MeV Neutron Energy Research Articles

Gallium oxide (Ga2O3) energy dependent scintillation response to fast neutrons and flash gamma-rays.

Gallium oxide is a newly emerged ultrawide bandgap (4.9eV) semiconductor that is suitable as a combined electronics and radiation detection platform. We have experimentally demonstrated fast neutron and gamma-ray scintillation from Czochralski-grown β-Ga2O3 in a recent series (October 2023) of experiments at the unmoderated pulsed neutron spallation source located at the Los Alamos Neutron Science Center. Using the neutron time-of-flight (TOF) technique and a fast-gated intensified CCD camera, we observed energy-dependent neutron scintillation for neutron energies ranging from 1 to 400MeV, including the 14.1MeV neutron energy relevant to D-T fusion. Neutron flux is quantified and calibrated by cascading the scintillator after the fission chamber, enabling a detailed analysis of temporal and energy-dependent characteristics of the scintillation events. A pronounced scintillation signal from the spallation gamma flash with a temporal full width of half maximum of ∼4ns is indicative of the material's rapid response. Neutron energy dependent scintillation is observed using the TOF method at a 22.6-m distance from the neutron source. These results highlight the possibility of developing a Ga2O3 based fusion neutron diagnostic platform integrated with both scintillation and electronics functions on the integrated chip scale.

Cross section measurement for the 14N(n,α0,1)11B reactions in the 4.5–11.5 MeV neutron energy region

Cross section measurement for the 14N(n,α0,1)11B reactions in the 4.5–11.5 MeV neutron energy region

Cross sections for 14 MeV neutron interaction with lutetium isotopes and their theoretical excitation functions* *Supported by the Natural Science Foundation of Henan (232300420130) and National Natural Science Foundation of China (11575090)

The cross-sections for the 175Lu(n,p)175Yb, 175Lu(n,α)172Tm, 176Lu(n,α)173Tm, 175Lu(n,2n)174mLu, and 175Lu(n,2n)174gLu reactions at 13.57, 14.03 14.62, and 14.86 MeV neutron energies were measured using an activation technique. The theoretical excitation functions of these reactions were calculated using the Talys-1.95 code. The reaction cross-section data experimentally obtained were analyzed and compared with experimental data reported in the literature, data from five major evaluated nuclear data libraries of IAEA, and theoretical values based on Talys-1.95. The data obtained at some neutron energies agree with some of the data reported in the literature and theoretical values based on Talys-1.95. The consistency of the theoretical curves of excitation functions based on Talys-1.95 with the data obtained in this study and those reported in the literature is higher than that of the evaluation curves of excitation functions for the 175Lu(n,p)175Yb, 175Lu(n,α)172 Tm, and 176Lu(n,α)173Tm reactions. This study is helpful because it provides new evaluated reaction cross-section data on lutetium (which is a fusion reactor material), improves the quality of neutron-induced reaction cross section data libraries, and advances the research on related applications.

Double differential cross section calculations of (n,p) reactions on 27Al and 24Mg targets

In structural fusion material research, double differential sections are necessary to determine heating and damage due to secondary particles that may occur in structural materials. Therefore, in this study, the double differential proton emission cross sections of 27Al and 24Mg target nuclei were calculated theoretically with the TALYS nuclear reaction code at 14 MeV neutron energy. The calculated values were compared with existing experimental data in the EXFOR library. Additionally, the contribution of direct, compound and preequilibrium reactions in theoretical calculations was investigated.

Measurement of neutron induced reaction cross-section of 99Mo

In the present work, we have measured 98Mo(n,γ)99Mo reaction cross-section using a 7Li(p,n)7Be neutron source at 1.67 ± 0.14, 2.06 ± 0.14 and 2.66 ± 0.16 MeV neutron energies. We have employed offline γ-ray spectroscopy to measure the induced activity of the sample. The 115In(n,n’γ)115mIn reaction was used as a monitor reaction. Different attributes propagating the uncertainty in the total result, measured cross-sections with their uncertainties and correlation coefficients are given in detail in the present study. The result is compared with the data libraries, EXFOR database and theoretical model outcome from different level density models.

