Fast Neutron Source Research Articles

Comparison of fast-neutron-induced reaction cross-sections in [formula omitted], [formula omitted], [formula omitted], [formula omitted], [formula omitted], [formula omitted], and [formula omitted] in quasi-monoenergetic fields of p+Li and d+TiT neutron sources

This paper describes cross-section measurements of fast-neutron-induced reactions using the neutron activation method. The study employed two accelerator-driven fast neutron sources, each possessing distinct characteristics during irradiation of analogous sets of samples. Specifically, the neutron source at the NPI Řež relied on the p+Li(C) reaction, while MONNET at JRC Geel utilized the d+TiT(Ag) reaction for neutron production. Building upon prior experiments conducted at NPI, we present new cross-sections for reactions, some of which suffer from a dearth of experimental data. The results presented in this paper originate from samples of Al, Au, Bi, Co, NaF, and Y that were irradiated with neutrons at energies of 17.6(3) MeV, and 18.5(4) MeV at MONNET, and an energy of 22.5(8) MeV at the NPI. Following neutron irradiation, the samples were analysed using gamma-ray spectrometry. Cross-sections of several fast-neutron-induced reactions were determined on the basis of the measurement of neutron spectra and corresponding reaction yields. By comparing the results of analogous experiments conducted in different experimental setups, it becomes possible to assess systematic uncertainties. The obtained cross-sections may contribute to the further development of both advanced nuclear reactors and to a more precise fast-neutron dosimetry.

Exploring of the suitability of garnets for immobilizing nuclear waste and radiation shielding

In this study, the structural, chemical composition and gamma and nuclear shielding properties of some garnets were investigated. The experimental studies were made on the mass attenuation coefficients (MAC), linear attenuation coefficients (LAC), mean free path (MFP), half-value layer (HVL), effective atomic numbers (Zeff) and the radiation protection efficiency (RPE) in the high energy region. The fast neutron removal cross sections (ΣR), exposure buildup factors (EBF) were theoretically determined. The neutron absorption equivalent dose was measured of the garnet by 241Am–Be fast neutron source. It may be concluded that P1.21 has better photon shielding than the other samples investigated. The equivalent dose rate absorbed by the P1.21 garnet sample was found to be 36.80 % μSv/h and the P1.21 sample shows promise as a neutron shielding option. Also, the detection of inclusion minerals were made by EPMA and SEM analysis. It is observed that P1.21 sample is pyrope, P1.3 sample is andradite and P1.6 sample is grossular series. Also, according to ICP-MS results, it is possible to say that the P1.21 sample, which contains high amount of uranium, thorium, Y and Zr and has a high radiation absorption capacity, is a potential garnet structure for nuclear waste immobilization. Garnet's crystal structure and chemical durability enable safe containment and long-term storage of radioactive elements, facilitating their secure disposal.

Structural, thermal, and radiation shielding properties of B2O3-TeO2-Bi2O3-CdO-Tm2O3 glasses: The role of Tm2O3

This study presents a detailed investigation into the structural, thermal, and radiation shielding properties of a new glass composition, labeled as (50-x)B2O3 + 30TeO2 + 10Bi2O3 + 10CdO + x Tm2O3 (where x = 0, 1, 2, 3, 4, 5 mol %). Various radiation shielding parameters such as mass attenuation coefficient, linear attenuation coefficient, half-value layer, mean free path, effective atomic number, and absorbed equivalent dose rates for neutrons were determined through experimental measurements. The study also employed theoretical calculations to assess the exposure buildup factor and effective removal cross-section for neutrons. Additionally, the SRIM program was utilized to investigate mass stopping power and projected range for proton and alpha particles. The structural characteristics of the glasses were analyzed using XRD and FT-IR, while thermal properties were examined through Differential Thermal Analysis (DTA). FTIR analysis revealed distinctive bands corresponding to Bi—O, B—O—B, and Te—O bonds. Glass transition temperatures (Tg) increased with Tm2O3 content, ranging from 422 °C to 444 °C. The radiation shielding properties of bismuth boro-tellurite glasses showed that a dose of 1.2025 μSv/h emitted from a fast neutron source was absorbed by 35.1574 % in pure glass and 40.0391 % in glass doped with 5 mol% Tm2O3. Incorporating Tm2O3 into the glass matrix significantly improved the shielding and thermal properties, thus enhancing their potential for advanced high-energy radiation absorption. This investigation contributes to a deeper understanding of how Tm2O3 enhances the structural and thermal attributes of these glasses, positioning them as promising materials for radiation shielding applications.

