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During fission an atomic nucleus generally splits into two fission products and between two and three neutrons. These neutrons are extremely energetic having energies in the low MeV range. These neutrons lose energy through scattering collisions, and if they are not captured by a nucleus will eventually reach thermal equilibrium at energies of a fraction of an eV. It is found that light nuclei such as hydrogen and beryllium are the most effective at slowing-down neutrons. During this slowing-down process, the neutron population will take on a spectrum of energies which will depend upon other things, the amount of fissionable material, the probability of the neutrons being scattered as opposed to being absorbed, and so on.

Journal of radioanalytical and nuclear chemistry (2021). doi:10.1007/s10967-021-08102-2, Published by Springer Science + Business Media B.V., Dordrecht [u.a.]

The (n,2n) nuclear reaction is important for neutron irradiation damage. However, its production of two indistinguishable secondary neutrons leads the calculation of displacement damage cross section to be complicated and uncertain. The present work investigates the (n,2n) reaction-induced displacement damage cross section by comparing five methods. The five approaches give consistent damage cross sections within 5% relative deviation for incident neutron energy below 20 MeV, the upper limit for nuclear reactor applications. However, the deviation becomes important for higher incident energy. The proposed KLab2 or KCM2 method is recommended in the case of non-negligible deviation between different calculations, while further investigation on the full correlation between two secondary neutrons as well as the distributions of excited recoils is suggested.

Boron neutron capture therapy (BNCT) was first proposed as early as 1936, and research on BNCT has progressed relatively slowly but steadily. BNCT is a potentially useful tool for cancer treatment that selectively damages cancer cells while sparing normal tissue. BNCT is based on the nuclear reaction that occurs when 10B capture low-energy thermal neutrons to yield high-linear energy transfer (LET) α particles and recoiling 7Li nuclei. A large number of 10B atoms have to be localized within the tumor cells for BNCT to be effective, and an adequate number of thermal neutrons need to be absorbed by the 10B atoms to generate lethal 10B (n, α)7Li reactions. Effective boron neutron capture therapy cannot be achieved without appropriate boron carriers. Improvement in boron delivery and the development of the best dosing paradigms for both boronophenylalanine (BPA) and sodium borocaptate (BSH) are of major importance, yet these still have not been optimized. Here, we present a review of this treatment modality from the perspectives of radiation oncology, biology, and physics. This manuscript provides a brief introduction of the mechanism of cancer-cell-selective killing by BNCT, radiobiological factors, and progress in the development of boron carriers and neutron sources as well as the results of clinical study.

The possibility of reconstructing the neutron spectrum of a photoneutron source using activation detectors is considered. Activating reactions suitable for reconstructing neutron spectra are selected. Reconstruction is done using the LOUHI program from the Nuclear Energy Agency (NEA) program bank. The optimum weight parameters of the terms in the minimized functional are selected according to data obtained with the model neutron spectrum.

The feasibility of monitoring maximum energy of neutrons $$E_{{\text{n}}}^{{\max }}$$ in the neutron flux from a source by analyzing pulse-height spectra from a 10B detector is investigated. Ionization losses of 4He and 7Li nuclei in two detector gaps are modeled. The clear experimental dependence of the position of the maximum and the shape of these spectra on $$E_{{\text{n}}}^{{{\text{max}}}}$$ and calculations in agreement with this dependence can be used to monitor the maximum energy of neutrons.

One of the methods that widely used in solving neutron transport equations in the nuclear fuel cell is the collision probability (CP) method. The neutron transport is very important to solve because the neutron distribution is related to the reactor power distribution. The important thing in the CP method is the CP matrix calculation, better known as has an important role in determining the neutron flux distribution in the reactor core. This study uses a linear flat flux model in each cell region for each energy group with white boundary condition. Although the type of reactor used in this study is a fast reactor, the matrix calculation still carried out in fast and thermal group energy. The matrix depends on the number of mesh in each cell region. The matrix formed from the mesh distribution will produce a matrix for each energy group. Because the boundary condition of the system is assumed that there are no contributions neutron source from the outside, the sum of the matrix must be less than one. In general, the results of the calculations in this study are following the theory

It is well known that prompt neutrons from nuclear fissions show anisotropic emission properties that have been studied for nuclear safeguards purposes. However, for further practical applications, thorough investigation of those properties from diverse perspectives needs to be done. In the present work, the energy dependency of prompt neutron anisotropic emissions by 244Cm and 240Pu was investigated using Monte Carlo simulations. In the results, the asymmetry of prompt neutron angular distributions showed a dependency on neutron energy thresholds, and its characteristic tendency was confirmed for both nuclides.

