Neutron Energy Research Articles

NeuSIM4: A comprehensive GEANT4 based neutron simulation code

A new neutron SIMulation program based on the versatile GEANT4 toolkit, neuSIM4, has been developed to describe interactions of neutrons in the NE213 liquid scintillator from 0.1 to 3000 MeV. neuSIM4 is designed to accommodate complicated modern detector geometry setups with multiple scintillator detectors, each of which can be outfitted with more than one photo-multiplier. To address a broad spectrum of neutron energies, two new neutron interaction physics models, KSCIN and NxQMD, have been implemented in GEANT4. For neutrons with energy below 110 MeV, we incorporate a total of eleven neutron induced reaction channels on hydrogen and carbon nuclei, including nine carbon inelastic reaction channels, into KSCIN. Beyond 110 MeV, we implement a neutron induced reaction model, NxQMD, in GEANT4. We use its results as reference to evaluate other neutron-interaction physics models in GEANT4. We find that results from an existing cascade physics model (INCL) in GEANT4 agree very well with the results from NxQMD, and results from both codes agree with new and existing light response data. To connect KSCIN to NxQMD or INCL, we introduce a transition region where the contribution of neuSIM4 linearly decreases with corresponding increased contributions from NxQMD or INCL. To demonstrate the application of the new code, we simulate the light response and performance of a 2 × 2 m2 neutron detector wall array consisting of 25 2m-long scintillation bars. We are able to compare the predicted light response functions to the shape of the experimental response functions and calculate the efficiency of the neutron detector array for neutron energies up to 200 MeV. These simulation results will be pivotal for understanding the performance of modern neutron arrays with intricate geometries, especially in the measurements of neutron energy spectra in heavy-ion reactions.

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The compound nucleus model is employed to calculate neutron capture on 76\\documentclass[12pt]{minimal} \\usepackage{amsmath} \\usepackage{wasysym} \\usepackage{amsfonts} \\usepackage{amssymb} \\usepackage{amsbsy} \\usepackage{mathrsfs} \\usepackage{upgreek} \\setlength{\\oddsidemargin}{-69pt} \\begin{document}$$^{76}$$\\end{document}Ge and the consecutive γ\\documentclass[12pt]{minimal} \\usepackage{amsmath} \\usepackage{wasysym} \\usepackage{amsfonts} \\usepackage{amssymb} \\usepackage{amsbsy} \\usepackage{mathrsfs} \\usepackage{upgreek} \\setlength{\\oddsidemargin}{-69pt} \\begin{document}$$\\gamma $$\\end{document} decay. Total, inelastic and capture cross sections are compared to experiments in order to restrict model parameters. The respective γ\\documentclass[12pt]{minimal} \\usepackage{amsmath} \\usepackage{wasysym} \\usepackage{amsfonts} \\usepackage{amssymb} \\usepackage{amsbsy} \\usepackage{mathrsfs} \\usepackage{upgreek} \\setlength{\\oddsidemargin}{-69pt} \\begin{document}$$\\gamma $$\\end{document} cascades are analyzed for possible background suppression schemes in future 0νββ\\documentclass[12pt]{minimal} \\usepackage{amsmath} \\usepackage{wasysym} \\usepackage{amsfonts} \\usepackage{amssymb} \\usepackage{amsbsy} \\usepackage{mathrsfs} \\usepackage{upgreek} \\setlength{\\oddsidemargin}{-69pt} \\begin{document}$$0\ u \\beta \\beta $$\\end{document} decay searches with germanium detectors. Neutron energies in the range from 0.2 keV to 5 MeV are investigated. The spin dependence of the γ\\documentclass[12pt]{minimal} \\usepackage{amsmath} \\usepackage{wasysym} \\usepackage{amsfonts} \\usepackage{amssymb} \\usepackage{amsbsy} \\usepackage{mathrsfs} \\usepackage{upgreek} \\setlength{\\oddsidemargin}{-69pt} \\begin{document}$$\\gamma $$\\end{document} cascades is proposed to be employed for determining the spin of thermal resonances.

Observation of energetic ion anisotropy using neutron diagnostics in the Large Helical Device

Energetic ion anisotropy was observed by tangential sightline compact neutron energy spectrometers (CNESs) in tangential neutral beam heated deuterium plasmas in Large Helical Device. Significant upper and lower energy shifts in D–D neutron energy from 2.45 MeV were measured according to the beam ion injection directions and CNES sightline using a conventional liquid scintillation detector with the unfolding technique and a novel Cs2LiYCl6:Ce with a 7Li-enrichment (CLYC7) scintillation detector without unfolding. The observed neutron energy spectrum was compared with that predicted by a numerical simulation based on orbit following models. Numerical simulation revealed that the Doppler shift in D–D neutron energy results from energetic ion anisotropy.

