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Copyright of Atomic Energy Science & Technology is the property of Editorial Board of Atomic Energy Science & Technology and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

We build a model to describe the net charges existing in the Sun and Earth. According to statistical mechanics, electrons on average move much faster than protons and neutrons at the same temperature. Electrons escape the Sun more easily than protons and neutrons, so the Sun becomes a charged star. We estimate the maximal net charges in the Sun by using statistical mechanics first. Then, we analyze the dynamical cycles between the positive and negative charged states. At a distance far away from the Sun, the effective net charges including the leaving protons and electrons are about 6.3 × 10 9 C with energies of 1 GeV initially. We also use another way based on the observations of the Earth's perihelion precession to estimate the minimum and maximum net charges between 1.15 × 10 8 C and 2.80 × 10 10 C in space from the Sun to Earth. The most charged particles from the Sun to the Earth are electrons, so both the Moon and Earth are impacted by them and very possibly have the same electricity. Next, we propose new physical mechanisms causing the slowdown of the Earth's spin and propose Coulomb's repulsive force resulting in the increasing distance between the Moon and Earth. As a result, it gives the net charges of 1.11 × 10 6 C surrounding the Earth and 8.29 × 10 3 C surrounding the Moon. Our estimations also correspond to early works. The charges surrounding the Sun and Earth cause the Earth to be long-term accelerated in the radial direction by Coulomb's force. Finally, using the effective net charges of the Sun and Earth, we calculate the increasing distance between 11.4 m and 19.4 m on average per century if the initial radial velocities of the Earth are in between 3.59 × 10 − 9 m / s and 6.12 × 10 − 9 m / s , which satisfies the observed reports. • A model describes the charges of the Sun and the force between the Sun and Earth. • The charged processes between two transient states are from negative to positive. • Both the Moon and Earth are impacted mostly by electrons from the Sun. • The negative charges of the Earth and Moon cause to exist Coulomb's repulsive force. • The Coulomb's force between the Sun and Earth can accelerate the Earth in the radial direction. [ABSTRACT FROM AUTHOR]

Purpose: The paper aims to discuss the response of the Micro-Mesh Gaseous Structure (Micromegas) detector to high-energy X-ray with 2.3 MeV energy using photon to neutron converters in addition to optimization of the detector components by Monte Carlo simulation. Materials and Methods: Methods of using Micromegas are different in terms of energy and intensity of high-energy X-ray. Response of Micromegas detector to X-ray with 2.3 MeV by different photoneutron converters was calculated by Monte Carlo N Particle X-Version (MCNPX) code. The number of electrons in the drift and multiplication regions and the depth-dose in the various regions of the detector were measured to study the response of the Micromegas detector to high energy X-ray. Also, the thickness of the upper electrode, and the type of gas in the detector were studied and optimized. Results: D2O with 2x10-5 efficiency is the best target to convert photons with 2.3 MeV energy to neutrons. It is the proper convertor to change the high energy X-ray into a photoneutron that can be detected by Micromegas. The optimum thickness of the upper electrode is 0.0026 cm for air and P10 gas in the detector. Conclusion: The results show that this detector can detect high-energy X-rays with energy above 2 MeV. The Monte Carlo results showed the output current of the Micromegas detector is 5.12 pA per one input hard X-ray photon. [ABSTRACT FROM AUTHOR]

Introduction: We investigate possible correlations between the stiffness of the symmetry energy at saturation density, the so-called L parameter, and the neutron skin thickness of 48Ca and 208Pb, for which the recent measurements from the CREX and PREX I + II experiments at the Thomas Jefferson Laboratory became available. Methods: We choose an ensemble of nucleonic equations of state (EoS) derived within microscopic (BHF, Variational, AFDMC) and phenomenological (Skyrme, RMF, DD-RMF) approaches. They are all compatible with the laboratory nuclear collisions data and with current observations of neutron stars (NS) mass and the tidal polarizability of a 1.4 M ⊙ NS, as deduced from the GW170817 event. Results: We find some degree of correlation between the L parameter and the neutron skin thickness whereas a much weaker correlation does exist with the tidal polarizability and the symmetry energy at saturation density. However, some EoS which are able to explain the CREX experimental data, are not compatible with the PREX I + II data, and viceversa. Conclusion: We confirm the results previously obtained with a different set of EoS models, and find a possible tension between the experimental data and the current understanding of the nuclear EoS. [ABSTRACT FROM AUTHOR]

Energy-resolved fast-neutron radiography is a powerful non-destructive technique that can be used to remotely measure the quantity and distribution of elements and isotopes in a sample. This is done by comparing the energy-dependent neutron transmission of a sample with the known cross-sections of individual isotopes. The reconstruction of the composition is possible due to the unique features (e.g. resonances) in the cross-sections of individual isotopes. At short-pulsed ( 1 ns) neutron sources, such information is accessible via time-of-flight neutron imaging in principle, but requires a detector with nanosecond temporal resolution. Conventional neutron detectors can meet this requirement only by heavily compromising spatial resolution or efficiency. Here, we present a unique approach on fast neutron resonance radiography using a scintillator-based event-mode imaging detector at a short-pulsed neutron source, including first results on spatially mapped resonance profiles using MeV neutrons. The event mode approach applied in the presented detector allows recording of individual neutron interactions with nanosecond precision in time and sub-mm resolution in space. As a result, the entire available neutron energy spectrum can be measured for each pulse. At the same time, the use of a thick scintillator screen and lenses to focus the produced light results in a highly flexible field of view and a high interaction probability in the sensitive volume of the detector. [ABSTRACT FROM AUTHOR]