Numerical intercomparison of PHITS and Geant4 Monte Carlo codes for fast neutron inelastic scattering applications

Fast neutron inelastic scattering is a promising non-destructive assay technique for various analytical applications. As an active neutron interrogation technique, its performance is a function of various different factors and parameters that require optimization. Monte Carlo simulation codes are indispensable for such tasks. However, the internal simulation routines implemented in such codes can rely on different physical models that can yield discrepancies in the simulation results. In this work we conduct an intercomparison of PHITS and Geant4 codes performance in application to fast neutron inelastic scattering simulations. The goal of this paper is twofold. First, we explain the differences in code configuration with respect to gamma and neutron transport, as well as internal simulation routines. Second, we conduct a performance assessment of the two codes using two different measurement configurations. One configuration consisted of a source of gamma rays in a broad energy range (100–9000 keV) and a CeBr3 detector. The other configuration consisted of a monoenergetic 2.5 MeV fast neutron source, Fe, Nd, Dy, B targets and a CeBr3 detector. Selected simulation configurations were chosen with a goal to compare the performance differences in neutron energy distribution, produced prompt gamma rays and energy deposition in CeBr3 detector between the two codes. Results of our study reveal a good coherence of both codes performance in the application of fast neutron inelastic scattering simulations. The simulation geometries and observed differences are described in detail.

Radiation shielding effectiveness, structural, and mechanical properties of HDPE/B4C composites reinforced with Fe2O3-Al2O3-Al-Fe fillers

The fast neutron and gamma-ray attenuation properties of pure high-density polyethylene and its composites were investigated, including 90 HDPE/10 B4C (wt%) and 60 HDPE/10 B4C/30X (wt%) (Where X represents iron oxide, aluminum oxide, iron, and aluminum). The examined samples were analyzed using a scanning electron microscope, and their mechanical properties were assessed. The composite of 60 HDPE/10 B4C/30 α-Fe2O3 (wt%) shows better mechanical properties than those of other composites under investigation. 239Pu–Be was used as a source of fast neutrons with a neutron yield of 1.7 × 106 n s−1 and was detected by the Stilbene scintillator. Neutron removal cross-section with dependent parameters, the mean free path and half-value layer of the prepared composites, were calculated. Furthermore, the gamma-ray transmission through the produced composites has been examined. The composite containing 30% iron oxide exhibited superior shielding characteristics for neutrons and gamma rays compared to the other analyzed samples. The computed values of the shielding characteristics indicate that the produced composites are highly effective for shielding fast neutrons and gamma rays in radiation facilities.

The Contributions to Registration Efficiency of The Fast Neutron Reactions on The Nuclei of The Heavy Oxide Scintillators

The results of the study of the contributions of the interaction reactions of fast neutron sources of 239Pu-Be and 252Cf to the counting efficiency of registration by oxide scintillators CdWO4, ZnWO4, Bi4Ge3O12 and Gd2SiO5, presented. The amount of gamma quanta per input neutron emitted from final nuclei excited in the reactions of inelastic scattering (n, nʹγ)in, resonant scattering (n, n)res and capture (n, γ)res and radiation capture (n, γ)cap was measured. PMT R1307 operating in single-electron mode was used as a photodetector, the background rate was ~ 5*103 s-1. The measured efficiency ε for scintillators ø40x40 mm was 752 for ZWO, 532 for CWO, 37 for GSO, and 23 for BGO in "counts/neutron" units, measurement error rate ~ 3-5%. The formation of the detector response is influenced by the parameters of the scintillator nuclei, such as the values of the interaction cross sections in the resonance region, the density of nuclear levels of the final nuclei, the lifetime of excited nuclear states, the upper limit of the resonance region of the cross section, as well as the scintillation time and geometric parameters of the scintillators. A phenomenological model of the response of an oxide scintillator to fast neutrons is proposed.

Photonuclear reaction in 67Zn

Mono-energetic [Formula: see text]-beams ([Formula: see text][Formula: see text]eV) based on thermal neutron capture, in a nuclear reactor, using the Mn([Formula: see text]) reaction were utilized for generating a fast neutron source from Zinc, via the [Formula: see text]Zn([Formula: see text]) reaction. One of the incident [Formula: see text]-lines of the Mn source at [Formula: see text][Formula: see text]keV, photoexcites by chance a resonance level in [Formula: see text]Zn, with subsequent emission of neutrons at an energy of 191[Formula: see text]keV. The cross-section for this process was measured and found to be [Formula: see text][Formula: see text]mb with an intensity of the order of 104[Formula: see text]n/s. The angular distribution of the 191[Formula: see text]keV neutron group was also measured.

Investigation of the structural, mechanical, radiation and neutron shielding properties of the TeO2-B2O3-Li2O-MoO3-CuO glass system

In this study, the structural, mechanical, radiation and neutron shielding properties of (78-x)(0.4TeO2-0.6B2O3)− 20Li2O-xMoO3-2CuO glass system, were investigated. A variety of characterisation methods, including optical absorption spectroscopy, FTIR, EPR, and X-ray diffraction, were used to examine the structural characteristics. Molecular orbital coefficients were calculated from EPR and optical absorption measurements. EPR lines of both Cu2+ and Mo5+ ions were observed in the EPR spectra. The mechanical properties were assessed through experimental measurements, while the elastic properties were evaluated using the semi-empirical approaches of the Makishima and Mackenzie model, along with the bond compression model. The inclusion of 4% MoO3 led to noteworthy enhancements in both hardness and fracture toughness. Hardness increased from 4.364 to 4.664 GPa, and fracture toughness improved from 0.980 to 1.226 MPa.m1/2. The mass attenuation coefficients (MAC), mean free paths (MFP), half-value layers (HVL), and effective atomic numbers (Zeff) were measured in the energy range of 59.54–661.62 keV. The measurement of equivalent dose rates resulting from the absorption of fast neutrons was conducted utilizing a 241Am-Be fast neutron source with an activity of 10 mCi. The exposure buildup factor (EBF) and the fast neutron removal cross section (ΣR) were therotically calculated. The potential suitability of glasses for nuclear security applications was assessed by evaluating their proton and alpha shielding properties. The study found that glasses containing higher amounts of TeO2 and MoO3, within the TeO2-B2O3-Li2O-MoO3-CuO composition, demonstrated enhanced capabilities for shielding against radiation and neutrons.

