Incident Neutron Spectrum Research Articles

Development of a simple calculation tool of dose distributions in a phantom for boron neutron capture therapy

A simple dose calculation tool, SiDE, was developed for dose evaluation in a water phantom for boron neutron capture therapy, which makes the calculation time much shorter compared with the conventional particle transportation Monte Carlo codes and is applicable to any type of incident neutron spectra to the phantom. As the SiDE can not only calculate quantitatively the dose distribution in the phantom but also output dose indexes such as advantage depth and peak tumor dose, a comparison between different boron neutron capture therapy neutron sources can be easily performed. Consistency with a Monte Carlo transportation code was verified through comparison with the conventional dose calculation with the Particle and Heavy Ion Transport Code System, and the calculation time was nearly 1/90 in the SiDE. The dose distributions for a reactor and accelerator-based neutron sources were compared, and the differences were found to be small although large differences existed between the incident spectra.

Genetic algorithm-based optimization of a target for the production of atmospheric-like neutrons via 100 MeV proton beam

This paper presents the optimization of a white neutron target for 100 MeV proton beam to produce a fast neutron spectrum shape similar to that of the atmospheric neutron at ground level. The neutron target consists of 0.5 mm-thick discs consecutively placed in front of a cooling water channel. A sensitivity analysis is performed with different materials to assimilate resulting neutron spectrum with the ground-level cosmic ray-induced neutron spectrum by applying the genetic algorithm (GA). Optimization of the spectrum is conducted in consideration of neutron impact to semiconductor devices. To reconstruct a target-generated neutron spectrum, parametrization on neutron generation and stopping power by target material and position is configured. The loss function in the GA is defined to evaluate similarity between the target neutron spectrum and the atmospheric neutron spectrum. The GA is modeled with a package of consecutive discs as a gene for optimization, and the length and material of target for each disc are decided throughout generations. As a result, the target is configured with six different materials, with approximately half of the weighted deviation of atmospheric neutrons being compared with any other single-solid neutron target. The GA-based optimization process shown here suggests further use for designing targets with other incident proton energies or neutron spectra for various applications.

Unfolding neutron spectrum using Doppler broadened γ peak shapes

The Doppler broadening of gamma rays occurs in the fast neutron analysis with photons emitted in inelastic scattering reactions on certain nuclei, such as 12C and 14N. In this paper, we developed an algorithm based on the Monte Carlo method to calculate the Doppler broadened gamma peak shape of the 4.438 MeV gamma ray emitted by 12C in graphite when exposed to fast neutrons of different energies. The SAND-II unfolding iterative method is applied to obtain the spectrum of the incident neutron beam by using the Doppler broadened gamma spectra. In the calculation example, the unfolded neutron spectrum shows reasonable agreements with the original incident neutron spectrum.

Thermal neutron scattering properties of Bismuth crystal filter

ABSTRACT Accurate thermal neutron scattering cross-section data of the Bismuth (Bi) filter is required for high-fidelity calculations of the thermal neutron beam’s performance after the filtration in reactor column. In this paper, the phonon density of states of Bi was calculated using the ab initio method, based on which the thermal neutron scattering cross-section data was generated by the introduction of Bi coherent scattering model. By comparing the numerical results with the experimental data, the applicability of the model is verified. The model is then applied to simulate the neutron beam-shaping effect in the filter with an incident neutron spectrum in a research reactor column. Compared with the free gas model, the thermal neutron fraction was enhanced remarkably with an accurate phonon spectrum, and the phenomenon of Bi neutron filtration can be captured.

Measurements of cross sections for high energy neutron induced reactions on Co and Bi

There are few experimental data for neutron cross-section libraries in (n,xn) reactions for various materials at energies above 20 MeV. For neutron energies above 20 MeV, these set of (n,xn) reactions are important for neutron fluence monitoring and spectra unfolding for future generation IV nuclear reactors. There were attempts to measure the cross-sections on natural cobalt and bismuth at incident neutron energies of about 90 MeV and 140 MeV. These measurements were made using the quasi-monoenergetic neutron facility at iThemba LABS, South Africa. In addition, at The Svedberg Laboratory facility in Sweden, similar experiments were performed on natural Yttrium. The measured cross-sections were compared with some of the few available data for neutron-induced reactions at high energies. Data collected from these two facilities, required corrections to be made for the contribution of the low energy tail (continuum) in the incident neutron spectrum. Preliminary results from iThemba LABS showed large discrepancy which we suspect was due to instability of the proton beam during the irradiations. Follow-up experiments are planned to accurately determine the uncertainty contributions, with additional data at other neutron energies.