The plastic scintillator detector is widely used to measure pulsed radiation fields, and a high-quality neutron energy response curve is crucial for accurately determining neutron yields. Traditionally, the neutron energy response curve can be obtained by the simulation of the performance of neutron detectors via the knowledge of the light yield of secondary particles or direct measuring energy response. Both methods record signals by a data acquisition system in the counting mode. This work establishes a new model to directly measure the energy response to neutrons of a plastic scintillator detector in current mode with white source spectra. The plastic scintillator detector can be triggered externally. There is no signal loss and dead time, and it is possible to calibrate the neutron detector with a high-intensity neutron source. Efficiencies can be determined at all energies simultaneously, and a rapid calibration of the plastic scintillator detector is available. To validate the experimental results, the theoretical values are obtained by Geant4 simulation, and the results reproduce the shapes of the experimental curves reasonably well. The neutron energy response curves of the detector show that when the thickness of the scintillator is reduced from 3 to 1 mm, the reduction ratio of gamma signal intensity is greater than that of neutron signal intensity, which leads to an increase in the neutron–gamma sensitivity ratio. Meanwhile, the thin scintillator thickness can obtain a somewhat flatter neutron energy response curve compared with other thicknesses. [ABSTRACT FROM AUTHOR]

In this work, we studied the light-output properties, efficiency function, as well as the pulse-shape discrimination (PSD) capability of p-Terphenyl scintillator. The selected solid cylindrical scintillation detector has a thickness of 45 mm and a diameter of 45 mm. Recently presented studies of light-output functions have only been measured for low-neutron energies. Our motivation has been to determine the light output function for p-Terphenyl scintillator more accurately over a wider neutron energy range. The measurements have been carried out with mono-energetic neutron beams in the wide energy range from 1.1 to 19 MeV. The neutron–gamma spectrometric system which we developed has been used for the measurement. The input analog signal from the detector was digitized with a fast 12-bits analog to digital converter with a sampling frequency of 1 GHz. Measured data from the detector are processed into the gamma and neutron spectra. The accurate light output function for the p-Therphenyl scintillator has been calculated. The pulse-shape discrimination capability, as well as the detection efficiency, of a p-Terphenyl scintillator are lower in comparison with a NE-213 equivalent detector. [ABSTRACT FROM AUTHOR]

Traditional computational methods for pressurized water reactors are unable to handle dispersed fuel particles as the double heterogeneity and the direct volumetric homogenization can result in significant errors. In contrast, reactivity-equivalent physical transformation techniques offer high precision for addressing the double heterogeneity introduced by dispersed fuel particles. This approach converts the double heterogeneity problem into a single heterogeneity problem, which is then subsequently investigated by using the conventional pressurized water reactor computational procedure. However, it is currently empirical and takes a lot of time to obtain the right k∞. In this paper, we train the RPT model by using the existing dataset of plate-dispersed fuel and rod-dispersed fuel by a machine learning method based on a linear regression model, and we then use the new data to make predictions and derive the corresponding similarity ratios. The burnup verification, density verification, fission rate verification, and neutron energy spectrum analysis are calculated through the OpenMC program. For plate-type fuel elements, the method maintains an accuracy within 200 pcm during depletion, with deviations in the 235U density and 235U fission rate within 0.1% and neutron energy spectrum errors within 6%. For rod-type fuel elements, the method maintains an accuracy within 100 pcm during depletion, with deviations in 235U and 239Pu density within 1.5% and neutron energy spectrum errors within 1%. The numerical validation indicates that the reactivity-equivalent physical transformation method based on the linear regression model not only greatly improves the computational efficiency, but also ensures a very high accuracy to deal with double heterogeneity in nuclear reactors. [ABSTRACT FROM AUTHOR]

From the viewpoints of the advantage depths (ADs), peak tumor dose and skin dose, we evaluated the effect on the dose distribution of neutron beam properties, namely the ratio between thermal and epithermal neutron fluxes (thermal/epithermal ratio), fast neutron component and γ -ray component. Several parameter surveys were conducted with respect to the beam properties of neutron sources for boron neutron capture therapy assuming boronophenylalanine as the boron agent using our dose calculation tool, called SiDE. The ADs decreased by 3% at a thermal/epithermal ratio of 20–30% compared with the current recommendation of 5%. The skin dose increased with the increasing thermal/epithermal ratio, reaching a restricted value of 14 Gyeq at a thermal/epithermal ratio of 48%. The fast neutron component was modified using two different models, namely the 'linear model', in which the fast neutron intensity decreases log-linearly with the increasing neutron energy, and the 'moderator thickness (MT) model', in which the fast neutron component is varied by adjusting the MT in a virtual beam shaping assembly. Although a higher fast neutron component indicated a higher skin dose, the increment was <10% at a fast neutron component of <1 × 10−12 Gy cm2 for both models. Furthermore, in the MT model, the epithermal neutron intensity at a fast neutron component of 6.8 × 10−13 Gy cm2 was 41% higher compared with that of 2 × 10−13 Gy cm2. The γ -ray component also caused no significant disadvantages up to several times larger compared with the current recommendation. [ABSTRACT FROM AUTHOR]

Copyright of Atomic Energy Science & Technology is the property of Editorial Board of Atomic Energy Science & Technology and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

Laser-driven deuterons generate neutrons with a mean energy of 2.5 MeV, through the 2H(d,n) fusion reaction in a deuterated polyethylene (dPE) tablet. The deuterium ions are accelerated by 12 fs, 21 mJ laser pulses interacting with a 0.2 µm thin dPE foil at a peak intensity of 1018 W/cm2. The laser was operated at 1 Hz repetition rate in bursts of 75 shots. The interaction was characterized and recorded for each laser shot. The ion spectra were measured in the forward and backward directions by Thomson ion spectrometers. Neutron events were detected by a time-of-flight (ToF) system consisting of four plastic scintillators positioned at various angles around the experimental chamber. The maximum cut-off energy of the forward accelerated protons and deuterons was close to 1.4 MeV and 1 MeV, while the mean values are 428 ± 63 keV and 433 ± 80 keV, respectively. Analysis of ToF distributions from 3128 shots resulted in an average yield of 1142 ± 59 neutrons per shot in the energy range of 1.5–4 MeV. The energy distribution of forward-directed neutrons peaks between 3 and 3.5 MeV. Angular dependence analysis showed a perpendicular minimum and a maximum along the deuteron beam, consistent with the expected distribution from the literature and our simulation results. [ABSTRACT FROM AUTHOR]