Machine Learning Based Compton Suppression for Nuclear Fusion Plasma Diagnostics

Diagnostics are critical on the path to commercial fusion reactors, since measurements and characterisation of the plasma is important for sustaining fusion reactions. Gamma spectroscopy is commonly used to provide information about the neutron energy spectrum from activation analysis, which can be used to calculate the neutron flux and fusion power. The detection limits for measuring nuclear dosimetry reactions used in such diagnostics are fundamentally related to Compton scattering events making up a background continuum in measured spectra. This background lies in the same energy region as peaks from low-energy gamma rays, leading to detection and characterisation limitations. This paper presents a digital machine learning Compton suppression algorithm (MLCSA), that uses state-of-the-art machine learning techniques to perform pulse shape discrimination for high purity germanium (HPGe) detectors. The MLCSA identifies key features of individual pulses to differentiate between those that are generated from photopeaks and Compton scatter events. Compton events are then rejected, reducing the low energy background. This novel suppression algorithm improves gamma spectroscopy results by lowering minimum detectable activity (MDA) limits and thus reducing the measurement time required to reach the desired detection limit. In this paper, the performance of the MLCSA is demonstrated using an HPGe detector, with a gamma spectrum containing americium-241 (Am-241) and cobalt-60 (Co-60). The MDA of Am-241 improved by 51% and the signal to background ratio improved by 49%, while the Co-60 peaks were partially preserved (reduced by 78%). The MLCSA requires no modelling of the specific detector and so has the potential to be detector agnostic, meaning the technique could be applied to a variety of detector types and applications.

Investigation of Cs27LiYCl6:Ce scintillator energy response for D-D fusion neutron spectrometer

During plasma experiments on EAST (Experimental Advanced Superconducting Tokamak), the discharges heated with neutral beam injection (NBI) for plasma results in the production of 2.45 MeV deuterium-deuterium (D-D) fusion neutrons. In order to measure the energy spectrum of D-D fusion neutrons, a new Cs27LiYCl6:Ce (CLYC-7) crystal with 7Li-enrichment has been developed and fabricated into a CLYC-7 scintillation detector. Before utilizing this detector for studying neutron energy spectra, it is crucial to explore the response function of the detector within a relevant neutron energy range. Therefore, the detector has been experimented using the 4.5 MeV electrostatic accelerator at Peking University to investigate its response to fast neutrons, energy resolution, and pulse discrimination capability in the neutron energy range 2.07 MeV–5.51 MeV. The neutron reactions occurring within the CLYC-7 scintillator were simulated using MCNPX, focusing on the reactions 35Cl(n,p) and 35Cl(n,α), as well as the energy response curve and the contributions of different particles (protons, alphas) to the neutron spectrum. The experimental results were successfully validated on EAST, where the 2.45 MeV fusion neutron energy spectrum was measured. That demonstrates the ability of the CLYC-7 scintillation detector to diagnose the fusion neutron energy spectrum for deuterium plasma discharge.

Calculation of the energy dependence of fission fragments yields and kinetic energy distributions for neutron-induced 235U fission* *Supported by the National Natural Science Foundation of China (12075105), the NSFC-Nuclear Technology Innovation Joint Fund (U2167203), the Major Science and Technology Projects of Gansu Province (22ZD6GB020), the Fundamental Research Funds for the Central Universities of China (lzujbky-2022-ey14, lzujbky-2022- kb07)

Fission fragments yields and average total kinetic energy are fundamental nuclear data for nuclear energy applications and the study of nuclear devices. Certain fission products, such as 95Zr, 99Mo, 140Ba, 144Ce, and 147Nd, serve as burnup monitors, assessing the number of fissions induced by neutrons on 235U. However, current experimental data for these fission products worldwide are inconsistent, introducing significant uncertainty into related scientific research. In this study, we employed the Potential-driving Model to calculate the independent yields of 235U and evaluate its advantages in such calculations. Additionally, we investigated the energy dependence of independent yields to select important products. Furthermore, we calculated the cumulative yields of 95Zr, 99Mo, 140Ba, 144Ce, and 147Nd, and compared them with existing literature data to explore the energy dependence of fission products for 235U. Given the lack of fission product yield data above 14.8 MeV, we extended our calculated incident neutron energy to 20 MeV, aiming to support future scientific research. The Geant4 physical model does not consider the influence of incident neutron energy on the average total kinetic energy of fission fragments; thus, we introduced the excitation function of the total kinetic energy of fission fragments recommended by Madland et al., which effectively describes the experimental data of the average total kinetic energy of fragments formed in 235U fission. In this paper, we comprehensively discuss the energy dependence of fission product yields and average total kinetic energy.

