquid water in living cells. However, the production process of these chemical products in the Bragg peak region used in particle therapy is not fully understood. To investigate this process, we conducted experiments to evaluate the radiolytic yields produced when a liquid water jet in vacuum is irradiated with MeV-energy carbon beams. We used secondary ion mass spectrometry to measure the products, such as hydronium cations (H3O+) and hydroxyl anions (OH-), produced along with ·OH radicals, which are significant inducers of DNA damage formation. In addition, we simulated the ionization process in liquid water by incident ions and secondary electrons using a Monte Carlo code for radiation transport. Our results showed that secondary electrons, rather than incident ions, are the primary cause of ionization in water. We found that the production yield of H3O+ or OH- was linked to the frequency of ionization by secondary electrons in water, with these electrons having energies between 10.9 and 550eV. These electrons are responsible for ionizing the outer-shell electrons of water molecules. Finally, we present that the elementary processes contribute to advancing radiation biophysics and biochemistry, which study the formation mechanism of DNA damage.

Fixed Source Sensitivity Calculations for Inertial Confinement Fusion Applications

A numerical code library was developed for the radiation transport code MCNP6.3 to calculate generalized response sensitivity coefficients for fixed source neutron transport problems with applications to inertial confinement fusion (ICF) experiments. The new MCNP6.3 dependency is used to generate a novel time convolution response that represents a neutron time-of-flight (nToF) signal. The traditional suite of macroscopic cross-section sensitivities and constrained fixed source probability distribution sensitivities are available for both the standard and the new response tallies in this library. However, novel sensitivity coefficients for the constrained hyperparameters of analytic fixed source probability distributions are emphasized in this work for their connection to ICF neutron transport models. Particularly, advanced Monte Carlo methods are developed for calculating the sensitivity of a nToF signal to perturbations in an ICF plasma’s ion temperature and burn history as well as perturbations in the target liner mass density and the shape parameters of the nToF detector’s impulse response function. Together, these capabilities form an advanced suite of computational tools that can be used to analyze and extract information from any ICF experimental platform.

Assessing the benefit of thorium fuel in a once through molten salt reactor

This study comprehensively evaluates thorium-based fuel utilisation in the Molten Salt Reactor (MSR) with a once-through fuel cycle. It examines the whether the use of thorium is beneficial in extending the fuel cycle length of an MSR, fuel utilisation efficiency, reactor safety, and reduction of long-lived radioactive waste. To observe those parameters, this research compares the use of thorium and uranium as fertile fuel in a single-fluid, once through MSR using Uranium-233 (U-233), Uranium-235 (U-235), and Reactor Grade Plutonium (RGPu) as well as the fissile driver. MCNP radiation transport code with ENDF/B-VII.0 continuous neutron library was used to analyse the neutronic parameters and fuel burnup in a multi-refuelling scheme for a total burnup time of 8 years. The findings indicate that the utilisation of thorium in conjunction with U-233 and U-235 significantly enhances fuel consumption and improve reactor safety. Such benefits, however, did not appear in RGPu. Moreover, thorium effectively reduces the production of long-lived radioactive waste, a crucial issue in nuclear waste management. Considering technical and environmental aspects, this study offers novel insights into using thorium as a more sustainable nuclear alternative under a once through fuel cycle, within a condition of multi-refuelling cycle which is only possible in an MSR.

New developments in the MCUNED-Plus code for radiation transport and coupled transport-activation computational simulations in accelerator-based facilities

The radiation hazard is one of the major aspects of concern in the design of a facility where it exists a risk of exposure to ionizing radiation. The nuclear analysis for radioprotection purposes requires an accurate description of the radiation fields. This is especially true in accelerator-based facilities where several types of particles are producing the radiation fields. Radiation fields exist during the accelerator operation, but delayed radiation may also be significant for a long time after operation ends. The radiation fields of concern present during the operation are due to the secondary neutrons and photons emitted by the interaction of accelerated particles with the accelerator components (beam leakage) or with the target. The radiation present after the operation is the residual photon field produced by activated materials, these materials being activated by both accelerated particles and secondary neutrons. This residual radiation field is relevant in high intensity accelerators like IFMIF-DONES.In order to address these complex radiation transport simulations, the D1SUNED code has been updated and the new code release renamed as MCUNED-Plus. The new developments include improvements in the light-ion transport like the implementation of a variance reduction for the production of secondary particles, and a new kinematics to reproduce the angular distribution of secondary particles emitted after deuteron breakup reaction. The calculation of the residual photon field in accelerator facility has also been improved by allowing to evaluate both light-ions and secondary neutrons induced shutdown dose rate in a single coupled simulation.