Unfolding Plastic Detector Responses to White and Mono-energetic Neutrons Based on Artificial Neural Networks

Neutron spectrum unfolding based on an artificial neural network (ANN) is a highly flexible and robust method that can be applied to any type of radiation detector and wide energy range of incident particles. In this study, we present details of neutron-spectrum deconvolution with well-established multilayer perceptron algorithms implemented in CERN ROOT with the aim of obtaining the incident neutron energy spectra for white neutron and mono-energetic beams with finite energy spreads. The ANN trained with experimental and simulation datasets successfully approximated incident neutron spectra with high accuracy for each case, indicating that a well-trained ANN has high potential for applications in radiation-related fields.

Single event upset on static random access memory devices due to spallation, reactor, and monoenergetic neutrons**Project supported by the National Natural Science Foundation of China (Grant Nos. 11690040 and 11690043) and the Foundation of State Key Laboratory of China (Grant Nos. SKLIPR1801Z and 6142802180304).

This paper presents new neutron-induced single event upset (SEU) data on the SRAM devices with the technology nodes from 40 nm to 500 nm due to spallation, reactor, and monoenergetic neutrons. The SEU effect is investigated as a function of incident neutron energy spectrum, technology node, byte pattern and neutron fluence rate. The experimental data show that the SEU effect mainly depends on the incident neutron spectrum and the technology node, and the SEU sensitivity induced by low-energy neutrons significantly increases with the technology downscaling. Monte–Carlo simulations of nuclear interactions with device architecture are utilized for comparing with the experimental results. This simulation approach allows us to investigate the key parameters of the SEU sensitivity, which are determined by the technology node and supply voltage. The simulation shows that the high-energy neutrons have more nuclear reaction channels to generate more secondary particles which lead to the significant enhancement of the SEU cross-sections, and thus revealing the physical mechanism for SEU sensitivity to the incident neutron spectrum.

Ionizing and non-ionizing kerma factors in silicon for China Spallation Neutron Source neutron spectrum

The quantification of ionizing energy deposition and non-ionizing energy deposition plays a critical role in precision neutron dosimetry and in the separation of the displacement damage effects and ionizing effects induced by neutron radiation on semiconductor devices. In this report, neutrons generated by the newly built China Spallation Neutron Source (CSNS) are simulated by Geant4 in semiconductor material silicon to calculate the ionizing and non-ionizing kerma factors. Furthermore, the integral method is applied to calculate neutron-induced ionizing at the CSNS and non-ionizing kerma factors according to the standard neutron nuclear database and the incident neutron spectrum. In addition, thermoluminescence dosimeters are utilized to measure the ionizing energy deposition and six series of bipolar junction transistors are used to measure the non-ionizing energy deposition based on their neutron damage constants. The calibrated kerma factors that were experimentally measured agreed well with the simulation and integral calculation results. This report describes a complete set of methods and fundamental data for the analysis of neutron-induced radiation effects at the CSNS on silicon-based semiconductor devices.

Neutron capture cross-section measurement and resolved resonance analysis of 237Np

ABSTRACTThe neutron capture cross-section of 237Np has been measured in the neutron energy region from 10 meV to 500 eV. A neutron time-of-flight method was employed using the NaI(Tl) spectrometer in the ANNRI beamline at the Japanese Proton Accelerator Research Complex (J-PARC). The experimental capture yield was derived using the pulse-height weighting technique and an energy-dependent cross-section was obtained relative to the incident neutron spectrum derived from a 10B(n, )7Li reaction yield. The absolute cross-section was determined by normalizing the results to JENDL-4.0 cross-section data at the first resonance of 237Np. The thermal cross-section was measured to be 177.6 ± 3.8 b. The resolved resonance region was analyzed with the REFIT code. An average radiation width = 40.3 ± 0.5 meV, a mean level spacing D0 = 0.60 ± 0.02 eV and a neutron strength function 104S0 = 1.02 ± 0.12 were obtained.