Negative and positive muon-induced single-event upsets (SEUs) are studied in 20-nm bulk planar SRAMs. Muon irradiation is performed using a mono-energetic source with varying the muon energy. The energy dependence of the cross sections (CSs) of SEUs and multiple-cell upsets (MCUs) shows the significant contribution of muon capture reactions for the negative muon, as reported in previous studies. Interestingly, MCU events are found for the positive muon, in contrast to the previous studies. The CSs for the negative and positive muons are compared with that for the other terrestrial radiations: high-energy neutrons, thermal neutrons, and alpha particles. The voltage dependence of the SEU CS, together with the empirical model for charge collection, demonstrates the difference in the contributing secondary ions among the negative muon, the high-energy neutron, and the thermal neutron. The MCU events are thoroughly analyzed in terms of their ratio to the total events and their fail bit patterns. The results reveal that the MCU characteristics for the negative muon are different from that for the other terrestrial radiations due to the muon capture reactions, where parasitic bipolar effects and the isotropic emission of secondary ions are important factors.

Characterizing low-energy neutrons ( et al. (2020), this article newly presents and analyzes the data measured at the low operating voltage of 0.4 V. The dependence of SEU cross section on the neutron energy is similar between the operating voltages of 0.4 and 1.0 V, including onset energy of around 6 MeV. The existence of MCUs at 4.1-MeV neutrons was also confirmed at both the operating voltages. Based on the measurement, we approximate the dependence of SEU and MCU cross sections as Weibull functions of the neutron energy. The terrestrial SER of SEUs and MCUs was calculated by folding the Weibull function and the flux spectrum. The calculated result indicates that the SER originating from the low-energy neutrons is less than 6% in the terrestrial environment at New York and Tokyo City. We confirm that disregarding the flux of neutrons below 10 MeV in the acceleration factor calculation at accelerated neutron tests, which follows the current standard defined in JESD89, could give a reasonable SER estimation accuracy for both SEUs and MCUs. On the other hand, for covering the beams having an extremely high proportion of low-energy neutrons, considering the flux of neutrons above 6 MeV would be an option for better SER estimation.

In the future, a hazardous asteroid will find itself on a collision course with Earth. For asteroids of moderate size or larger, a nuclear device is one of humanity’s only technologies capable of mitigating this threat via deflection on a timescale of less than a decade. This work examined how the output neutron energy from a nuclear device standoff detonation affects the deflection of a notional asteroid that is 300 meters in diameter and composed of silicon dioxide at a bulk density of 1.855 g/cm 3 . 14.1 MeV and 1 MeV neutron energy sources were modeled in MCNP to quantify the energy deposition in the asteroid target. The asteroid’s irradiated region was discretized in angle by tracing the rays emanating from the point of detonation and in depth by considering the neutron mean-free-paths. This high-fidelity approach was shown to deviate from previous analytic approximations commonly used for asteroid energy deposition. 50 kt and 1 Mt neutron yields of the energy deposition mappings were imported into a hydrodynamic asteroid model in ALE3D to simulate the deflective response due to blow-off ejecta. Underexplored in literature, changing the neutron energy was found to have up to a 70% impact on deflection performance due to induced differences in the energy deposition profile and in the energy coupling efficiency. The magnitude of energy deposition accounted for most of the observed variation in the asteroid velocity change, making the coupling efficiency more significant than the spatial profile characteristics. These findings are vital for determining the optimal source neutron energy spectrum for asteroid deflection applications.

A virtual mechanism used for describing low-energy ternary fission is generalized to spontaneous and induced quaternary fission. A formula is obtained for the width of quaternary fission as a virtual process. The height and penetrability of the Coulomb barrier are estimated for the spontaneous fission of 248Cm and 252Cf nuclei and the thermal neutron induced fission of 233U and 235U nuclei with the emission of pairs of α particles.

A two-coordinate 10B detector is used to study the spatial distribution of a neutron flux with maximum energies $$E_{{\text{n}}}^{{{\text{max}}}}$$ = 3.3–5.1 MeV from the outlet channel of a photoneutron source. The maximum of neutron flux intensity measured along the horizontal axis is found to shift backward, relative to the direction of the primary electron beam upon an increase in maximum neutron energy $$E_{{\text{n}}}^{{{\text{max}}}}$$ .