Thermal structure of circumbinary discs: Circumbinary planets should be icy, not rocky

The process of forming a circumbinary planet is thought to be intimately related to the structure of the nascent circumbinary disc. It has been shown that the structure of a circumbinary disc depends strongly on three-dimensional effects and on detailed modelling of the thermodynamics. Here, we employ three-dimensional hydrodynamical simulations, combined with a proper treatment of the thermal physics using the RADMC-3D radiation transport code, to examine the location of the snow line in circumbinary discs. The models have application to the circumbinary planets that have been discovered in recent years by the Kepler and TESS transit surveys. We find that the snow line is located in a narrow region of the circumbinary disc, close to the inner cavity that is carved out by the central binary, at typical orbital distances of ~ 1.5–2 au for the system parameters considered. In this region, previous work has shown that both grain growth and pebble accretion are likely to be inefficient because of the presence of hydrodynamical turbulence. Hence, in situ planet formation interior to the snow line is unlikely to occur and circumbinary planets should preferentially be icy, not rocky.

Comparison of MCNP and Microshield Dose Savings Determinations for Remote Methods of Transuranic Contamination Characterization.

The maturation of robotic and remote systems presents opportunities to expand the use of technologies that have typically been restricted to high-dose/high-risk nuclear work for moderate- or low-risk work to further reduce radiation exposure to workers. This study quantifies the potential dose savings achieved through the use of robotic techniques for characterizing transuranic-contaminated waste items and compares dose estimates from a simplistic, user-friendly deterministic radiation transport code and a more robust, complex Monte Carlo code. Three scenarios of transuranic-contaminated waste items described in published reports are modeled using representative source geometries in MicroShield and MCNP radiation transport codes. Estimated dose rates are determined at points ranging from 30 cm to 300 cm from the face of the waste item to represent the increase in distance allowed by robotic or remote system implementation for characterization activities. The dose rate savings are then converted to detriment cost savings using a dollar-per-person-dose conversion factor to provide a financial context. The radiation transport simulations show no consistent bias in estimated dose rate by varying simulation methodology or using geometrical simplifications-in some cases, MicroShield produces higher dose rate estimates while MCNP estimates are higher in other cases. In the MCNP simulations, the volume source geometry consistently produces a higher dose rate than the slab source geometry, but the MicroShield dose rate estimates do not display the same trend. Dose savings range from 1.60 × 10-5 μSv h-1 to 1.75 × 101 μSv h-1 with associated detriment cost savings from < 0.010 USD/person-h to 14 USD/person-h.

PUFFIn – A user friendly fast interface for calculating and visualizing the dose distribution in materials

This article describes a new graphical user interface that uses the PENELOPE Monte Carlo code to calculate dose distributions in materials. It is named PUFFIn, for the Penelope User-Friendly Fast Interface and was developed as an educational and scoping analysis tool for non-experts in radiation modeling. PUFFIn enables the user to visualize and compare the dose distributions in objects (e.g., sterilized healthcare products) irradiated with cobalt-60 gamma-rays, electron beam (E-beam) or X-rays. From such comparisons, the user can determine the most efficient product and/or packaging designs for any given radiation field – whether for conceptual or existing products and packaging.PUFFIn is distributed in a complete, self-contained package of software, including the PENELOPE radiation transport code, a standard graphics package, and a set of simple exercises. The package is available to any user at no cost and requires a minimal amount of training compared to other similar software. PUFFIn's capabilities are described, as well as validation measurements performed at Texas A&M University E-beam facility and the Aerial E-beam Facility in France.

Calculations of backpatch factors and conversion coefficients for the personal dose equivalent for X-ray beams used in radiodiagnostic

The purpose of this study is to use the Monte Carlo N-Particle radiation transport code (MCNP) to calculate the conversion coefficients of air kerma for personal dose equivalent (Hp [Sv]) and Back Scattering Factors (BSF) for X-ray beams used in diagnostic radiography at potentials of 40, 50, 60, 70, 80, 90, 100, 120, and 150 kVp. Additionally, effective dose is determined using male and female anthropomorphic voxel-based simulators (REX and REGINA) as recommended by the International Commission on Radiological Protection (ICRP) Publication 110, along with tissue weight factors from ICRP Publication 103 to simulate whole-body irradiation. The results indicate that the relationship between operational metrics such as Hp and risk limitation metrics like effective dose varies from 0.14 to 0.71 for Radiation Qualities in Radiation (RQR) beams from the X-ray source assembly and up to 0.64 for Radiation Qualities based on a phantom with an added aluminium filter (RQA). When comparing the obtained values for air kerma conversion coefficients for personal dose equivalent with the value adopted by the National Commission for Nuclear Energy (CNEN) of 1.14 Sv/Gy for individual dose determination, it is observed that at low energies, the values are overestimated by up to 44%, while at high energies, the values are underestimated by up to 67%.