Empirical formula for (n, f) reaction cross sections of Uranium isotopes at 1–20 MeV neutrons

This article describes how to calculate neutron-induced fission reaction cross sections using a proposed empirical formula and the EMPIRE 3.2.3 and TALYS 1.95 computer codes for Uranium isotopes up to the third fission plateau. In this study, the excitation functions of 232U (n, f), 233U (n, f), 234U (n, f), 235U (n, f), 236U (n, f), 237U (n, f) and 238U (n, f) nuclear reactions were calculated at 1–20 MeV neutron energies. The results were contrasted to measured values from the Experimental Nuclear Reaction Data (EXFOR) as well as the evaluated data from Evaluated Nuclear Data File (ENDF) such as (EAF-2010, JEFF-3.3, ENDF/B-VIII.0 and TENDL-2019). Overall, the calculated, experimental, and evaluated fission cross-sections are in concordance.

Montecarlo Application on Nucleon Dynamics in Calculating Fission Yield at 14 MeV Neutron Energy

Nuclear data is a completeness that must be present in every activity related to nuclear technology. So high is the role of nuclear data, it is necessary to have very complete nuclear data. The need for nuclear data is not in line with the resulting experimental products. The amount of experimental data needs to be completed. This is because the operational costs for these experiments are costly. Thus, theoretical modeling calculations are inevitably the right choice to replace experimental results. Many theoretical models have been developed to obtain satisfactory results. They were starting from microscopic models to macroscopic models. A common obstacle is that microscopic models must be simplified and efficient to produce massive nuclear data. Meanwhile, the constraints on the macroscopic model could be more accurate. This paper will present a calculation that tries to produce accurate but uncomplicated and economical data. This technique uses the basic principles of random numbers and classical nucleon dynamics in the nucleus. At the end of the paper, the results of calculations are presented, which are very accurate and, at the same time, show the dynamics of the nucleons that occur.

Development of a compact fast-neutron spectrometer for nuclear emergency response applications

We have developed a Compact Fast Neutron Spectrometer (CFNS) for passive assay of special nuclear material (SNM) through the observation of fast neutrons. The CFNS consists of eight organic glass scintillators (OGS) coupled to silicon photomultipliers and a waveform digitizer, which are integrated within a human-portable box. The CFNS determines the neutron energy profile by spectrum unfolding using the Maximum-Likelihood Expectation Maximization method. The detector acquisition system was optimized to have a dynamic range of up to 10 MeV neutron energy. Bulk special nuclear material (SNM) measurements from the National Criticality Experiments Research Center were analyzed for SNM validation/examination. The results show that the CFNS can be used to distinguish between fission and (α, n) neutron emitters, regardless of intervening material type (Cu and polyethylene) and thickness, by taking the ratio of neutron counts at different regions in the unfolded energy spectrum. Additionally, by fitting an exponential curve to the unfolded energy spectrum of PuO2 and Pu neutron emitters, the CFNS showed the ability of distinguishing between pure Pu oxide, pure Pu metal and mixed oxide-metal configurations.

76Se(n,2n)75Se reaction cross section measurement at 14.77 MeV neutron energy

The cross section of 76Se(n,2n)75Se reaction at 14.77 ± 0.17 MeV neutron energy was measured with offline gamma spectrometry. Furthermore, covariance analysis for the measured cross section with respect to the other neutron induced reaction cross sections for the same target was reported. The experimental results were compared with reported experimental data from the EXFOR database and evaluated data from the ENDF/B-VIII.0, JENDL-5, and TENDL-2021 libraries. Statistical model calculations with the TALYS-1.96, and EMPIRE-3.2.3 codes were carried out for optimized parameters corresponding to experimental trends. The measured cross section is in good agreement with the data from the EXFOR database, and the evaluated data libraries.

Recoil Proton Telescopes and Parallel Plate Avalanche Counters for the 235U(n,f) cross section measurement relative to H(n,n)H between 10 and 450 MeV neutron energy

With the aim of measuring the 235U(n,f) cross section at the n_TOF facility at CERN over a wide neutron energy range, a detection system consisting of two fission detectors and three detectors for neutron flux determination was realized. The neutron flux detectors are Recoil Proton Telescopes (RPTs), based on scintillators and solid state detectors, conceived to detect recoil protons from the neutron-proton elastic scattering reaction. This system, along with a fission chamber and an array of parallel plate avalanche counters for fission event detection, was installed for the measurement at the n_TOF facility in 2018, at CERN.An overview of the performances of two RPTs — especially developed for this measurement — and of theparallel plate avalanche counters are described in this article. In particular, the characterizationin terms of detection efficiency by Monte Carlo simulations and response to neutron beam, the studyof the background, dead time correction and characterization of the samples, are reported. Theresults of the present investigation show that the performances of these detectors are suitable foraccurate measurements of fission reaction cross sections in the range from 10 to 450 MeV.