Fast neutrons and gamma-ray imaging detection is an effective way to detect and identify radioactive material in the field of nuclear security. A compact coded aperture imaging (CAI) camera was designed to be sensitive to both gamma and neutron radiation based on plastic scintillators and multi-pixel photon counters (MPPC). MPPCs coupling with the 13 × 13 pixelated plastic scintillators one-to-one were utilized to reduce the scale of the CAI system while maintaining good positional performance. The symmetric charge division (SCD) circuit was adopted to reduce the 169 signals output from the MPPC array to 26. Each waveform was collected and processed with four Domino Ring Sampler 4 (DRS4) chips and two 16-channel analog-to-digital converter (ADC) modules. As the pulse shapes of fast neutrons would be broadened after elastic scattering multiple times in the scintillators, the Anger-Logic method was applied to eliminate multiple elastic scattering events so that good pulse shape discrimination (PSD) performance can be achieved. The imaging and detection ability of the camera was evaluated using the 241Am-Be (5.9 × 105 n/s) neutron source and 137Cs (370 MBq) gamma source. The camera can be used to detect fast neutrons (0.5–10 MeV) and gamma rays (0.2–2.5 MeV). Furthermore, it can implement efficient neutron/gamma PSD capabilities in the mixed-field environment. The figure of merit (FOM) of the camera calculated at 400keVee energy cut is 0.93. A compact MPPC-based CAI camera was designed to detect and discriminate fast neutrons and gamma rays. Its good PSD performance was well suited to distinguish fast neutrons from gamma rays in a dual-particle environment. The portable design makes it promising for complex monitoring scenarios in nuclear security.

The neutron capture cross sections of 244Cm and 246Cm were measured for the neutron energy range of 1–1000 eV via the neutron time-of-flight method with Accurate Neutron-Nucleus Reaction measurem...

Active interrogation (AI) is a promising technique to detect shielded special nuclear materials (SNMs). At the University of Michigan, we are developing a photon-based AI system that uses bremsstrahlung radiation from an electron linear accelerator (linac) as an ionizing source and trans-stilbene organic scintillating detectors for neutron detection. Stilbene scintillators are sensitive to fast neutrons and photons and have excellent pulse shape discrimination (PSD) capabilities. The traditional charge integration (CI) method commonly used for PSD analysis eliminates piled-up pulses and relies on a particle discrimination line to separate neutrons and photons. The presence of the intense photon flux during AI creates a significant number of piled-up events in the stilbene scintillator, thereby posing a great challenge to the traditional CI method. Identifying true single neutron pulses becomes challenging due to the presence of a pile-up cloud and overlapping neutron, photon and pile-up clouds in the PSD analysis. To mitigate the effect of pulse pile-up and identify true single neutron pulses from stilbene scintillators, an artificial neural network (ANN) system is developed. The developed ANN system identifies single neutron pulses and neutron-photon combinations from piled-up events. The results obtained from a 252Cf measurement in the presence of the intense photon flux show that the developed ANN system outperforms the traditional CI method. Since many piled-up events lie above the particle discrimination line, they get misclassified as neutrons by the traditional CI method resulting in 27% overestimation of the net neutron count rate during the linac pulse. The overall net neutron count rate (single and restored neutrons) during the linac pulse, estimated by the ANN system is 62.32% of the ground truth. Energy spectroscopy of the ANN attributed single neutron pulses further provides evidence on the detection of prompt fission neutrons from the 252Cf fission source.

The authors review recent experimental data about relative biological effectiveness (RBE) of low absorbed doses (0.01–0.5 Gy) from fast neutrons and accelerated multicharged ions currently accepted as career limits for cosmonauts from the standpoint of late effects risk mitigation. According to the new radiobiological findings, literary maximum values of the RBEM coefficients for low doses and dose rates are increased two or three times for late cytogenetic effects; benign cell transformation and tumor risks; lifetime reduction; disorders in the brain cortex neurons and damages to organs; lenticular opacity and cataract. The analysis-based revision of the quality coefficients – linear energy transfer (LET) relationship leads to a considerable increase of equivalent doses received from galactic cosmic rays and fast neutrons, and calculated total radiation risk over cosmonaut's lifetime.