Simultaneous measurements for fast neutron flux and tritium production rate using pulse shape discrimination and single crystal CVD diamond detector

This paper presents the development of a simultaneous measurement method for fast neutron energy spectra and tritium production rates within mixed radiation fields using a single crystal chemical vapour deposition diamond detector combined with a lithium fluoride (LiF) foil. The method involves the separation of pulses with rectangular shapes and the determination of the depth position within the single crystal diamond (SCD) struck by fast neutrons or nuclear reaction products including recoil tritons from the LiF foil based on pulse width, extracting pulse events occurred at the specific bulk region and the surface region of the SCD. Subsequently, unfolding techniques were employed to analyse the energy deposition spectrum of pulses at the specific bulk region which are induced only by fast neutrons, allowing the deduction of the fast neutron energy spectrum. To evaluate the tritium production rate, the energy deposition spectrum of pulses from events occurring at the SCD surface facing the LiF foil was analysed. By estimating the energy deposition spectrum solely induced by fast neutrons striking the SCD surface and subtracting it from the energy deposition spectrum of events at the SCD surface, the contribution of energetic ions, such as recoil tritons generated by the 6Li(n,α)3H reaction in the LiF foil, was determined. The fast neutron flux and tritium production rate obtained through this study were consistent with particle transport calculations, demonstrating the successful development of a method suitable for performance testing of fusion reactor blankets.

Investigation of breakup process in 9Be(α,n)12C reaction

Neutron energies and angular distributions were measured in 9Be(α,n)12C reaction for α energies of 5.5 and 6.5 MeV. Three major neutron groups were observed in the spectrum, which correspond to the ground and first two excited states of 12C. Measured data could only be explained by the TALYS calculation if reaction at more than one location within the target is considered for a given beam energy. The preferred locations are driven by the resonance energy levels existing in 13C. Neutron yield due to the 9Be breakup process was determined which is found to be 12.6 ± 0.2% and 18.4 ± 0.5% of the total reaction cross-section for 5.5 and 6.5 MeV respectively.

Shielding mass estimation model for gas-cooled space nuclear reactors

The shielding is expected to be “mass-effective” for space nuclear power systems (SNPs) because it accounts for a significant portion of the total weight. Hence, the shield mass should be estimated rapidly but accurately especially in the early stage of SNPs’ design. In this paper, a simple mass estimation model was developed based on the preliminarily optimized shield design for the Jupiter Icy Moons Orbiter (JIMO) reactor. In the model, the reactor-induced 1 MeV equivalent silicon neutron fluence and photon absorbed dose at the unshielded payload were predicted by a circular surface source hypothesis at the core bottom, and the proportional coefficients were determined by the Monte Carlo calculation. Moreover, the attenuation factors of the shield were derived by solving the neutron/gamma coupled transport process within the shield body, and the correlations were established between the attenuation effect and the shield thicknesses. Afterwards, when given the radiation limits at the payload, the shield geometry and associated mass could be easily determined. Several detailed calculations corresponding to the variation of shield design were used to verify the estimation model. Those variations include the thickness of the Be-B4C layers, the effective density of the B4C and the number of the shielding layers. Additionally, the potential changes of the reactor such as the neutron energy spectrum and the structural materials in the core were also considered to evaluate the applicability of the model. Totally, the estimation model agrees with the detailed calculation within 15 % in a wide range, which suggests the model can be incorporated into the system code to help optimize the overall performance of the SNPs.