Impact of Nuclear Data Uncertainty in the Modeling of Neutron-Induced Recoil Atom Energy Distributions in Silicon

Recoil ion distributions in silicon and the resulting distribution of the linear energy transfer (LET) are important metrics in microdosimetric studies and in the investigation of neutron-induced single-event effects. A rigorous methodology is presented for quantifying the uncertainty in these metrics due to the underlying uncertainty contributors, including that due to the nuclear data, recoil ion electronic stopping power, and incident neutron spectrum. The methodology uses a Monte Carlo-based approach so that the nonlinear uncertainty propagation is rigorously treated as the response function folded with the full incident neutron spectrum. The uncertainty is captured in the form of both recoil energy and LET-dependent covariance matrices. The uncertainty contributions from the nuclear data are shown to have strong energy-dependent correlations which are comparable in magnitude to that from the uncertainty found in the spectrum characterization for high fidelity reference neutron fields. The uncertainty from the stopping power has the largest magnitude of the largest contributors, but it shows a very strong energy-dependent correlation that translates into a systematic uncertainty that may cancel out in many applications.

Incident neutron spectra on the first wall and their application to energetic ion diagnostics in beam-injected deuterium–tritium tokamak plasmas

A diagnostic method for small non-Maxwellian tails in fuel-ion velocity distribution functions is proposed; this method uses the anisotropy of neutron emissions, and it is based on the numerical analysis of the incident fast neutron spectrum on the first wall of a fusion device. Neutron energy spectra are investigated for each incident position along the first wall and each angle of incidence assuming an ITER-like deuterium–tritium plasma; it is heated by tangential-neutral-beam injection. Evaluating the incident neutron spectra at all wall positions and angles of incidence enables the selective measurement of non-Gaussian components in the neutron emission spectrum for energetic ion diagnostics; in addition, the optimal detector position and orientation can be determined. At the optimal detector position and orientation, the ratio of non-Gaussian components to the Gaussian peak can be two orders of magnitude greater than the ratio in the neutron emission spectrum. This result can improve the accuracy of energetic ion diagnostics in plasmas when small, anisotropic non-Maxwellian tails are formed in fuel ion velocity distribution functions. We focus on the non-Gaussian components greater than 14 MeV, where the effect of the background noise (i.e., slowing-down neutrons by scattering throughout the machine structure) can be ignored.

Estimation of neutron energy distributions from prompt gamma emissions

A technique of estimating the incident neutron energy distribution from emitted prompt gamma intensities from a system exposed to neutrons is presented. The emitted prompt gamma intensities or the measured photo peaks in a gamma detector are related to the incident neutron energy distribution through a convolution of the response of the system generating the prompt gammas to mono-energetic neutrons. Presently, the system studied is a cylinder of high density polyethylene (HDPE) placed inside another cylinder of borated HDPE (BHDPE) having an outer Pb-cover and exposed to neutrons. The emitted five prompt gamma peaks from hydrogen, boron, carbon and lead can be utilized to unfold the incident neutron energy distribution as an under-determined deconvolution problem. Such an under-determined set of equations are solved using the genetic algorithm based Monte Carlo de-convolution code GAMCD. Feasibility of the proposed technique is demonstrated theoretically using the Monte Carlo calculated response matrix and intensities of emitted prompt gammas from the Pb-covered BHDPE–HDPE system in the case of several incident neutron spectra spanning different energy ranges.

Neutron conversion and cascaded cooling in paramagnetic systems for a high-flux source of very cold neutrons

A new neutron-cooling mechanism is proposed with potential benefits for novel intense sources of very cold neutrons with wavelengths $g2$ nm, and for enhancing the production of ultracold neutrons. It employs inelastic magnetic scattering in weakly absorbing, cold paramagnetic systems. Kinetic energy is removed from the neutron stepwise in constant decrements determined by the Zeeman energy of paramagnetic atoms or ions in an external magnetic field, or by zero-field level splittings in magnetic molecules. The stationary neutron transport equation is analyzed for an infinite, homogeneous medium with Maxwellian neutron sources, using inelastic scattering cross sections derived in an appendix. Nonmagnetic inelastic scattering processes are neglected. The solution therefore still underestimates very cold neutron densities that should be achievable in a real medium. Molecular oxygen with its triplet ground state appears particularly promising, notably as a host in fully deuterated ${\mathrm{O}}_{2}$-clathrate hydrate. Other possibilities are dry ${\mathrm{O}}_{2}\text{\ensuremath{-}}^{4}\mathrm{He}$ van der Waals clusters and ${\mathrm{O}}_{2}$ intercalated in fcc-${\mathrm{C}}_{60}$. For conversion of cold to ultracold neutrons, where an incident neutron imparts only a single energy quantum to the medium, the paramagnetic scattering in the clathrate system is found to be stronger, by more than an order of magnitude, than the single-phonon emission in superfluid helium, when evaluated for an incident neutron spectrum with the optimum temperature for the respective medium. Moreover, the multistep paramagnetic cooling cascade leads to further strong enhancements of very cold neutron densities, e.g., by a factor 14 (57) for an initial neutron temperature of $30\phantom{\rule{4pt}{0ex}}\mathrm{K} (100\phantom{\rule{4pt}{0ex}}\mathrm{K})$, for the moderator held at about $1.3\phantom{\rule{4pt}{0ex}}\mathrm{K}$. Due to a favorable Bragg cutoff of the ${\mathrm{O}}_{2}$ clathrate, the cascade-cooling can take effect in a moderator with linear extensions smaller than a meter.