The cross-sections of the 187Re(n,γ)188mRe and 187Re(n,γ)188gRe reactions were measured at the four neutron energies of 13.5, 14.1, 14.4 and 14.8 MeV through the activation method and using off-line γ-ray spectrometry. The previously developed formulas were used to subtract the effects of γ rays with close energies and the contribution from the excited state. A high-resolution γ-ray spectrometer with high-purity germanium detector was used to measure the induced γ activities. The simulation computations of the excitation functions concerning the above-mentioned two nuclear reactions were performed by adopting the nuclear theoretical model program system Talys-1.9. The theoretical excitation curves in the incident neutron energy range from 2.53 × 10−8 to 20 MeV were acquired. The fitting line of our results in the neutron energy range of 13.5–14.8 MeV, within experimental uncertainty, agrees very well with the theoretical excitation curve based on the computer code Talys-1.9.

The method of Lebesgue moments for simulating the reversal of resonances, resonance self-shielding, and block effect in the neutron spectra of extended heterogeneous objects, such as nuclear reactors, radiation shielding, and installations for studying the properties of matter, is developed. The method uses a more accurate averaging procedure over neutron energy than the group averaging. The main components of the method are the refinement of the resonance structure of neutron cross sections by dividing the energy scale into a series of sets called carriers of resonances, Lebesgue averaging of cross sections and neutron flux within carriers, and the expansion of the neutron flux in a series in basis functions that depend on the magnitude of the neutron cross sections. The expansion coefficients (the so-called Lebesgue moments) can be calculated by any available method for solving the neutron transport equation.

The authors review the available data on damage to cortical neurons and incipient lenticular opacity in small laboratory animals after exposure to fast neutrons, accelerated multi-charged ions (AMCIs), and different doses of X- and γ-rays. Analysis and possible revision of maximum coefficients of relative biological effectiveness (RBEM) for these effects, as well as RBEM coefficients for other space radiation-induced impacts, including blast transformation, cytogenetic alterations, increased probability of tumor development in organs and tissues, and reduction in estimated life expectancy, will lay the grounds for reconsideration of the existing normative documents and calculations of generalized and equivalent doses for organs and tissues based on the absorbed dose values. At present, Russian GOST 25 645.218-90 “Space Crew Radiation Safety during Space Fight” and recommendations of the International Commission on Radiological Protection (ICRP, papers 60 and 61, 1994) reflect the dependence of space radiation quality factors (QFs) on linear energy transfer.

The near-threshold fissioning of 234U in the 0.2–0.9 MeV range of neutron energy displays resonance-like behavior with sharp variation of the angular distribution of fission fragments. The behavior of the fission cross section and angular distribution of fragments is attributed to resonance-like vibrational excitations of class III in the second isomeric potential well. The fission cross section and angular distribution of fragments are calculated with the Hauser–Feshbach model using the TALYS code. The calculated fission cross sections and the anisotropy of fragments are compared to experimental findings. Quantum numbers are obtained for the vibrational excitation of a highly deformed nucleus.

Sources of fusion neutrons with an energy of about 10 MeV can be a driver in hybrid fusion–fission reactor. They can be used for the disposal of radioactive wastes involved in the closure of the nuclear fuel cycle. Currently, the projects of such systems rely on the use of the D–T reaction and the production of tritium in the blanket. In terms of availability of fuel components, the D–D reaction is attractive. The energy of the neutrons produced in the D–D reaction directly is not high enough, but fast neutrons of 14 MeV are produced by the burning of the tritium produced in the D–D reactions. The possibilities of using a D–D plasma confined in a magnetic trap to generate fast neutrons are analyzed. To increase the reaction rate, a powerful heating by injection of neutral deuterium atoms is considered. Under such conditions, a significant population of fast deuterons is maintained. The requirements on the parameters of the plasma and the magnetic trap are discussed to present the possible concept of a fusion neutron source based on deuterium plasma.

The paper considers the effects produced by arranging a lead cavity in the central part of a lead-cooled fast reactor core. It is shown that the most favorable effect from the presence of an inner lead cavity at the center of the core can be obtained when radiogenic lead with a dominant 208Pb isotope content is used as the material of the cavity, coolant, and neutron reflector. This is explained by the extremely low neutron absorption of 208Pb in a wide range of neutron energies. If the thickness of the 208Pb inner cavity is chosen properly, then a spectral region with the prevailing share of resonance, epithermal, and thermal neutrons is created in the cavity. This results in a sharp increase in the mean prompt neutron lifetime (by few orders of magnitude) and enhancement of the stabilizing Doppler effect. The neutron migration from the inner cavity to the reactor core is able to intensify the fission chain reaction (FCR) and contribute positively to reactor reactivity. In addition, the high flux of slowed neutrons in the inner 208Pb-cavity makes it possible to expect efficient transmutation of long-lived fission products.