A successful attempt to produce a multi-function glass shield for attenuating neutrons and gamma rays by reinforcing a silicate glass network with boron and bismuth has been accomplished. A composition of 20SiO2-80Na2O (BSiBi0) was proposed to be used as a host glass network and prepared using the melt/annealing techniques. The low concentration of SiO2 in BSiBi0 was not sufficient to form a stable glass network. Then, the proposed BSiBi0 was modified with 10, 20, 30, and 40 mol% of each of B2O3 and Bi2O3 (BSiBi1, BSiBi2, BSiBi3, and BSiBi4) simultaneously. The structural effects of adding B3+ and Bi3+ were studied through X-ray diffraction, density, and FTIR, which all showed enhancement of glass forming ability, a former role of Bi3+ ions, and crowded the glass network by BO4 units. The derived structural parameters - molar volume, mean silicon – silicon separation, mean boron – boron separation, oxygen packing density, packing density, and number of bridging/non-bridging oxygen - were extensively discussed to explore the impact of B3+ and Bi3+ on the formed network. The richness of the proposed host glass network by B3+ and Bi3+ enhanced its thermal stability. The obtained elastic properties by ultrasonic measurements reflect the increase of the glass rigidity with increasing concentrations of B3+ and Bi3+ ions. The obtained glasses have high visible light transparency and almost complete UV absorption. The measured shielding parameters against two types of neutron energies (total slow and slow) and a wide range of gamma rays' energies showed a significant improvement in the shielding efficiency of the considered glasses. The total slow neutrons, slow neutrons, and gamma rays' attenuation abilities were improved by 22.9, 135.5, and 73.8 - 199.5%. High thermal stability, elasticity, visible light transparency, and neutrons and gamma rays' attenuation performance features give the produced glasses, especially BSiBi4 glass, preference as shielding materials in nuclear fields. [ABSTRACT FROM AUTHOR]

Commercial neutron dose meters (NDMs) provide estimates of neutron doses that may be inaccurate due to varying energy distribution of neutrons in workplace radiation fields. In this study, two NDMs (RadEye and DINEUTRON) were investigated by comparing their responses to a calibrated NDM (ELSE-Nuclear), which is traceable to a primary standards laboratory, and a validated ³He-based multishell neutron spectrometer (He-MuNS). The instruments were first calibrated using a Cf-252 reference field at the DOST-PNRI Neutron Laboratory (PNL). The Cf-252 field was then modified by introducing materials (concrete, aluminum, graphite, lead, polyethylene, and water phantom) to simulate workplace radiation fields. Calibration factors that were obtained using ELSE-Nuclear were found to be insufficient, as they presented different doses after the field was modified. Hence, He-BSS was used to derive correction factors (CF) for the three NDMs. Only concrete, aluminum, and graphite were selected for experimentation due to their significant results from computational studies of neutron modification, their undeniable presence in medical LINAC facilities, and their interaction with neutrons (Astronomo et al. 2021; Roslan et al. 2019; Borrelli et al. 2020). Results show that for facilities with neutron fields comparable to the concrete-modified field (58.8% thermal, 1.8% epithermal, and 39.4% fast), CF for RadEye, DINEUTRON, and ELSE-Nuclear are (0.493), (2.058), and (0.920), respectively. For the aluminum-modified field (15.5% thermal, 8.1% epithermal, and 76.4% fast), the CFs are (1.212), (1.894), and (0.872) respectively. Lastly, for the graphite-modified field (26.7% thermal, 16.3% epithermal, and 57% fast), the CFs are (0.785), (2.415), and (0.865), respectively. The results of this study can be applied by NDM users to facilitate more accurate neutron monitoring for radiation protection purposes. [ABSTRACT FROM AUTHOR]

The goal of this work is to characterize the secondary neutron spectra produced by 1 GeV/u56Fe beam colliding with a thick cylindric aluminum target and to perform a quantitative comparison with simulated results obtained with Monte Carlo codes. The measurements were performed using extended-range Bonner sphere spectrometers at two positions (15° and 40°) with respect to the beam direction. The secondary radiation field was simulated using four Monte Carlo codes (FLUKA, MCNP6, Geant4 and PHITS) and several physical models of nuclei transport and interaction. Neutron and proton energy distributions were simulated for the experimental measurement positions. The simulated neutron spectra, together with data measured with Bonner sphere spectrometers, after carrying out the correction of the contributions induced by the secondary protons, were used as input for the MAXED spectrum unfolding code to obtain the measured neutron spectra. Unfolded neutron spectra were compared with simulated ones to carry out a quantitative analysis of the performance of the chosen Monte Carlo codes and their corresponding physical models. This comparison showed that, because of experimental uncertainties and physical models, there are no unique solutions for each measurement location, but a range of solutions where the true experimental neutron spectra probably lie. The results showed deviations between 4.23% and 8.42% for some simulated spectra. Regarding the total integral values of neutron fluence and ambient equivalent dose, the unfolded neutron spectra showed deviations lower than 2%. [ABSTRACT FROM AUTHOR]

This study investigates the performances of tungsten, lead, and bismuth elements, which are widely used for shielding gamma-ray, from the neutron shielding point of view using the Monte Carlo N-Particle (MCNP6.2) simulation code. The study analyzed the neutron shielding capacities of Al matrix composites containing 30% W, Pb, and Bi. In addition, the shielding properties of secondary gamma-ray with an energy of 0.478 MeV resulting from neutron-boron interaction were also investigated. The results show that for thermal and intermediate energy neutrons, Al-30%W composite gives very good results by separating from Al-30%Pb and Al-30%Bi composites. However, Al-%30Pb and Al-%30Bi composites are also effective for fast neutrons. Regarding the gamma-ray, it was calculated that the Al-30%Pb composite exhibited the highest shielding performance. These findings provide important data on the contribution of W, Pb, and Bi elements used as gamma-ray shielding to neutron shielding in different neutron energy ranges and reveal that these elements should be used strategically with appropriate combinations depending on the neutron energy in neutron shielding designs. [ABSTRACT FROM AUTHOR]