Activation analysis study on Li-ion batteries for nuclear forensic applications

The nuclear materials environment has been increasing significantly in complexity over the past couple of decades. The prevention of attacks from nuclear weapons is becoming more difficult, and nuclear forensics is a deterrent by providing detailed information on any type of nuclear event for proper attribution. One component of the nuclear forensic analysis is a measurement of the neutron spectrum. As an example, the neutron component provides information on the composition of the weapons, whether boosting is involved or the mechanisms used in creating a supercritical state. As 6Li has a large cross-section for thermal neutrons, the lithium battery is a primary candidate for assessing the neutron spectrum after detonation. The absorption process for 6Li yields tritium, which can be measured at a later point after the nuclear event, as long as the battery can be processed in a manner to successfully extract the tritium content. In addition, measuring the activated constituents after exposure provides a means to reconstruct the incident neutron spectrum. The battery consists of a spiral or folded layers of material that have unique, energy dependent interactions associated with the incident neutron flux. A detailed analysis on the batteries included a pre-irradiated mass spectrometry analysis to be used as input for neutron spectrum reconstruction. A set of batteries were exposed to a hard neutron spectrum delivered by the University of Massachusetts, Lowell research reactor Fast Neutron Irradiator (FNI). The gamma spectra were measured from the batteries within a few days and within a week after the exposure to obtain sufficient data on the activated materials in the batteries. The activity was calculated for a number of select isotopes, indicating the number of associated neutron interactions. The results from tritium extraction are marginal. A measurable increase in detected particles (gammas and betas) below 50keV not self-attenuated by the battery was observed, yet as the spectra are coarse, the gamma information is not separable from tritium spectra. The activation analysis was successful, and the incident neutron spectrum was reconstructed using materials found in lithium batteries.

Cross section measurement of the 159Tb(n, γ)Tb160 nuclear reaction

Abstract The cross section of the 159Tb(n, γ)Tb160 reaction was measured in four mono-energetic neutron fields of energy 3.7, 4.3, 5.4, and 6.85 MeV, respectively, with the activation technique applied to metal discs of natural composition. To ensure an acceptable precision of the results all major sources of uncertainties were taken into account. Calculations of detector efficiency, incident neutron spectrum and correction factors were performed with the Monte Carlo code (MCNPX), whereas theoretical excitation functions were calculated with the TALYS-1.2 code and compared to the experimental cross section values. This paper presents both measurements and calculation leading to the cross section values.

Maxwellian Averaged Neutron Capture Cross Sections in 56Fe and 57Fe

Measurements of the keV-neutron capture cross sections and radiative γ-ray spectrum of 56Fe and 57Fe are performed based on a 7Li(p,n)7Be reaction neutron source. The incident neutron spectrum on a capture sample is measured by means of a time-of-flight (TOF) method with a 6Li-glass detector. The radiative capture γ-rays emitted from an iron (56Fe or 57Fe) or standard gold (197Au) sample are detected by a large anti-Compton NaI(Tl) spectrometer covered with a heavy shield. The capture yields of samples are obtained by applying a pulse-height weighting technique to the corresponding capture γ-ray pulse-height spectrum. The Maxwellian averaged neutron capture cross sections of 56Fe and 57Fe are derived according to the present capture cross section results.