In this contribution we describe NEPIR, the fast-neutron irradiation facility under construction at the 70 MeV cyclotron SPES facility of the INFN laboratory of Legnaro (LNL). NEPIR will be constructed in stages, according to the available funds. The initial configuration, based on a thick Be neutron production target, will be operational in 2022; it will be used for shielding studies against fast neutrons for space applications and to investigate neutron-induced Single Event Effects (SEE) in microelectronic devices and systems. In its final configuration NEPIR will have two target systems: one will deliver a Quasi Mono-energetic Neutron (QMN) beam, of general interest, with an adjustable energy peak in the 20–70 MeV range; the second target will deliver a specialized continuous energy neutron beam for studying the effects of fast neutrons produced in cosmic ray air-showers in electronic devices and systems. We review the use of NEPIR to characterize the sensitivity of electronics, describe the neutron production targets and the facility layout. In closing we describe ways, presently under investigation, to use the 15 MV XTU Tandem of LNL to produce nearly monochromatic fast neutrons that would complement the QMN system by allowing one to probe for SEE below 20 MeV.

The 92Mo(n,α)89Zr, 97Mo(n,p)97Nb, 148Nd(n,2n)147Nd and 146Nd(n,p)146Pr reaction cross sections with the neutron of energy of 14.54 ± 0.24 MeV have been measured by using the method of activation and off-line γ-ray spectrometry. The neutron was generated from the D-T reaction at the PURNIMA neutron generator of BARC. The 197Au(n,2n)196Au monitor reaction was used for the estimation of neutron flux. The covariance analysis for the uncertainties of the cross sections has been carried out by considering the partial uncertainties of different attributes. The present data have been compared with the literature data, evaluated data and theoretically calculated values from TALYS-1.9 code.

International audience; During the flight, an aircraft is submitted to a radiation environment composed of cosmic-ray-induced particles (CRIP) of which neutrons are responsible for approximately 40% of the crew effective dose and are the main cause of single event effects (SEE) in avionics systems at flight altitudes. A model of Learjet aircraft was developed on Monte Carlo simulation using the MCNPX code in order to detail the CRIP field inside the aircraft. The radiation source modeling was previously developed by a computational platform that simulates the energy and angular distributions of the CRIP along the atmosphere. In this article, we determined the variation of the neutron radiation field in several positions inside the aircraft at 11- and 18-km altitudes and for both equatorial and polar regions. The results suggest that the maximum variation of neutron fluence rate between different positions inside the aircraft shows a tendency of higher differences for a lower energy threshold (thermal and E > 1 MeV) in comparison with those differences for a higher energy threshold (E > 10 MeV). Moreover, the angular distribution results show relevant differences between positions inside aircraft, mainly for thermal neutrons close to the fuel. The general tendency is to enhance these discrepancies for devices with new technologies, due to their lower energy threshold for SEE occurrences.

The Neutron Detector facility for detecting the neutral component of cosmic rays has been developed by the Dolgoprudny Scientific Station of the Lebedev Physical Institute. The description and performance characteristics of the facility are presented. Each module of the facility is based on SNM-18 counters. Unlike standard neutron monitors, this facility is sensitive to thermal neutrons. The integration time of the counters is 1 ms. The function of continuous monitoring of meteorological data and data on the state of the supply voltages of the facility has been created. The results of the barometric-coefficient determination for the Neutron Detector facility and the analysis of the first experimental data are also presented.

This paper details and implements a framework for evaluating thermal neutron scattering cross sections that provide S ( α , β ) data and covariance data for hydrogen in light water. This methodolog...

Neutron-based experimental techniques have been continuously improved, refined and strengthened since the pioneering experiments conducted by Clifford Shull and Bertram Brockhouse in the mid twentieth century. The possibility to reveal structure and dynamics at different scales of distances and times, provided a deep insight into the microscopic nature of condensed matter systems. The advent of scattering techniques, firmly linked to instrument development (diffractometers, spectrometers), made neutron science attractive for scientists working in different fields, such as physics, biology, chemistry and engineering. Together with the development of intense sources and sophisticated instruments, the role of neutron detection techniques is instrumental for an effective use of the intense fluxes of neutron beams that became available in the last three decades. Detectors are then essential for the development of new and effective instrumentation that in turn can trigger new ideas for science. Neutrons made available at large scale facilities extend from ultra-cold to fast neutrons. Sources providing monochromatic fast neutron beams, such as DD or DT sources (also in the form of portable devices) are used for many applications, including at industrial level. Thus, the unique properties of neutrons in terms of their interaction with matter are related to the extended range of energies or (equivalently) wavelengths over which they can be produced at both compact and/or large scales facilities. The scope of this review is, starting from the main physical mechanism for neutron detection, to provide a survey on well assessed and newly developed neutron detection systems using both passive and active methods and their applications. It will provide an overview of the current state of neutron detection by describing different approaches and pointing out open problems to be faced.