Purpose: This study aims to quantify the secondary radiation dose caused by photoneutrons during prostate cancer treatment using an 18 MV medical linear accelerator (LINAC) through Monte Carlo simulations and experimental validation. Methods: Monte Carlo simulations were performed using G4Linac_MT to model the 18 MV photon beam of an Elekta LINAC. The simulation results were validated against experimental measurements. Neutron characteristics, including penetration, cross-section interactions, Linear Energy Transfer (LET), and dose contributions, were analyzed using an adult male ICRP phantom. Prostate treatment scenarios involved 3D-CRT plans with 4-fields, 5-fields, and 7-fields. Specific absorbed fractions (SAFs) in various organs were also evaluated. Results: Simulation and experimental measurements showed strong agreement, with a dose error of approximately 0.74%, and 97% of dose points passed a 2%/2 mm gamma index. Intermediate neutrons constituted 87.05%, while 12.95% were fast neutrons. Neutron dose contributions were 0.63%, 0.33%, and 0.77% for the 3D-CRT 4-field, 5-field, and 7-field plans, respectively. SAF values decreased as neutron energy increased, highlighting reduced neutron interaction efficiency at higher energies. Conclusions: Monte Carlo simulation is a reliable approach for evaluating neutron dose contributions in high-energy X-ray LINACs. Optimization of treatment plans is essential to minimize neutron-induced dose contributions. [ABSTRACT FROM AUTHOR]

Copyright of Atomic Energy Science & Technology is the property of Editorial Board of Atomic Energy Science & Technology and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

Boron Neutron Capture Therapy (BNCT) is performed using high-intensity neutron sources; however, the energy of the primary neutrons is too high for direct patient irradiation. Thus, neutron moderation is mandatory and is performed using a device known as a Beam Shaping Assembly (BSA). Due to the differences in flux and energy spectra between neutron sources, each facility needs a dedicated BSA design, whether it is based on a nuclear reactor or, more recently, on an accelerator. Since moderation involves the loss of neutrons, typically by a factor of 1000, it is necessary to generate a very high flux before neutrons pass through the BSA. We propose a novel approach that eliminates the necessity of a BSA, BSA-free, by generating neutrons suitable in flux and energy for direct patient irradiation through the Sc(p,n) Ti reaction using near-threshold protons. Our findings demonstrate that all IAEA quality factors for BNCT can be met with existing proton accelerators. Additionally, figures of merit studied provide similar results compared to real BNCT facilities. This breakthrough opens up new avenues in BNCT, among others, the control of the neutron penetration within the human body by small changing in the proton energy. Also, it is expected simplified accelerator-based facilities in terms of manufacturing and maintenance and operation. This work is a study based on experimental data and Monte Carlo simulations. Technical challenges and safety are addressed in Discussion section. This novel proposal is under evaluation as patent. [ABSTRACT FROM AUTHOR]

This article presents a kinetic study of fusion burn waves in compressed deuterium-tritium and proton-boron plasmas. The study uses particle-in-cell simulations with Monte Carlo collisions to analyze the energy balance in the expansion of a hot-spot, capturing alpha particle heating and radiative losses. The simulations show that the energy radiated by the hot-spot is recaptured by the surrounding high-density fuel, resulting in a net fusion energy production that is twice the initial hot-spot energy. The study provides insights into the feasibility of proton-boron fusion as an alternative fusion fuel. The given text describes a simulation model for studying Coulomb collisions, fusion reactions, and bremsstrahlung radiation in high-density, highly-collisional systems. The model uses a binary Monte Carlo collision method to simulate these interactions between particles. The text explains the implementation of bremsstrahlung emission and inverse bremsstrahlung absorption in the simulation model. The model is verified through simulations of bremsstrahlung emission from an electron beam incident on solid boron, laser heating of a hydrogen plasma, and photon emission and absorption in a p-11B plasma. The given text discusses the results of simulations on various plasma scenarios. The simulations include laser heating of a hydrogen plasma, fusion burn waves in compressed DT and p-11B plasmas, and the measurement of radiation and fusion power density in a fully ionized p-11B plasma. The results show close agreement between the simulations and other [Extracted from the article]

Graphite is widely used in molten salt reactors (MSRs) because of its excellent properties. However, the irradiation lifespan of graphite in MSR is much shorter than the life of a conventional nuclear power plant, which lowers the load factor and increases the economic burden. In this paper, to extend the graphite irradiation lifespan, one uniform radial fast neutron field was designed by changing the fuel distribution in different radial fuel channels, which is called continuous core zoning. A positive correlation correction algorithm was used to adjust the fuel volumetric fraction (VF) of each zone. The optimization was carried out among different reactor sizes and different thicknesses of the reflector to study flattening characteristics. After optimization, the difference in fast neutron flux between different zones of the flattened region was less than 1%. Compared to the unoptimized case, the peak value of the fast neutron flux can be most reduced by 48.6%. The study employed a single‐channel heat transfer model to investigate the temperature distribution within the both pre‐ and postoptimization core structure. The flow matching inlet condition and the average velocity inlet condition were considered. The results showed that to eliminate hot spots, a flow matching design is needed; otherwise, the temperature will rise greatly. Then the lifetime of graphite was calculated by combining fast neutron flux and temperature. Under average velocity inlet condition, graphite lifespan in CORE550‐20 increased from 11.4 years of the unoptimized core to 18.9 years of the optimized core, only representing a 66% enhancement due to higher temperatures. Under flow matching conditions, the lifespan of graphite in CORE450‐00 can be extended from 13.3 years of the unoptimized core to 26.8 years of the optimized core, indicating a 102% improvement. [ABSTRACT FROM AUTHOR]

The Scibar Cosmic-Ray Telescope (SciCRT) is the most promising detector of the Sierra Negra Cosmic Rays Observatory (SN-CRO). At this location, being a target and a tracker of secondary cosmic rays, the SciCRT offers a high probability of observing solar energetic particles and lower energy galactic cosmic rays (LEGCR); also, it allows the identification of incoming particles by measuring their energy deposition. In this work we present a Geant4-based simulation of the energy deposited by neutrons, γ -rays, protons, electrons and muons in the optimally running SciCRT components. We also calculated the detection efficiency of the SciCRT at its current state. Our simulation results provide new information about the SciCRT detection response that may be used as a basis to estimate and analyze the energy spectra of primary particles. [ABSTRACT FROM AUTHOR]