A method to predict space radiation biological effectiveness for non-cancer effects following intense Solar Particle Events

In addition to the continuous exposure to cosmic rays, astronauts in space are occasionally exposed to Solar Particle Events (SPE), which involve less energetic particles but can deliver much higher doses. The latter can exceed several Gy in a few hours for the most intense SPEs, for which non-stochastic effects are thus a major concern. To identify adequate shielding conditions that would allow respecting the dose limits established by the various space agencies, the absorbed dose in the considered organ/tissue must be multiplied by the corresponding Relative Biological Effectiveness (RBE), which is a complex quantity depending on several factors including particle type and energy, considered biological effect, level of effect (and thus absorbed dose), etc.While in several studies only the particle-type dependence of RBE is taken into account, in this work we developed and applied a new approach where, thanks to an interface between the FLUKA Monte Carlo transport code and the BIANCA biophysical model, the RBE dependence on particle energy and absorbed dose was also considered. Furthermore, we included in the considered SPE spectra primary particles heavier than protons, which in many studies are neglected. This approach was then applied to the October 2003 SPE (the most intense SPE of solar cycle 23, also known as “Halloween event”) and the January 2005 event, which was characterized by a lower fluence but a harder spectrum, i.e., with higher-energy particles. The calculation outcomes were then discussed and compared with the current dose limits established for skin and blood forming organs in case of 30-days missions.This work showed that the BIANCA model, if interfaced to a radiation transport code, can be used to calculate the RBE values associated to Solar Particle Events. More generally, this work emphasizes the importance of taking into account the RBE dependence on particle energy and dose when calculating equivalent doses.

Expert‐in‐the‐loop design of integral nuclear data experiments

AbstractNuclear data are fundamental inputs to radiation transport codes used for reactor design and criticality safety. The design of experiments to reduce nuclear data uncertainty has been a challenge for many years, but advances in the sensitivity calculations of radiation transport codes within the last two decades have made optimal experimental design possible. The design of integral nuclear experiments poses numerous challenges not emphasized in classical optimal design, in particular, constrained design spaces (in both a statistical and engineering sense), severely under‐determined systems, and optimality uncertainty. We present a design pipeline to optimize critical experiments that uses constrained Bayesian optimization within an iterative expert‐in‐the‐loop framework. We show a successfully completed experiment campaign designed with this framework that involved two critical configurations and multiple measurements that targeted compensating errors in 239Pu nuclear data.

A new ground level neutron monitor for space weather assessment

We report on a new ground-level neutron monitor design for studying cosmic rays and fluxes of solar energetic particles at the Earth’s surface. The first-of-its-kind instrument, named the NM-2023 after the year it was standardised and following convention, will be installed at a United Kingdom Meteorological Office observatory (expected completion mid 2024) and will reintroduce such monitoring in the UK for the first time since ca. 1984. Monte Carlo radiation transport code is used for the development and application of parameterised models to investigate alternative neutron detectors, their location and bulk material geometry in a realistic cosmic ray neutron field. Benchmarked against a model of the current and most widespread design standardised in 1964 (the NM-64), two main parameterisation studies are conducted; a simplified standard model and a concept slab parameterisation. We show that the NM-64 standard is well optimised for the intended large-diameter boron trifluoride (BF3\\documentclass[12pt]{minimal} \\usepackage{amsmath} \\usepackage{wasysym} \\usepackage{amsfonts} \\usepackage{amssymb} \\usepackage{amsbsy} \\usepackage{mathrsfs} \\usepackage{upgreek} \\setlength{\\oddsidemargin}{-69pt} \\begin{document}$$_3$$\\end{document}) proportional counters but not for multiple smaller diameter counters. The new design (based on a novel slab arrangement) produces comparable counting efficiencies to an NM-64 with six BF3\\documentclass[12pt]{minimal} \\usepackage{amsmath} \\usepackage{wasysym} \\usepackage{amsfonts} \\usepackage{amssymb} \\usepackage{amsbsy} \\usepackage{mathrsfs} \\usepackage{upgreek} \\setlength{\\oddsidemargin}{-69pt} \\begin{document}$$_3$$\\end{document} counters and has the added advantage of being more compact, lower cost and avoids the use of highly toxic BF3\\documentclass[12pt]{minimal} \\usepackage{amsmath} \\usepackage{wasysym} \\usepackage{amsfonts} \\usepackage{amssymb} \\usepackage{amsbsy} \\usepackage{mathrsfs} \\usepackage{upgreek} \\setlength{\\oddsidemargin}{-69pt} \\begin{document}$$_3$$\\end{document}.