Research on the application of pulse shape discrimination method in (n, lcp) reaction cross-section measurements

The China Spallation Neutron Source Back-n white neutron source supplies neutrons with a wide energy range for nuclear data measurements and nuclear techniques. In cross-section measurements of neutron induced light charged-particle emission (n, lcp) reactions at Back-n, the identification of charged particles are usually achieved with the ΔE-E method or a two-dimensional spectrum of the signal amplitude of the particles deposited in the detector and the corresponding neutron energies derived from the time of flight, called amplitude-En method. Nevertheless, a deficiency still exists in the measurement range and the identification performance with these methods. The pulse shape discrimination (PSD) method with a silicon detector is supposed to be a supplement to the amplitude-En and ΔE-E methods. With silicon detectors, experiments have been performed to compare the performance of the amplitude-En and the PSD methods. The PSD method shows a good performance in identifying light charged particles in the MeV neutron energy region. Alphas and tritons emitted from 6Li(n, t) reaction can be identified with the low energy threshold around 2.5 MeV. This provides an effective method for the cross-section measurement of (n, lcp) reactions at Back-n.

Shielding Capability Research on Composite Base Materials in Hybrid Neutron-Gamma Mixed Radiation Fields

The 16N monitoring system operates in a mixed neutron-gamma radiation field and is subject to high background radiation, thus triggering instability in the 16N monitoring system measurement data. Due to its property of actual physical process simulation, the Monte Carlo method was adopted to establish the model of the 16N monitoring system and design a structure-functionally integrated shield to realize neutron-gamma mixed radiation shielding. First, the optimal shielding layer with a thickness of 4 cm was determined in this working environment, which had a significant shielding effect on the background radiation and improved the measurement of the characteristic energy spectrum and the shielding effect on neutrons was better than gamma shielding with the increase in the shield thickness. Then, functional fillers such as B, Gd, W, and Pb were added to the matrix to compare the shielding rates of three matrix materials of polyethylene, epoxy resin, and 6061 aluminum alloy at 1 MeV neutron and gamma energy. The shielding performance of epoxy resin as the matrix material was better than that of the aluminum alloy and polyethylene, and the shielding rate of boron-containing epoxy resin was 44.8%. The γ-ray mass attenuation coefficients of lead and tungsten in the three matrix materials were simulated to determine the best material for the gamma shielding performance. Finally, the optimal materials for neutron shielding and gamma shielding were combined, and the shielding performance of single-layer shielding and double-layer shielding in mixed radiation field was compared. The optimal shielding material-boron-containing epoxy resin was determined as the shielding layer of the 16N monitoring system to realize the integration of structure and function, which provides a theoretical basis for the selection of shielding materials in a special working environment.

63Cu(n, α)60Co cross sections in the MeV region

Cross sections of the 63Cu(n, α)60Co reaction in the MeV neutron energy region were measured using both the direct measurement method and the activation method. A twin-gridded ionization chamber was used for the direct measurement while the high-purity germanium detector was used for the activation method. The measured cross sections using the two methods are consistent within measurement uncertainties. The present results support the evaluation data in the ENDF/B-VII.1 library instead of the latest evaluation data in the ENDF/B-VIII.0 library, which is helpful to clarify discrepancies in measurement and evaluation data. TALYS-1.9 code analysis was performed which shows that the measurement cross sections can be well reproduced with minor adjustments of the input parameters. Through the calculation, it is found that the compound mechanism predominates the 63Cu(n, α)60Co reaction with ratios higher than 0.7 for neutron energy less than 20 MeV.