Zirconium hydride is commonly used for next-generation reactor designs due to its excellent hydrogen retention capacity at temperatures below 1000 K. These types of reactors operate at thermal neutron energies and require accurate representation of thermal scattering laws (TSLs) to optimize moderator performance and evaluate the safety indicators for reactor design. In this work, we present an atomic-scale representation of sub-stoichiometric ZrH2−x (0.3 ≤ x ≤ 0.6) , which relies on ab initio molecular dynamics (AIMD) in tandem with velocity auto-correlation (VAC) analysis to generate phonon density of states (DOS) for TSL development. The novel NJOY+NCrystal tool, developed by the European Spallation Source community, was utilized to generate the TSL formulations in the A Compact ENDF (ACE) format for its utility in neutron transport software. First, stoichiometric zirconium hydride cross sections were benchmarked with experiments. Then sub-stoichiometric zirconium hydride TSLs were developed. Significant deviations were observed between the new δ -phase ZrH2−x TSLs and the TSLs in the current ENDF release. It was also observed that varying the hydrogen vacancy defect concentration and sites did not cause as significant a change in the TSLs (e.g., ZrH1.4 vs. ZrH1.7) as was caused by the lattice transformation from ϵ - to δ -phase. [ABSTRACT FROM AUTHOR]

A scheme for a quasi-monoenergetic high-flux neutron source with femtosecond duration and highly anisotropic angular distribution is proposed. This scheme is based on bulk acceleration of deuteron ions in an optical trap or density grating formed by two counter-propagating laser pulses at an intensity of ∼ 1 0 16 W / c m 2 in a near-critical-density plasma. The deuterons are first pre-accelerated to an energy of tens of keV in the ambipolar fields formed in the optical trap. Their energy is boosted to the MeV level by another one or two laser pulses at an intensity of ∼ 1 0 20 W / c m 2 , enabling fusion reactions to be triggered with high efficiency. In contrast to previously proposed pitcher–catcher configurations, our scheme can provide spatially periodic acceleration structures and effective collisions between deuterons inside the whole target volume. Subsequently, neutrons are generated directly inside the optical trap. Our simulations show that neutron pulses with energy 2–8 MeV, yield 1018–1019n/s, and total number 106–107 in a duration ∼ 400 fs can be obtained with a 25 μm target. Moreover, the neutron pulses exhibit unique angularly dependent energy spectra and flux distributions, predominantly along the axis of the energy-boosting lasers. Such microsize femtosecond neutron pulses may find many applications, such as high-resolution fast neutron imaging and nuclear physics research. [ABSTRACT FROM AUTHOR]

Fracture structures are important natural gas transport spaces in shale gas reservoirs, and their storage state in shale gas reservoirs seriously affects gas production and extraction efficiency. This work uses numerical modeling techniques to investigate the logging response law of the thermal and epithermal neutrons in the gas-containing fracture environment at various angles, applying neutron logging as a technical method. To increase the precision of the evaluation of the natural gas storage condition in shale gas reservoirs, the angle of the fractures' neutron logging data is analyzed. It is found that even in an environment with the same porosity of the fractures, there are significant differences in the logging results due to the different angles of the fracture alignment: 1. the neutron counts in the high-angle (70–90°) fracture environment are 2.25 times higher than in the low-angle (0–20°), but the diffusion area of the neutrons is only 10.58% of that in the low-angle (0–20°); 2. in the neutron energy spectrum, neutron counts are spreading to the high-energy region (7–13 MeV) along with the increase in the angle of the fracture, and the feature is especially prominent in the approximately vertical (60–90°) fracture environment, which is an increase of 528.12% in comparison with the counts in the approximately horizontal angle (0–30°) environment. The main reason for these differences is the variation in the volume of the fracture within the source radiation. This volumetric difference results from the variation in fracture angles (even though the fracture porosity is the same). In view of the above phenomenon, this paper proposes the concept of "effective fracture volume", which can intuitively reflect the degree of influence of fracture angle on neutron logging results. Further, based on the unique characteristics of shale gas reservoirs and neutrons, this paper provides important theoretical support for the modification of the porosity of the field operation, the evaluation of the physical characteristics of the gas endowment space, and the assessment. [ABSTRACT FROM AUTHOR]

The asteroid impacts can be devastating for the life on the Earth. Thus, mitigation of this unwanted event is of crucial importance, and nuclear devices are state-of-the-art choice. The presented paper focuses on nuclear data validation of silicon and oxygen. Nuclear data validation is an essential task because nuclear data are the deciding factor for energy deposition profiles, which serve as input for subsequent hydrodynamic simulation, thus directly affecting the description of asteroid deflection. The silicon and oxygen in the form of silicate materials are abundant in asteroids. The set of neutron and gamma transport experiments using a primary neutron standard 252Cf neutron source was performed. The fast neutron spectrum in the energy range of 2–10 MeV was measured using a stilbene detector. The best agreement for fast neutron leakage spectra was reached using INDEN evaluation for 28Si and ENDF/B-VIII.0 16O evaluation. Activation foils sensitive in different energy ranges were utilized. The results depend on the examined reaction and distance from the source. Finally, the prompt gamma rays from capture and inelastic scattering were measured by an HPGe detector. With the exception of 28Si(n,n′) 1779 keV peak, other prompt gamma production peaks show significant discrepancies. The calculations for comparisons of different transport libraries were performed using Monte Carlo MCNP6.2 code. The results show the necessity of further improvement of evaluations for oxygen and silicon. • Energy deposition profile highly depends on used transport library. • Effect of different transport libraries on calculations. • Measurement of leakage neutron spectra from sand cylinder • Measurement of reaction rates of activated materials. • Measurement of prompt gammas. [ABSTRACT FROM AUTHOR]

Neutron choppers, which work with rotating blades and remove neutrons with unwanted energy, can allow neutrons with certain energy to pass through. In the design of a neutron velocity selector and the crucial role of these blades in its operation, simulation and experimental studies were performed to investigate the material and type of coating of the blades in relation to thermal neutron absorption. In this study, blades made of cadmium, aluminum, and carbon fiber with different coatings were exposed to thermal neutrons, and the neutron transmission was calculated using the MCNPX code. Based on the simulation results, blades with cadmium coatings and coatings of B4C, LiF, and H3BO3 with specific thicknesses were designed. The thermal neutron absorption of the constructed blades was measured in a thermal neutron flux from an Am-Be source using a BF3 detector. The results show that the cadmium blade without coating (1 mm thick), the aluminum blade (0.5 mm thick), and the carbon fiber blade (0.9 mm thick), both with B4C coatings, have the highest thermal neutron absorption values. The simulation and experimental results show that a B4C coating with a concentration of 35-45 gr/cm² is suitable for neutron velocity selector blades. [ABSTRACT FROM AUTHOR]