Study on the keV neutron capture reaction in 56Fe and 57Fe

The neutron capture cross-sections and the radiative capture gamma-ray spectra from the broad resonances of 56Fe and 57Fe in the neutron energy range from 10 to 90keV and 550keV have been measured with an anti-Compton NaI(Tl) detector. Pulsed keV neutrons were produced from the 7Li\( (p,n)\) 7Be reaction by bombarding the lithium target with the 1.5ns bunched proton beam from the 3MV Pelletron accelerator. The incident neutron spectrum on a capture sample was measured by means of a time-of-flight (TOF) method with a 6Li -glass detector. The number of weighted capture counts of the iron or gold sample was obtained by applying a pulse height weighting technique to the corresponding capture gamma-ray pulse height spectrum. The neutron capture gamma-ray spectra were obtained by unfolding the observed capture gamma-ray pulse height spectra. To achieve further understanding on the mechanism of neutron radiative capture reaction and study on physics models, theoretical calculations of the \( \gamma\) -ray spectra for 56Fe and 57Fe with the POD program have been performed by applying the Hauser-Feshbach statistical model. The dominant ingredients to perform the statistical calculation were the Optical Model Potential (OMP), the level densities described by the Mengoni-Nakajima approach, and the \( \gamma\) -ray transmission coefficients described by \( \gamma\) -ray strength functions. The comparison of the theoretical calculations, performed only for the 550keV point, show a good agreement with the present experimental results.

Li-Ion Batteries Used as Ubiquitous Neutron Sensors for Nuclear Forensics

With an increasingly complex worldwide nuclear environment, nuclear forensics provides a deterrent through proper attribution. Identifying the type of event is important in the attribution process, and the lithium battery has the potential to provide significant information for this purpose. The incident neutron spectral fluence is related to the type of nuclear event. The lithium battery is built using materials that can be used to assess the number of incident neutrons, as well as provide threshold detection dependant on the neutron energy. This study looks at correlating the production of various long-lived radioactive isotopes to the incident neutron spectrum. The analysis uses known nuclear cross sections and battery properties to calculate the number of long-lived radioactive elements that can be produced in the solid winding of the active battery elements, which include the cathode, separator, and anode. Some isotopes are only produced above a threshold energy, such as 58Co , which requires 10.6 MeV neutrons. Other isotopes such as 3H are produced through various reaction channels, mostly 6Li(n,α)3H, which is an exothermic reaction, and is not associated with a threshold energy.

The Nanoscale Ordered MAterials Diffractometer NOMAD at the Spallation Neutron Source SNS

The Nanoscale Ordered MAterials Diffractometer (NOMAD) is neutron time-of-flight diffractometer designed to determine pair distribution functions of a wide range of materials ranging from short range ordered liquids to long range ordered crystals. Due to a large neutron flux provided by the Spallation Neutron Source SNS and a large detector coverage neutron count-rates exceed comparable instruments by one to two orders of magnitude. This is achieved while maintaining a relatively high momentum transfer resolution of a δQ/Q∼0.8% FWHM (typical), and a possible δQ/Qof0.24% FWHM (best). The real space resolution is related to the maximum momentum transfer; a maximum momentum transfer of 50Å−1 can be obtained routinely and the maximum momentum transfer given by the detector configuration and the incident neutron spectrum is 125Å-1. High stability of the source and the detector allow small contrast isotope experiments to be performed. A detailed description of the instrument is given and the results of experiments with standard samples are discussed.

Light-Ion Production in The Interaction of 175 MeV Neutrons with Iron and Bismuth

We have measured double-differential (angle and energy) cross sections for light-ion (p, d, t, (3)He, and a) production in the interaction of quasi-monoenergetic 175 MeV neutrons with iron and bismuth. Measurements have been performed at the The Svedberg Laboratory, Uppsala (Sweden), using the Medley setup which allows low-energy thresholds and wide energy and angular ranges. Medley is a spectrometer system consisting of eight three-element telescopes placed at angles from 20 degrees to 160 degrees, in steps of 20 degrees. Each telescope is composed of two silicon surface barrier detectors and a CsI(Tl) scintillator, to perform particle identification, fully stop the produced light-ions and measure their kinetic energy. The time-of-flight was used to reduce the contribution from the low energy tail in the accepted incident neutron spectrum. We report double-differential production cross sections for protons, deuterons, tritons, (3)He and alpha particles and compare them with model calculations with TALYS-1.2.