An improved methodology for modeling short pulse buried layer x-ray emission spectra

Radiation-hydrodynamic and spectroscopic modeling are important aspects of high energy density experimental design. In this paper, we improve the performance and capabilities over those obtainable with a previous methodology used for simulating x-ray emission spectra from buried layer targets heated by short pulse lasers. The improvement incorporates post-processing HYDRA radiation-hydrodynamic output with a non-local thermodynamic equilibrium atomic-kinetics radiation transport code, Cretin. Each code uses an independent radiation field which allows decoupling HYDRA's radiation group structure from Cretin's spectral output to improve the speed and flexibility of the design methodology. The execution time decreases from 2–3 days to a few hours while the flexibility of the improved methodology allows for performing sensitivity studies including a comparison of steady-state and time-dependent atomic kinetics and differences in the radiation group structure.

Neutronic Design Modification of Passive Compact-Molten Salt Reactor

Passive compact molten salt reactor (PCMSR) is a design concept of a molten salt reactor (MSR) currently under development in Universitas Gadjah Mada, Indonesia. It is designed as a thermal breeder reactor using thorium fuel cycle. However, our previous study shows that the original PCMSR design was incorrectly modelled, primarily overestimating its thorium breeding capability. To improve PCMSR neutronic design, we modified the core configuration by the addition of radial fuel channel layers previously nonexistent in original PCMSR core design in various configurations. Neutronic parameters of modified PCMSR geometries in the beginning of life (BOL) were simulated using MCNP6.2 radiation transport code with ENDF/B-VII.0 library. All variations of fuel layer addition show improvement in both temperature coefficient of reactivity (TCR) and breeding ratio (BR), with TCR values became more negative and BR values are larger than unity, ensuring proper breeding capability. Configuration Inner Core-Outer Blanket (IC-OB) achieves the largest BR and lowest doubling time (DT), whilst its TCR is an improvement from the original design. Therefore, IC-OB fuel layer configuration can be applied to redesign the original PCMSR and used in various design optimization scenarios.

A step-by-step simulation code for estimating yields of water radiolysis species based on electron track-structure mode in the PHITS code

Objective. Time-dependent yields of chemical products resulting from water radiolysis play a great role in evaluating DNA damage response after exposure to ionizing radiation. Particle and Heavy Ion Transport code System (PHITS) is a general-purpose Monte Carlo simulation code for radiation transport, which simulates atomic interactions originating from discrete energy levels of ionizations and electronic excitations as well as molecular excitations as physical stages. However, no chemical code for simulating water radiolysis products exists in the PHITS package. Approach. Here, we developed a chemical simulation code dedicated to the PHITS code, hereafter called PHITS-Chem code, which enables the calculation of the G values of water radiolysis species (•OH, eaq −, H2, H2O2 etc) by electron beams. Main results. The estimated G values during 1 μs are in agreement with the experimental ones and other simulations. This PHITS-Chem code also simulates the radiolysis in the presence of OH radical scavengers, such as tris(hydroxymethyl)aminomethane and dimethyl sulfoxide. Thank to this feature, the contributions of direct and indirect effects on DNA damage induction under various scavenging capacities can be analyzed. Significance. This chemical code coupled with PHITS could contribute to elucidating the mechanism of radiation effects by connecting physical, physicochemical, and chemical processes.