At the Korea Multi-purpose Accelerator Complex (KOMAC), a 100-MeV high-power proton linac has been used for proton-beam irradiation services for various applications and research programs, including material science, bio-medical science, semiconductor applications, as well as nuclear physics and basic science, since commissioning of the machine in 2013. Recently, demands for fast neutrons from users in the semiconductor industries are increasing to test the semiconductor device SER (soft error rate) or SEE (single event effect). At present, we provide pilot services to users with pulsed white-spectrum neutrons (max. neutron energy ∼100 MeV) generated at the beam dump, which is made of copper. However, the neutron energy should be increased at least to 200 MeV to meet international standards such as ISO26262 or JEDEC89A. For the upgrade of the linac energy to 200 MeV, we conducted a preliminary design study, including superconducting radio-frequency (SRF) based linac structure, a neutron generation target and a neutron utilization facility. The upgrade plan for the KOMAC linac, along with the preliminary results of the design study on the superconducting linac and neutron generation facility, will be given in this paper.

To measure the total neutron cross sections, the authors developed a convenient technique using silicon semiconductor detectors with the simple and compact detecting system. The created technique was used to measure A + n total cross sections at the energy En = 14.1 MeV for several nuclei in different mass region. The neutron generator NG-150 in D + T regime of neutrons generation is used as fast neutrons source. The results of measurements and their comparison with the appropriate literature values are also presented.

Distributions of thermal neutron flux density are measured with respect to the center of a W–Be photoneutron source and the axis of an output neutron channel. The measurements are made using activation detectors. The distributions are approximated by simple functions with parameters that depend on the detectors’ coordinates. The spatial distribution is obtained for thermal neutrons in the horizontal plane relative to the axis of the photoneutron source’s output channel.

The thermal neutron scattering cross-section of a solid depends on the energy (or wavelength) of the incident neutrons. Devising a method to calculate the energy dependence from first principles, without the approximations built in the scattering theory, has been a major undertaking in nuclear engineering. Here, we demonstrate such a calculation method using the program OCLIMAX. Our approach eliminates various approximations and limitations involved in a regular calculation with the LEAPR module of NJOY code, and the results are compared with available experimental and theoretical data. It is also demonstrated how additional insight can be obtained from the calculated full dynamical structure factor. The results reported here show the great potential and excellent platform provided by OCLIMAX for future development in the study of neutron thermalization in solid materials for different applications.

International audience; In addition to high-energy hadrons, which include neutrons, protons, and pions above 20 MeV, thermal neutrons (ThNs) are a major concern in terms of soft error rate (SER) for electronics operating in the large hadron collider (LHC) accelerator at the European Organization for Nuclear Research (CERN). Most of the electronic devices still contain Boron-10 inside their structure, which makes them sensitive to ThNs. The LHC radiation environment in different tunnel and shielded areas is analyzed through measurements and FLUKA simulations, showing that the ThN fluence can be considerably higher than the high-energy one, up to a factor of 50. State-of-the-art commercial-off-the-shelf (COTS) components such as SRAM, field-programmable gate arrays (FPGA), and Flash memories of different technologies are studied to derive the expected single-event upset (SEU) rate due to ThNs, relative to the high-energy hadron contribution. We find that for the studied parts and most of the accelerator applications, ThNs are the dominating source of upsets with respect to the high energy particles yielding even to neglect the latter in some cases. Indeed, they can induce, in electronics, up to more than 90% of the total upsets. The estimation is performed also for ground-level and avionic applications, and although in general, ThNs are not the main source of SER, in Flash memories they can play the same role as high energy neutrons. Related radiation hardness assurance (RHA) considerations for the qualification of components and systems against ThNs are presented.

The propagation of thermal neutrons through media characterized by different degrees of ordering is analyzed. It is shown that the standard Fermi pseudopotential concept should be modified in order to obtain a unified description of thermal-neutron propagation in crystalline and amorphous media. It is found that satisfactory results are obtained upon employing a pseudopotential that correctly reproduces the amplitude of scattering on a single center in the second order of perturbation theory rather in the first order. General formulas that describe thermal-neutron propagation through media of arbitrary degree of ordering are derived.