Cross section measurements of low-energy charged particle induced reactions using moderated neutron counter arrays

A high-and-flat efficiency moderated neutron detection array of 3He counters has been recently developed for photoneutron cross section measurements at the future ELI-NP gamma-ray beam source. We have designed a three rings geometry of 31 counters with a ˜37% efficiency flat within 5% in the 10 keV to 5 MeV neutron energy interval. A commissioning experiment was performed using proton beams delivered by the 9 MV Tandem accelerator of IFIN-HH. We measured the neutron production cross sections for proton induced reactions on nat Cu and 27Al. The (p, xn) reactions on nat Cu were investigated in the 4.5 MeV to 14 MeV proton energy range with 250 keV steps, using pulsed and continuous proton beams. The low energy measurements below 10 MeV served to validate the detection efficiency calibration against well-known natCu(p, n) cross sections. The 27Al(p, n) reaction was investigated in the 5.8 MeV to 6.75 MeV energy range with 5 keV steps. Preliminary cross section results are here compared with preceding data.

Few-layer hexagonal boron nitride / 3D printable polyurethane composite for neutron radiation shielding applications

Functional polymer composites can confer a range of benefits in practical applications that go beyond the individual properties of the constituent materials. Here we investigate and characterize the neutron absorbing capability of few-layer hexagonal boron nitride (h-BN) in composite with a 3D-printable thermoplastic polyurethane, and present experiment and simulation data to understand the processes and mechanisms in play. Shielding and protection from neutrons can be necessary in a range of terrestrial and space-based applications. The neutron absorption of composites with varying fractions of h-BN is strongly energy-dependent in the low-energy regime below 10 meV, and a composite containing 20 wt% h-BN shows a 70-fold reduction in the transmission relative to pure polyurethane at 0.5 meV neutron energies. This is attributed to the strong neutron capture cross-section of the naturally abundant boron-10 isotope, with energy-dependent measurements up to 100 meV confirming this point. Using inelastic neutron spectroscopy, we identify additional effects from the hydrogen in the polyurethane which both scatters diffusively and moderates neutrons inelastically via its phonon spectrum, enhancing the neutron absorption characteristics. Two models – based on analytic functions and Monte Carlo numerical techniques – are presented, and show excellent agreement with experiment results. The 3D-printability of the composite is demonstrated, and the opportunities and challenges for deploying these composites in neutron radiation protection applications are discussed.

New cross-section measurements for the (n, 2n), (n, p), and (n, α) reactions on bromine in the 14 MeV region with detailed uncertainty quantification* *Supported by the National Natural Science Foundation of China (12165006, 11875016).

Measurement of the cross-sections of the 79Br(n, 2n)78Br, 81Br(n, p)81mSe, 81Br(n, α)78As, and 79Br(n, α)76As reactions was performed at specific neutron energies, precisely, 13.5±0.2, 14.1±0.2, 14.4±0.2, and 14.8±0.2 MeV, relative to the standard 93Nb(n, 2n)92mNb and 27Al(n, α)24Na reference reactions using offline γ-ray spectrometry and neutron activation. Monoenergetic neutrons were generated at the China Academy of Engineering Physics via a 3H(d, n)4He reaction using the K-400 Neutron Generator equipped with a solid 3H-Ti based target. The activity of the reaction produce was obtained using a high-purity germanium detector. The cross-sections of the (n, 2n), (n, p), and (n, α) reactions on the bromine isotopes were measured in the 13–15 MeV neutron energy range. The covariance analysis approach was employed for a thorough inspection of any uncertainties within the measured cross-section data. A discussion and comparison of the observed outcome were carried out with previously published data, especially with the results of the JENDL-4.0, JEFF-3.3, TENDL-2019, and ENDF/B-VIII.0 data libraries, along with the theoretical excitation function curve derived by employing the TALYS-1.95 program. Improved cross-section restrictions for the investigated processes in the 13–15 MeV neutron energy range will be obtained using the current findings, which will help to raise the caliber of associated databases. Furthermore, the parameters of relevant nuclear reaction models can be verified using this data.

Estimations for [formula omitted] reaction cross sections at around 14.5MeV using Levenberg-Marquardt algorithm-based artificial neural network

Prediction of neutron-induced reaction cross-sections at around the 14.5 MeV neutron energy is crucial to calculate nuclear transmutation rates, nuclear heating, and radiation damage from gas formation in fusion reactor technology In this research, the new approach of (n,α) reaction cross-section is presented. It has been assessed by utilizing the artificial neural network (ANN) when compared to more advanced algorithms, the Levenberg-Marquardt algorithm-based ANN can be exceedingly fast. The correlation coefficients for a training R-value of 0.99283, a validation R-value of 0.991190, a testing R-value of 0.97337, and an overall R-value of 0.98515 demonstrate that Levenberg-Marquardt algorithm-based ANN is well suited for this purpose. . The obtained results were compared to theoretical calculations of TALYS 1.95 nuclear code. As a consequence, it has been demonstrated that the ANN model can be used to determine the systemic study for (n, α) reaction cross-sections.