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The calculation of effective doses is crucial in many medical and radiation fields in order to ensure safety and compliance with regulatory limits. Traditionally, Monte Carlo codes using detailed human body computational phantoms have been used for such calculations. Monte Carlo dose calculations can be time-consuming and require expertise in different processes when building the computational phantom and dose calculations. This study employs various machine learning (ML) algorithms to predict the organ doses and effective dose conversion coefficients (DCCs) from different anthropomorphic phantoms. A comprehensive data set comprising neutron energy bins, organ labels, masses, and densities is compiled from Monte Carlo studies, and it is used to train and evaluate the supervised ML models. This study includes a broad range of phantoms, including those from the International Commission on Radiation Protection (ICRP-110, ICRP-116 phantom), the Visible-Human Project (VIP-man phantom), and the Medical Internal Radiation Dose Committee (MIRD-Phantom), with row data prepared using numerical data and organ categorical labeled data. Extreme gradient boosting (XGB), gradient boosting (GB), and the random forest-based Extra Trees regressor are employed to assess the performance of the ML models against published ICRP neutron DCC values using the mean square error, mean absolute error, and R2 metrics. The results demonstrate that the ML predictions significantly vary in lower energy ranges and vary less in higher neutron energy ranges while showing good agreement with ICRP values at mid-range energies. Moreover, the categorical data models align closely with the reference doses, suggesting the potential of ML in predicting effective doses for custom phantoms based on regional populations, such as the Saudi voxel-based model. This study paves the way for efficient dose prediction using ML, particularly in scenarios requiring rapid results without extensive computational resources or expertise. The findings also indicate potential improvements in data representation and the inclusion of larger data sets to refine model accuracy and prevent overfitting. Thus, ML methods can serve as valuable techniques for the continued development of personalized dosimetry. [ABSTRACT FROM AUTHOR]

The Timepix2-based hybrid-pixel detector with a 500 μm thick silicon sensor was employed for fast-neutrons registration to be applied in neutron radiography of metallic printed circuit heat exchanger (PCHE). Two energies of neutrons were experimentally tested. The detection of 3.55 MeV neutrons from the deuteron–deuteron (DD) reaction was compared to 15.7 MeV neutrons from the deuteron–tritium (DT) neutron generator. In order to distinguish the signal induced by the registered neutrons from the accelerator background, filtration of the recorded particle spectral tracks was applied. The benefit of applying hydrogen-based converter layer for 3.55 MeV neutrons was observable. On the other hand, in the case of 15.7 MeV neutrons, the direct registration by interaction with the sensor Si significantly dominates the conversion. [ABSTRACT FROM AUTHOR]

In this paper, the temperature dependence of single event upset (SEU) cross-section in 28 nm embedded Static Random Access Memory (SRAM) of System in Package (SiP) was investigated. An atmospheric neutron beam with an energy range of MeV~GeV was utilized. The SEU cross-section increased by 39.8% when the temperature increased from 296 K to 382 K. Further Technology Computer Aided Design (TCAD) simulation results show that the temperature has a weak impact on the peak pulse current, which is mainly caused by the change of bipolar amplification effect with temperature. As the temperature increases, the critical charge of the device decreases by about 4.8%. The impact of temperature on the SEU cross-section is determined competitively by the peak pulse current and the critical charge. The impact of temperature on critical charge is expected to become more severe as the feature size is further advanced. [ABSTRACT FROM AUTHOR]

Deuterium-tritium fusion reactions produce energy in the form of 14.1 MeV neutrons, and hence fusion reactor components will be exposed to high energy neutron irradiation while also being subjected to thermal, mechanical and magnetic loads. Exposure to neutron irradiation has numerous consequences, including swelling and dimensional changes, comparable in magnitude to the peak transient thermal deformations occurring in plasma-facing components. Irradiation also dynamically alters the various thermo-mechanical properties, relating temperature, stress and swelling in a strongly non-linear way. Experimental data on the effect of neutron exposure spanning the design parameter space are very sparse and this highlights the relevance of computer simulations. In this study we explore the equivalence between the body force/surface traction approach and the eigenstrain formalism for treating anisotropic irradiation-induced swelling. We find that both commercial and massively parallelised open source software for finite element method (FEM) simulations are suitable for assessing the effect of neutron exposure on the mechanically loaded reactor components. We demonstrate how two primary effects of irradiation, radiation swelling and the degradation of thermal conductivity, affect the distributions of stress and temperature in the divertor of the ITER tokamak. Significant uncertainties characterising the magnitude of swelling and models for treating it suggest that on the basis of the presently available data, only an order of magnitude estimate can be given to the stress developing in reactor components most exposed to irradiation during service. [ABSTRACT FROM AUTHOR]

Identifying and characterizing the decay chains forms a crucial part of nuclide identification in the synthesis of super-heavy elements (SHEs), which is achieved via cold or hot fusion reactions. A systematic theoretical calculation of alpha-decay half-lives in this region of the periodic system may be useful in the identification of new nuclei in both types of reactions. Here, we obtain the alpha-decay half-lives in the framework of the dynamical cluster-decay model (DCM) based on quantum mechanical fragmentation theory (QMFT), built using pre-formation probability and penetration probability. We calculate the α-decay half-life using the DCM (ℓ = 0 case), which considers the compound nucleus hot due to its recoil energy after neutron emission, providing a more accurate description than other models. Decay half-lives are compared with experimental results. [ABSTRACT FROM AUTHOR]