In neutron optics, we analyse the propagation of slow (thermal) neutrons along a semi-infinite waveguide limited by a large repulsive potential (reminding the propagation of light along an optical fibre). We set an ideal case of a straight empty two-dimensional semi-infinite waveguide limited by an infinitely repulsive potential (Dirichlet conditions on its boundaries). The neutron wave function is given through an integral representation involving the incoming wave with energyE, Green's function and certain functions defined on the boundaries. The latter functions follow from the Dirichlet conditions, thereby proposing a new formalism in neutron optics. We develop various approximations (through Fourier and Hilbert transforms) and numerical computations. We get: (a) the expected extinction of the incoming wave for very large penetration into the waveguide, (b) the generation of propagation modes and their number for suitably large penetration, asEincreases, (c) an estimate of the critical angle.

We present the results of accelerated radiation testing on an AMD accelerated processing unit, three Nvidia graphic processing units, an Intel accelerator, a field-programmable gate array, and two double-data-rate memories under thermal and high-energy neutrons separately. The sensitivity depends on the device type and the code being executed and we show that thermal neutrons contribute to the error rate of modern computing devices under certain conditions.

Fast neutrons can perform two functions that are critical to the nuclear power industry: breed fissile material from fertile material and burn minor actinides in spent fuel. The fusion reaction in a fusion reactor between deuterium and tritium nuclei yields an alpha particle and a neutron with 14.5-MeV energy. These neutrons can easily convert fertile material (such as 238U and 232Th) into fissile material (such as 239Pu and 233U) and destroy minor actinides (Pu, Np, Am, and Cm) present in the spent nuclear fuel and significantly reduce its radiotoxicity. However, each neutron generated implies one tritium consumed and tritium is generated by nuclear reactions between a neutron and lithium nuclei. Therefore, neutron economy becomes a crucial issue and an adequate neutronic modeling of the neutron source becomes critical. In this article, the Automatic System for TRansport Analysis (ASTRA) and Monte Carlo N-Particle (MCNP) codes are used in conjunction to characterize the neutron flux coming out of a low-aspect-ratio tokamak machine. ASTRA is first used to determine temperature and density profiles within the plasma volume and MCNP is used to evaluate the neutron flux. Since the simulation of an extended neutron source with complex geometry such as the confined plasma inside the tokamak is difficult, steps are taken to simplify the geometry by defining two simple sources, a cylindrical wall and a disk, which should have similar neutron emission distributions to the volumetric source. This reduction is done using a preprocessing Monte Carlo algorithm to establish probability density functions of neutron emission for the cylindrical wall as a function of the height and for the disk as a function of the radius, based on the emission characteristics of the extended source. The results that compare the neutron flux leaving a volume for four different configurations of neutron sources (isotropic source at the geometrical origin, ring source at the plasma geometrical center, cylindrical wall source, and distributed multipoint source) are presented. It is found that the processing of the two simple sources has a higher computational cost in terms of both processing time and random number consumption.

The 100Mo (n, 2n) 99Mo and 96Mo (n,p) 96Nb reactions have been studied relative to the 197Au (n, 2n) 196Au monitor reaction, at the neutron energy of 14.54 MeV, based on an experiment performed using Purnima neutron generator. Extended unscented transformation technique has been applied for computing the contribution of uncertainty of each attribute, which is then propagated, for the measurement and covariance analysis of the cross section of 100Mo (n, 2n) 99Mo and 96Mo (n, p) 96Nb reactions. The results obtained agree closely with the experimental data, model data and data in various evaluated nuclear data libraries.

The 110Pd(n,2n)109Pd, 102Pd(n,2n)101Pd, 105Pd(n,p)105Rh and 106Pd(n,p)106mRh reaction cross sections have been measured with respect to the 197Au(n,2n)196Au monitor reaction at the neutron energy of 14.54 ± 0.24 MeV by using the method of activation and off-line γ-ray spectrometry. The mono-energetic neutron beam was generated from the D–T reaction. The uncertainties in the various basic nuclear parameters in the measured reactions and their correlations were estimated by using covariance analysis. The present data were compared with the EXFOR based literature data, evaluated data of various libraries available in national nuclear data center and with the calculated values from TALYS-1.9 code.