Passive single-sphere multi-detector spectrometry system for profiling of neutron fields in the thermal to 20 MeV energy range

This study explores the development and application of a custom-designed single-sphere multi-detector spectrometer (SSMDSS) for characterizing neutron fields. Stray neutrons in reactor and accelerator facilities present complex, energy-diverse fields. Accurately assessing these fields is crucial for personnel safety. The SSMDSS system is designed to address this challenge, offering a compact and passive dosimetry solution for constrained workspaces. We studied the performance of SSMDSS in two environments: radio-nuclide based neutron sources and stray neutron fields at a research reactor hall. Our investigation demonstrates that SSMDSS provide reliable neutron energy distributions. The performance of the SSMDSS was compared with commercially available ROSPEC system. SSMDSS exhibits superior resolution due to its design, leading to a more accurate measurement of low-energy fluence when compared with ROSPEC. This distinction is reflected in various measured parameters, including average energy and fluence-to-dose conversion coefficients. Inside the reactor hall, SSMDSS measured a fluence-to-dose conversion coefficient of 121 pSv.cm2 compared to ROSPEC's 83 pSv.cm2. Additionally, SSMDSS measured a higher average energy (0.25 MeV) compared to ROSPEC (0.14 MeV). While ROSPEC remains a valuable tool for real-time data collection in dynamic fields with high gamma backgrounds, SSMDSS presents itself as a viable alternative for neutron spectrometry in neutron work fields due to its compact design and passive dosimetry capabilities. This makes it particularly suitable for deployment in constrained workspaces with potential stray neutron exposure.

Influence mechanism of lithium surface oxidation on neutron yield of lithium target for accelerator-based neutron source

Lithium is generally adopted as the target to generate epithermal neutrons based on 7Li(p,n)7Be nuclear reaction for the application of accelerator-based boron neutron capture therapy (AB-BNCT). The stability of neutron yields and neutron energy spectrum are key factors for the therapeutic effects of the AB-BNCT. Owing to the active chemical properties of lithium, the surface oxidation formation of lithium may influence the soundness of the target system and the stability of neutron yields. Experimental and simulation methods were performed for a better understanding of the passivation layer formation after air exposure and the influence of the passivation layer on the neutron yield and energy spectrum. On one hand, X-ray photoelectron spectroscopy was adopted to study the chemical states of lithium after air exposure. On the other hand, particle and heavy ion transport code system (PHITS) was used to evaluate the effect of surface chemical states variation on neutron yield and energy spectrum. The simulation results indicate that the formation of Li2O with a thickness of 2 μm could mainly influence the neutron yields at the neutron emission angle of 20–90° with a maximum reduction of ∼ 10.4%.

Neutron spectroscopy for Boron Neutron Capture Therapy beams characterization

Boron Neutron Capture Therapy (BNCT) is a therapeutic treatment for malignant tumors that utilizes the nuclear reactions that happen when thermal neutrons are captured by boron-10 atoms to selectively destroy designated cells. Boron-10 atoms are biochemically accumulated inside the tumor target, which is then irradiated with thermal neutrons.In recent years, the possibility to obtain accelerator based intense neutron beams has given a boost to the diffusion of BNCT also in Europe, removing the need of nuclear reactors. In this contest, the monitoring and characterization of the epithermal neutron beams dedicated to BNCT becomes an important issue.The directional neutron spectrometer called NCT-WES (Neutron Capture Therapy Wide Energy Spectrometer) is a single moderator neutron spectrometer composed of a polyethylene cylinder embedding six semiconductor-based detectors positioned at different depths along the cylinder axes. The position of the six detectors is studied in order to maximize the response of each one in a selected neutron energy range. The unfolding of the six simultaneous readings allows to reconstruct the incoming neutron energy spectrum as in a parallelized Bonner Sphere System. A cylindrical collimator situated in the front of the spectrometer makes the instrument sensitive to neutrons coming only from a given direction, which allows to exclude the contribution of the room scattered radiation. The experimental validation of the spectrometer, obtained through several measuring campaigns, is reported and discussed.

Optimization of photoneutron source for use in subcritical reactors

This work is an effort to maximize the number of output neutrons of a photoneutron target for 100 MeV incident electrons. The work steps were; to select the appropriate material, select the appropriate shape, and find the optimized dimensions. The simulations of this work were done with MCNPX2.6 simulation code. At first, U-235 was selected as the best material for the target, among some heavy atoms from the number of outlet neutron point of view. Then the best shape for the target was selected from geometric shapes that have been considered for the target and the dimensions of the target were optimized. After these parts, the energy spectrum of outlet neutrons was estimated and after that, the deposited energy of neutrons, electrons, and photons in the target was estimated and drowned with Tec plot.