This paper investigates the neutron diffusion model with two energy groups in spherical reactors. In particular, the integer-order two energy groups neutron diffusion model in spherical reactors is resolved using the Laplace transform method by regarding the spherical radius r as a time domain. Next, we transform the neutron diffusion model into fractional-order versions using the Caputo differentiator, resulting in what is referred to as the fractional-order two-energy-group neutron diffusion model. To address this fractional-order system, we introduce a novel approach to reduce a system of 2a -order to a duplicated system of a -order, where 0 <α 1. This converted system is then solved using one of the recent modifications of the fractional Euler method called the Modified Fractional Euler Method (MFEM). Several numerical simulations are depicted to verify our findings. [ABSTRACT FROM AUTHOR]

The nuclear structure effects of correlated pairs of fission fragments in the mass distribution spectrum of the fissioning nucleus are investigated using the fragmentation theory based on the asymmetric two-centre shell model. The model is applied to explain the observed structure in relative mass distribution yields of even Z–even N correlated fission fragment pairs in the low-energy heavy-ion induced fusion–fission reactions 208 Pb (18 O,f) and 238 U (18 O,f). In each of these reactions, the respective fused system is considered to utilise its excitation energy in 0–14 neutron evaporation channels before proceeding to decay. The fissioning isotopes ( 226 - 212 Th and 256 - 242 Fm) resulted after neutron evaporation (in the step of Δ n = 2 ) contribute to the observed even–even individual fission fragments mass distribution spectra. It is shown that the most probable decay channel consisting of correlated pair of magic and deformed shell fission fragments always has a larger mass distribution yield than that containing both magic shell structure fission fragments. The pear-shaped structure that emerges from the combined magic plus deformed shell fission fragments might have relevance to an occurrence of stable octupole deformation of a fissioning nucleus before saddle configuration, which was first noticed by Johansson (S A E Johansson, Nucl. Phys. 22 , 529 (1961)) and more recently, authenticated by Scamps and Simenel (G Scamps and C Simenel, Nature 564 , 382 (2018)). Thus, the combined quantum shell effects of magic and deformed fission fragments both in potential and the cranking mass parameters decide the most probable decay modes (i.e., the correlated pairs of fission fragments with maximum yields) in fusion–fission reactions ( 208 Pb (18 O,f) and 238 U (18 O,f)). [ABSTRACT FROM AUTHOR]

We use previously obtained experimental results by neutron interferometry to effectively constrain the parameter space of several prominent dark energy models. This investigation encompasses the environment-dependent dilaton field, a compelling contender for dark energy that emerges naturally within the strong coupling limit of string theory, alongside symmetron and chameleon fields. Our study presents substantial improvements over previous constraints of the dilaton and symmetron fields, improving parameter constraints by several orders of magnitude. However, the analysis does not yield any new constraints on the chameleon field. Furthermore, we establish constraints for the projected neutron split interferometer, which has recently concluded a decisive proof-of-principle demonstration. Our symmetron simulations reveal that, depending on the parameter values, there are multiple static solutions with an increasing number of nodes and increasing energy inside a cylindrical vacuum chamber. This agrees with results obtained earlier in the literature for infinitely extended parallel plates. Interestingly, while these multiple solutions can correspond to domain walls forming inside the vacuum chamber, we also find solutions that do not reach their vacuum expectation value inside the vacuum chamber, but display multiple nodes nonetheless. [ABSTRACT FROM AUTHOR]

We present detailed characterization of laser-driven fusion and neutron production ($\sim {10}^5$ /second) using 8 mJ, 40 fs laser pulses on a thin (<1 μm) D ${}_2$ O liquid sheet employing a measurement suite. At relativistic intensity ($\sim 5\times {10}^{18}$ W/cm ${}^2$) and high repetition rate (1 kHz), the system produces deuterium–deuterium (D-D) fusion, allowing for consistent neutron generation. Evidence of D-D fusion neutron production is verified by a measurement suite with three independent detection systems: an EJ-309 organic scintillator with pulse-shape discrimination, a ${}^3\mathrm{He}$ proportional counter and a set of 36 bubble detectors. Time-of-flight analysis of the scintillator data shows the energy of the produced neutrons to be consistent with 2.45 MeV. Particle-in-cell simulations using the WarpX code support significant neutron production from D-D fusion events in the laser–target interaction region. This high-repetition-rate laser-driven neutron source could provide a low-cost, on-demand test bed for radiation hardening and imaging applications. [ABSTRACT FROM AUTHOR]

Radiation hardening is studied for stainless austenitic and ferritic-martensitic chromium steels after ion and neutron irradiation at various temperatures. Austenitic and ferritic-martensitic steels irradiated up to 30 dpa in various nuclear reactors and ion accelerators are studied at various temperatures. A change in Vickers microhardness is used as the radiation hardening parameter. A methodology is developed that allows one to determine the ion irradiation parameters, which ensure the radiation hardening of ferritic-martensitic and austenitic steels, as close as possible to the radiation hardening of the same steels under neutron irradiation. A transferability function is introduced to connect the irradiation temperatures for ion and neutron irradiation that provides the same radiation hardening. On the basis of the obtained experimental data, after ion and neutron irradiation the transferability functions are determined for the investigated austenitic and ferritic-martensitic steels, which connect the temperatures for ion and neutron irradiation and provide the same radiation hardening at a given damage dose. [ABSTRACT FROM AUTHOR]

The neutron capture cross section of 181 Ta is relevant to s-process of nuclear astrophysics, extraterrestrial samples analysis in planetary geology and new generation nuclear energy system design. The 181 Ta( n , γ ) cross section had been measured between 1 eV and 800 keV at the back-streaming white neutron facility (Back-n) of China spallation neutron source(CSNS) using the time-of-flight (TOF) technique and C 6 D 6 liquid scintillator detectors. The experimental results are compared with the data of several evaluated libraries and previous experiments in the resolved and unresolved resonance region. Resonance parameters are extracted using the R-Matrix code SAMMY in the 1–700 eV region. The astrophysical Maxwell average cross section(MACS) from kT = 5 to 100 keV is calculated over a sufficiently wide range of neutron energies. For the characteristic thermal energy of an astrophysical site, at kT = 30keV the MACS value of 181 Ta is 834 ± 75 mb, which shows an obvious discrepancy with the Karlsruhe Astrophysical Database of Nucleosynthesis in Stars (KADoNiS) recommended value 766 ± 15 mb. The new measurements strongly constrain the MACS of 181 Ta( n , γ ) reaction in the stellar s-process temperatures. [ABSTRACT FROM AUTHOR]