Background High energy electron linear accelerators (LINACs) producing photon beams with energies higher than 10 MeV are widely used in radiation therapy. In these beams, fast neutrons are generated, which results in undesired contamination of the therapeutic beam. In this study, measurements and Monte Carlo (MC) simulations were used to obtain neutron spectra and dose equivalents in vicinity of linear accelerator. Materials and methods LINAC Siemens Oncor Expression in Osijek University Hospital is placed in vault that was previously used for 60Co machine. Then, the shielding of the vault was enhanced using lead and steel plates. Measurements of neutron dose equivalent around LINAC and the vault were done using CR-39 solid state nuclear track detectors. To compensate energy dependence of detectors, neutron energy spectra was calculated in measuring positions using MC simulations. Results The vault is a source of photoneutrons, but a vast majority of neutrons originates from accelerator head. Neutron spectra obtained from MC simulations show significant changes between the measuring positions. Annual neutron dose equivalent per year was estimated to be less than 324 μSv in the measuring points outside of the vault. Conclusions Since detectors used in this paper are very dependent on neutron energy, it is extremely important to know the neutron spectra in measuring points. Though, patient dosimetry should include neutrons, estimated annual neutron doses outside the vault were far below exposure limit of ionizing radiation for workers.

Fast-Neutron Multiplicity Counter based on Liquid Scintillator Detector can directly measure the fast neutron multiplicity emitted by UO2 rod. HPGe gamma spectrometer; which has superior energy resolution; is routinely used for the gamma energy spectrum measurement. Combing Fast-Neutron Multiplicity Counter and HPGe γ-spectrometer, the n/γ joint measurement system is developed. The fast neutron multiplicity and gamma energy spectrum of UO2 rod assemblies under different conditions are measured by the n/γ joint measurement system. The induced fission rate and the 235U abundance, thereby the 235U mass; can be obtained for UO2 rod assemblies. The 235U mass deviation of the measured value from the reference value is less than 5%. The results show that the n/γ joint measurement system is effective and applicable in the measurement of the 235U mass in samples. Keywords: Fast-neutron multiplicity counter, n/γ joint measurement system, Liquid scintillator detector, HPGe γ spectrometer, UO2 rod assembly, 235U mass

The purpose of the current study was to evaluate a spectrum formulation set employed to modify the neutron spectrum of D-D fusion neutrons in a IS plasma focus device using GEANT4, MCNPX2.6, and FLUKA codes. The set consists of a moderator, reflector, collimator and filters of fast neutron and gamma radiation, which placed on the path of 2.45 MeV neutron energy. The treated neutrons eliminate cancerous tissue with minimal damage to other healthy tissue in a method called neutron therapy. The system optimized for a total neutron yield of 109 (n/s). The numerical results indicate that the GEANT4 code for the cubic geometry in the Beam Shaping Assembly 3 (BSA3) is the best choice for the energy of epithermal neutrons. Keywords: IS plasma focus device, Beam-shaping assembly, Neutron therapy, GEANT4, MCNPX2.6, FLUKA

Neutron diffraction has been a very prominent tool to investigate high-temperature superconductors, in particular through the discovery of an incommensurate magnetic signal known as stripes. We here report the findings of a neutron diffraction experiment on the superconductor (La,Sr)$_2$CuO$_4$, where a spurious signal appeared to be magnetic stripes. The signal strength was found to be strongly dependent on the neutron energy, peaking at $E = 4.6$~meV. We therefore attribute the origin of this signal to be a combination of multiple scattering and crystal twinning. A forward calculation of the scattering intensity including these two effects almost completely recovers our experimental observations. We emphasise the need for employing such analysis when searching for ways to avoid spurious scattering signals., Comment: 8 pages, 5 figures

In the framework of the radioactive waste drum characterization using neutron coincidence counting, the Nuclear Measurement Laboratory of French Alternative Energies and Atomic Energy Commission (CEA) Cadarache is studying plastic scintillators as an alternative to ideal but costly 3He gas proportional counters. Plastic scintillators are at least five times cheaper for the same detection efficiency, and in addition, they detect fast neutrons three orders of magnitude more quickly than 3He detectors. However, they are sensitive to gamma rays with no pulse shape discrimination abilities for large detection volumes, which implies the necessity to identify precisely gamma background sources that may affect the useful signal. This article presents a detailed analysis of the gamma-ray spectrum of a radioactive waste drum containing glove box filters contaminated by plutonium dioxide. Classical gamma emissions following alpha and beta disintegrations are identified, and also those accompanying inelastic scattering (n, $\text{n}^{\prime }$ ) or radiative capture (n, $\gamma$ ) reactions in the whole waste drum, and ( $\alpha $ , n) or ( $\alpha $ , p) reactions in the filtration media, which can lead to neutron-gamma coincidences parasitizing useful coincidences from plutonium spontaneous fissions.