Evaluation of neutron cross sections of 48Ti based on the Unified-Monte-Carlo-B method* *Supported by the National Key Research and Development (R&D) Program (2022YFA1602403) and Continuous Support Basic Scientific Research Project BJ010261223282

A cross section evaluation of neutron induced reactions on 48Ti is undertaken using the Unified Monte Carlo-B (UMC-B) approach. The evaluation concentrates on estimating the covariance and the use of the UMC-B allows avoiding the deficiencies of linear regression brought by the traditional least squares method. Eight main neutron and charged particle emission reactions from n+48Ti in the fast neutron energy region below 20 MeV are studied in this work. The posterior probability density function (PDF) of each neutron cross section is obtained in a UMC-B Bayesian approach by convoluting the model PDFs sampled based on model parameters and the likelihood functions for the experimental data. Nineteen model parameters including level density, pair corrections, optical model and Kalbach matrix element parameter are stochastically sampled with the assumption of normal distributions to estimate the model uncertainty. The Cholesky factorization approach is applied to consider potential parameter correlations. Finally, the posterior covariance matrices are generated using the UMC-B generated weights. The new evaluated results are compared with the CENDL-3.2, ENDF/B-VIII.0, JEFF-3.3, TENDL-2021 and JENDL-5 evaluations and differences are discussed.

Developing and using international standards for neutron ambient dose equivalent rate meters

Measurements of neutron ambient dose equivalent rates are challenging because there are no perfect instruments that cover the entire 9 orders of magnitude neutron energy range (thermal to 20 MeV). Hence quantifying performance requirements with the international standard IEC 61005 for neutron ambient dose equivalent rate meters presents unique challenges. This paper discusses the requirements of this standard, which have been refined and improved over the last 20 years. Examples of type testing results for the constancy of ambient dose rate response, variation of the response due to neutron energy, response time, and various environmental disturbances are examined. IEC 61005 provides manufacturers with internationally acceptable requirements and provides consistent test methods for compliance with stated performance requirements.

Developed concrete reinforced by waste of lead fishing weights and steel nails as gamma rays and neutrons shields for nuclear power reactors

Wastes of steel nails WSN and wastes of lead fishing weights WLFW were successfully recycled in concrete to enhance its ability to attenuate neutrons and gamma rays. A series of studies on the effect of partially replacing coarse aggregates of concrete with different percentages (5, 10, and 20%) of WSN or WLFW on the density, slump, compressive strength, and attenuation of neutrons and gamma rays were implemented. The compressive strength of WSN-reinforced concrete ranged between 36.2 and 45.2 MPa, while that of WLFW-reinforced concrete ranged between 28.6 and 45.1 MPa. The ability of WSN- or WLFW-reinforced concrete to attenuate neutrons and gamma rays was tested by using two types of neutron energies; slow and total slow neutrons; and three gamma rays energies 661.64, 1173.23, and 1332.51 keV. Slow and total slow neutrons macroscopic cross-section, linear and mass attenuation coefficients of gamma rays, and half value layer for both types of nuclear radiation were obtained. The results showed that there is a significant improvement in the slow and total slow neutron attenuation efficiency of the considered concrete when reinforced by WSN or WLFW. On the other hand, due to the increase in the atomic number and density of both Fe present in WSN and Pb in WLFW, the attenuation ability of the studied gamma ray energies increased. The WLFW-reinforced concrete showed better value in the gamma rays and neutrons attenuation parameters compared to the WSN-reinforced concrete. Hence, the considered concretes reinforced by WSN or WLFW are a distinguished choice as radiation shields in nuclear power reactors, in addition to their adaptability to high concentrations of waste.