Changes in molecular conformation and electronic structure of DNA under 12C ions based on first-principles calculations

Fundamental studies on the interaction between ionizing radiation and deoxyribonucleic acid (DNA) can contribute to understanding the biological effects of radiation protection and radiotherapy. The transitional process from ionized and excited DNA to DNA lesions, such as strand breaks and base lesions, remains uncertain. In this study, we focused in particular on the direct effects of radiation on DNA and calculated the number of ionizations induced within 10 base pairs of DNA using the Particle and Heavy Ion Transport code System, a general-purpose Monte Carlo code for radiation transport. The electronic structure of the ionized DNA is subsequently calculated using the first-principles calculation software OpenMX. Through these calculations, we successfully identified the redox site of DNA and quantitatively evaluated the amount of molecular bond cleavage. The scientific findings of this study will elucidate the cleavage mechanism of DNA in the high-density radiation field realized by ion beam irradiation by determining the redox sites.

Activity measurements and calculations for gamma-emitting radionuclides in concrete drill cores of unit 2 of the Greifswald NPP

Due to Germany’s nuclear phase-out, decommissioning and final disposal of construction materials of Nuclear Power Plants (NPP) become increasingly important. The reliable determination of radionuclides produced by neutron activation, the activity as a function of time since shutdown and investigation of subsequent radionuclide mobility are subject of a research project. Drill cores of the concrete shielding of unit 2 of the Greifswald NPP were retrieved. Specific activities of gamma emitters and the elemental composition were measured. The radiation transport code MCNP 6 was used for the calculation of spectral neutron fluences. A neutron radiation field calculation reveals that the maximum neutron fluence at the concrete component is located in the floor just below the RPV. The concrete structures closest to the reactor core are shielded efficiently against neutron radiation by the annular water tank. Measured and calculated specific activities of 152Eu, 154Eu and 60Co for the cement screed at the position of the maximum neutron fluence are surprisingly low. A specific exemption (i.e. release from radiation protection surveillance but mandatory for final disposal) of the screed sample according to Germany’s Radiation Protection Ordnance is expected to be possible approximately 4 decades after the shutdown of the NPP.

Overview of PHITS Ver.3.34 with particular focus on track-structure calculation

This paper presents the latest updates on PHITS, a versatile radiation transport code, focusing specifically on track-structure models. Track structure calculations are methods used to simulate the movement of charged particles while explicitly considering each atomic reaction. Initially developed for radiation biology, these calculation methods aimed to analyze the radiation-induced damage to DNA and chromosomes. Several track-structure calculation models, including PHITS-ETS, PHITS-ETS for Si, PHITS-KURBUC, ETSART, and ITSART, have been developed and implemented for PHITS. These models allow users to study the behavior of various particles at the nano-scale across a wide range of materials. Furthermore, potential applications of track-structure calculations have also been proposed so far. This collection of track-structure calculation models, which encompasses diverse conditions, opens up new avenues for research in the field of radiation effects.

TRIPOLI-4® calculations of neutron multiplicity counting rates of 3He array detectors and verification tests of an automatic MCNP-to-TRIPOLI conversion tool for recent ICSBEP subcritical experiments

TRIPOLI-4® is a general-purpose Monte Carlo radiation transport code developed at CEA Paris-Saclay center. It uses continuousenergy nuclear data to simulate neutron transport for diverse nuclear engineering fields. Neutron multiplication calculations of subcritical systems are important for application areas including criticality safety, reactor ex-core monitoring, and non-proliferation of nuclear materials. Subcritical experiments were performed by the LANL teams to improve subcritical simulation capabilities of MCNP code and to validate 3He based array detectors. Three of these experimental series based on Ni, W, and Cu reflected Pu spheres and related MCNP calculations were published on 2020 ICSBEP handbook. New automation conversion tool of MCNP geometries to TRIPOLI-4® ones has been available. In this study, the traditional manual modelling of above Pu-W and Pu-Cu cases (17 configurations including Cu and polyethylene reflectors) was first prepared. The MCNP/TRIPOLI-4® converted ones were secondly set for Pu-Cu cases. Finally the singles counting rate R1 (cts/s) in a single NPOD detector for Pu-W cases and in two NoMAD detectors for Pu-Cu cases were calculated. Generally, a good agreement of R1 was obtained between the TRIPOLI-4® calculations and the ICSBEP reported results. When using the polyethylene reflector layers in some Pu-Cu experiments, a difference of 5 to 16 % for R1 was observed between TRIPOLI-4 (Converted) results and the ICSBEP measured ones.

Implementation of a Novel Monte Carlo Radiation Transport Code for Modeling Nonthermal Fe K$\alpha$ Line Generation in a Dense MagLIF Plasma

Implementation of a Novel Monte Carlo Radiation Transport Code for Modeling Nonthermal Fe K$\alpha$ Line Generation in a Dense MagLIF Plasma