The work describes features of the procedure for obtaining a response matrix to neutron radiation with energies from 1 to 19 MeV for a single-crystal stilbene scintillation detector. This response matrix is subsequently used in solving the problem of unfolding the neutron energy distribution at the detector location. The procedure for obtaining the matrix includes both experimental work and simulations of nuclear physics processes. In order to carry out the simulations, the Geant4 library package with the connected optical physics accounting module was used. The resulting matrix can be used as part of the amplitude method for unfolding the spectra of neutrons with energies in the range of 1–19 MeV with a step of 0.1 MeV. The obtained response matrix takes into account the main physical processes and instrumental effects of detectors based on organic scintillators, including the dependence of the light yield in the scintillator on the type and energy of charged particles, as well as the energy resolution of the detector depending on the energy of detected particles. In comparison with earlier results obtained using response matrices that do not take into account the above formation features of hardware spectra, it was possible to reduce the lower limit of the detected neutron energy from 1.5 to 1.0 MeV, as well as to increase the reliability of unfolded neutron energy distributions at the measurement point. The error in determining the neutron energy in the range of 1–15 MeV was not more than 200 keV. [ABSTRACT FROM AUTHOR]

The impact of 0.1–10 MeV neutron-induced single-event upsets (NSEUs) on terrestrial SER (soft error rate) prediction is analyzed for several broad spectrum neutron sources based on 0.1–200 MeV NSEU energetic-dependent cross-sectional data of a 40 nm SRAM. The results show that for test facilities with abundant <10 MeV neutrons, such as ISIS ChipIR, J-PARC BL10, CSNS Back-n, and CSNS BL09, the terrestrial SER is overestimated. The overestimation could reach 78%–287%. By reducing the cutoff energy (Emin) from 10 to 6 MeV, the overestimation can be decreased. However, the overestimation can still reach up to 164% in this case, significantly larger than the results reported by previous studies. By analyzing the contribution of neutrons in different energy bins to NSEUs, the discrepancy is proved to be attributed to the sensitivity of NSEUs to 0.1–1 MeV neutrons, along with the overpopulation of 0.1–1 MeV neutrons for certain test facilities, such as ISIS ChipIR and CSNS BL09. Therefore, it is necessary to take 0.1–1 MeV NSEU sensitivity into account when studying the prediction method using broad spectrum neutron sources. [ABSTRACT FROM AUTHOR]

This paper proposes a data-driven method to solve the fixed-energy inverse scattering problem for radially symmetric potentials using radial basis function (RBF) neural networks in an open-loop control system. The method estimates the scattering potentials in the Fourier domain by training an appropriate number of RBF networks, while the control step is carried out in the coordinate space by using the measured phase shifts as control parameters. The system is trained by both finite and singular input potentials, and is capable of modeling a great variety of scattering events. The method is applied to neutron–α scattering at 10 MeV incident neutron energy, where the underlying central part of the potential is estimated by using the measured l  = 0, 1, 2 phase shifts as inputs. The obtained potential is physically sensible and the recalculated phase shifts are within a few percent relative error. [ABSTRACT FROM AUTHOR]

The neutron's lifetime is a critical parameter in the standard model. Its measurements, particularly measurements using both beamline and ultracold neutron storage techniques, have revealed significant tension. In this work, we review the status of the tension between various measurements, especially in light of the insights provided by the β -decay correlation measurements. We revisit the lifetime measurement in a material storage chamber, dominated by losses from scattering off the walls of the storage chamber. The neutron energy spectra and associated uncertainties were, for the first time, well-characterized using storage data alone. Such models have applications in the extraction of the mean time between wall bounces, which is a key parameter for neutron storage disappearance experiments in search of neutron oscillation. A comparison between the loss model and the number of neutrons stored in a single chamber allowed us to extract a neutron lifetime of τ n * = 880 (+ 158 / − 78) stat. (+ 230 / − 114) sys. s   (68.3 %   C. I.) . Though the uncertainty of this lifetime is not competitive with currently available measurements, the highlight of this work is that we precisely identified the systematic sources of uncertainty that contribute to the neutron lifetime measurements in material storage bottles, namely from the uncertainty in the energy spectra, as well as from the storage chamber surface parameters of the Fermi potential and loss per bounce. In doing so, we highlight the underestimation of the uncertainties in the previous Monte Carlo simulations of experiments using the technique of ultracold neutron storage in material bottles. [ABSTRACT FROM AUTHOR]

Background: In a high-energy medical linear accelerator (linac), if the interaction of photon energy is higher than the neutron binding energy of high atomic material, it emits a neutron field through photonuclear (γ, n) reaction. Aim: The current study, evaluates the photoneutron dose equivalent (PNDE) produced between the 10 MV flattened and unflattened beams as a function of field sizes in the Elekta Versa HDTM linac. Materials and Methods: The PNDE produced from Versa linac was recorded along the patient plane using the bubble detector personal neutron dosimeter and from the measured PNDE values, the theoretical PNDE values were simulated for various field sizes using nonlinear least-squares curve-fitting as a function of a polynomial. The percentage of deviation (PoD) and Chi-square (χ2) tests were performed between the measured and simulated PNDE values to study the reliability and validity. Results: The results show that the mean PoD between the measured and simulated PNDE values for respective positions of a field size of FF beam was found to be -1.99% for 0.3×0.3, -4.39% for 5×5, -3.868% for 10×10, 0.590% for 15×15, 9.18% for 20×20, -4.133% for 25×25, and 0.467% for 30×30 cm2. Similarly, the mean PoD between the measured and simulated PNDE values for flattening filter-free (FFF) beam was found to be 1.36% for 0.3×0.3, -1.39% for 5×5, -5.38% for 10×10, 4.41% for 15×15, 3.84% for 20×20, 5.69% for 25×25, and -1.75% for 30×30 cm2. The maximum deviation between the measured and simulated PNDE values lies within the range ± 5%. Conclusions: From the study, it is observed that the FFF beam produces lesser neutron contamination than the FF beam. [ABSTRACT FROM AUTHOR]