Excitation functions for fast-neutron induced reactions on zinc

Cross sections of the 64Zn(n, p)64Cu, natZn(n, x)67Cu, 66Zn(n, 2n)65Zn and 70Zn(n, 2n)69mZn reactions have been measured by using the activation technique at 14.0 MeV neutron energy. The neutrons were produced via the 3H(d, n)4He reaction. The present experimental data illustrated satisfactory agreement with most of the previously reported experimental data. Experimental data are compared with evaluated nuclear data of the CENDL-3.2, ENDF/B-VIII.0, JENDL-5, BROND-3.1 and JEFF-3.3 libraries. Besides, these excitation functions were calculated by using theoretical model of the TALYS-1.96 code from thresholds up to 20 MeV. A group set of parameters was obtained which better reproduce the experimental data than the default parameters.

Semi-empirical formula for (n, α) reaction cross sections at 14–15 MeV neutrons

Semi-empirical formula for calculating the average (n, α) reaction cross-section was derived by using the EXFOR experimental data at 14–15 MeV neutron energies for 133 target nuclei in the mass number range 9≤A≤209. In this new systematic, the non-elastic cross-section portion has been modified by (A13+1)−0.5196∗(BEA), while the asymmetry parameter-dependent term has been maintained. Using EMPIRE 3.2.3 and TALYS 1.95, the (n, α) cross sections for selected nuclei within the same energy range were calculated. The experimental data can be predicted with greater precision when the cross sections derived from the proposed formula are compared to those derived from codes and earlier systematics.

Measurement of neutron-capture cross sections of Ge70,72 using the DANCE facility

Background: Approximately half of all atomic nuclei heavier than iron are synthesized by the slow neutron-capture process. The weak component of this process is not well understood and the reaction rates of each isotope in the s-process path affect nucleosynthesis abundances downstream.Purpose: To measure the neutron-capture cross sections of two weak s-process nuclei, $^{70,72}\mathrm{Ge}$, using the neutron time-of-flight technique. Measuring the capture cross sections for isotopes in this region of the chart of nuclides has proven challenging due to dominant scattering cross sections.Method: Samples consisted of pellets made of pressed enriched metallic powders. The $^{70,72}\mathrm{Ge}$ neutron-capture cross sections were measured as a function of neutron energy using the Detector for Advanced Neutron Capture Experiments at Los Alamos National Laboratory.Results: Neutron-capture cross sections were measured from 10 eV to 1 MeV. These are the first measurements for $^{70,72}\mathrm{Ge}$ between 300 keV and 1 MeV neutron energy. Maxwellian-averaged cross sections were calculated in the astrophysically relevant neutron energy range (5 keV $\ensuremath{\le}kT\ensuremath{\le}$ 100 keV). Their value at $kT=30$ keV was found to be $89\ifmmode\pm\else\textpm\fi{}11$ mb for $^{70}\mathrm{Ge}$ and $58\ifmmode\pm\else\textpm\fi{}5$ mb for $^{72}\mathrm{Ge}$. Both values are in agreement with recent time-of-flight measurements at n_TOF (neutron Time-Of-Flight facility at the European Organization for Nuclear Research).Conclusions: The average cross section results from this work for $^{70}\mathrm{Ge}$ show minor ($<1\ensuremath{\sigma}$) disagreement with a recent measurement by the n_TOF collaboration at higher neutron energies. This corresponds to the neutron energy region that had previously never been measured ($>300$ keV). Two reaction library databases underestimate the $^{72}\mathrm{Ge}$ average cross section below 30 keV according to n_TOF and DANCE. This is likely due to capture resonances that are missing from the theoretical cross sections in the databases that were identified in both time-of-flight measurements. Additionally, a rudimentary analysis of the impact of both cross section measurements on stellar nucleosynthesis abundances using the NETZ nucleosynthesis tool is presented.