EURADOS intercomparison for whole-body neutron dosemeters from 2012 to 2022

Within the EURADOS (European Radiation Dosimetry Group) self-sustained programme of regular intercomparisons (ICs) for individual monitoring, three campaigns had been performed in 2012, 2017 and 2022 to assess the performance of neutron personal dosemeters routinely used to measure personal dose equivalent, Hp(10). The irradiation schemes were established to provide participants with useful information regarding linearity, reproducibility and response of their dosimetry systems at different energies and angles of incidence. Small alterations were made in irradiation plans for subsequent exercises. Combinations of standard calibration fields and simulated workplace fields were employed, using bare and moderated neutron sources as well as monoenergetic and thermal neutron fields. Neutron energies used to evaluate dosemeter performance ranged from thermal to several MeV. Dose values varied between 0.3 mSv and 15 mSv. The number of participating laboratories in the three campaigns was stable: 31 in IC2012n, 32 in IC2017n and 29 in IC2022n. The dosimetry systems could have been categorized according to their detection principle into albedo and track dosemeters, or a combination of both. In 2012, a category ‘other’ had to be introduced to consider active personal dosemeters and fission track detectors. In 2012, participants were asked to provided results in a two-step procedure: (i) with no information on the radiation fields, and (ii) after receiving additional simplified a priori information on the energy distribution of the neutron fields to enable correction for dosemeter response. In 2017 and 2022, based on a one-step procedure, however, participants might have requested a priori information upon registration, which was reported in the certificate of participation. Intercomparison results were analysed against the acceptance criteria of the ISO 14146:2018 standard, although there were no standardized performance requirements for neutron personal dosemeters at the time of IC2012n. The paper compares the results of the three exercises, identifies major problems for individual monitoring services and shows trends in dosimetric performance. In all three intercomparisons, a few systems had shown problems related to calibration. Data do not allow to draw a universally applicable conclusion with regard to potentially superior performance of one particular dosimetry system.

Measurements of 232Th(n,2n) 231Th reaction cross-sections in the neutron energy of 12–18.5 MeV

Measurements of 232Th(n,2n) 231Th reaction cross-sections in the neutron energy of 12–18.5 MeV

Method of the known cross sections for calibration of the fast neutron spectrometer with a single-crystal stilbene based detector

The present work is devoted to implementation of the stilbene-based neutron spectrometer energy calibration method. The results of experiments with portable neutron generators and 238PuBe source and scattering materials with known cross sections are used for this method. It is shown that the submitted method makes it possible to carry out fast neutron spectrometry in the energy range from 1 to 15 MeV with the uncertainty of the unfolded neutron energy no more than 200 keV. Neutron spectra unfolding was carried out based on the measured spectra and a Geant4 simulated response matrix. Unfolded spectra were compared with the literature data and reference spectra.

Fast neutron production at the LNL Tandem from the 7\\documentclass[12pt]{minimal} \\usepackage{amsmath} \\usepackage{wasysym} \\usepackage{amsfonts} \\usepackage{amssymb} \\usepackage{amsbsy} \\usepackage{mathrsfs} \\usepackage{upgreek} \\setlength{\\oddsidemargin}{-69pt} \\begin{document}$$^7$$\\end{document}Li(14\\documentclass[12pt]{minimal} \\usepackage{amsmath} \\usepackage{wasysym} \\usepackage{amsfonts} \\usepackage{amssymb} \\usepackage{amsbsy} \\usepackage{mathrsfs} \\usepackage{upgreek} \\setlength{\\oddsidemargin}{-69pt} \\begin{document}$$^{14}$$\\end{document}N,xn)X reaction

Fast neutron beams (En>\\documentclass[12pt]{minimal} \\usepackage{amsmath} \\usepackage{wasysym} \\usepackage{amsfonts} \\usepackage{amssymb} \\usepackage{amsbsy} \\usepackage{mathrsfs} \\usepackage{upgreek} \\setlength{\\oddsidemargin}{-69pt} \\begin{document}$$_n > $$\\end{document} 1 MeV) are of relevance for many scientific and industrial applications. This paper explores fast neutron production using a TANDEM accelerator at the Legnaro National Laboratories, via an energetic ion beam (90 MeV 14N\\documentclass[12pt]{minimal} \\usepackage{amsmath} \\usepackage{wasysym} \\usepackage{amsfonts} \\usepackage{amssymb} \\usepackage{amsbsy} \\usepackage{mathrsfs} \\usepackage{upgreek} \\setlength{\\oddsidemargin}{-69pt} \\begin{document}$$^{14}N$$\\end{document}) onto a lithium target. The high energy models for nuclear collision of FLUKA foresee large neutron yields for reactions of this kind. The experiment aimed at validating the expected neutron yields from FLUKA simulations, using two separate and independent set-ups: one based on the multi-foil activation technique, and the other on the time of flight technique, by using liquid scintillator detectors. The results of the experiment show clear agreement of the measured spectra with the FLUKA simulations, both in the shape and the magnitude of the neutron flux at the measured positions. The neutron spectrum is centered around the 8 MeV range with mild tails, and a maximum neutron energy spanning up to 50 MeV. These advantageous results provide a starting point in the development of fast neutron beams based on high energy ion beams from medium-sized accelerator facilities.