A neutron source for the study of biological objects formed from superficially touching cones made of borated spheroplastics

At the Prometheus medical accelerator with a proton beam energy of 225 MeV, a source of fast and epithermal neutrons was constructed and measurements of neutron dose profiles at the output of the neutron channel were carried out using the BDMN-100 detector. A heavy NaI target was used to produce fast neutrons. Together with the research laboratory of the Central Laboratory of Avangard JSC, a new protective material against neutrons called wikineutron was developed, with a different percentage of 10B. This new material has been studied many times at the Prometheus proton accelerator and the Pakhra electron accelerator. Based on the developed new protective materials against neutrons, a shadow protection was formed, made in the form of surface contacting cones, forming a channel of fast and epithermal neutrons. Fast neutrons can be used for remote therapy. Also, a neutron beam can be used to study biological objects and cells. It is also possible to use a neutron source from the Prometheus accelerator for the treatment of superficial tumors. The developed neutron channel can be used for medical work on the creation of new radiopharmaceuticals containing boron and other highly absorbing elements. The developed neutron source is a compact low-power source of therapeutic neutrons, which can be used for the treatment of superficial types of cancer. The main goal of the work was: the formation, based on the developed neutron-absorbing materials, of a neutron channel, which has a simple design and can be used for boron-neutron capture therapy and nadepithermal therapy; creation on the neutron channel of a beam of epithermal neutrons and supra-epithermal neutrons to assess the effectiveness of the use of radiopharmaceuticals.The developed neutron-absorbing materials made it possible to create a neutron channel of epithermal neutrons and supra-epithermal neutrons for therapy and the development of gold-based radiosensitizers.

Development of a High Sampling Rate Data Acquisition System Working in a High Pulse Count Rate Region for Radiation Diagnostics in Nuclear Fusion Plasma Research

In this study, a high sampling rate data acquisition system with the ability to provide timestamp, pulse shape information, and waveform simultaneously under a sub megahertz pulse counting rate was developed for radiation diagnostics for magnetic confinement nuclear fusion plasma research. The testing of the data acquisition system under the high pulse counting rate condition using real signals was performed in an accelerator-based deuterium-deuterium fusion neutron source (Fast Neutron Source) at the Japan Atomic Energy Agency. We found that the pulse counts acquired by the system linearly increased up to 6 × 105 cps, and the count loss at 106 cps was estimated to be ~10%. The data acquisition system was applied to deuterium-deuterium neutron profile diagnostics in the deuterium gas operation of a helical-type magnetic confinement plasma device, called the Large Helical Device, to observe the radial profile of neutron emissivity for the first time in a three-dimensional magnetic confinement fusion device. Time-resolved measurements of the deuterium-deuterium fusion emission profile were performed. The experimentally observed radial neutron emission profile was consistent with numerical predictions based on the orbit-following models using experimental data. The data acquisition system was shown to have the desired performance.

On-line determination of the water cut and chlorine impurities in crude oil using a pulsed beam of fast neutrons

A neutron-based sensor has been developed able to measure the water cut and the chlorine content in crude oil. The sensor is based upon the use of a pulsed fast neutron source. Two different cases were studied. Case ‘‘A’’: chlorine detection with no water present and case ‘‘B’’: water cut measurements with no chlorine present. The minimum detection limits for case A and case B were found to be (62 ± 6) mg/L and (2 ± 2) vol%, respectively.

Monte Carlo simulation-aided design of a thermal neutron generator system from 241Am-Be isotopic sources

Collimated thermal neutron beams are obtained from neutron extraction channels in nuclear reactors for various applications in research and technology, such as neutron imaging techniques (neutron radiography, neutron radioscopy, neutron tomography, and neutron-based autoradiography). Practical setups for neutron radiography using ion beams from particle accelerators and radioisotopic sources of fast neutrons have been also developed. However, only radioisotopic sources enable autonomous and transportable thermalization systems that can produce thermal neutron collimated beams. This work presents the performance results for a prototype of a compact system that generates a collimated beam of thermal neutrons using low-activity isotopic 241Am-Be sources. It was designed with the aid of Monte Carlo simulation using the PHITS v 3.17 program. Experimental measurements of the fluence of the neutron beam produced by the built prototype showed good agreement with the simulated values by the Monte Carlo method.

Sludge interface measurement in the storage tank utilizing neutron backscattering technique: A field experiment

Modern industrial plant operations commonly necessitate a measurement system of the liquids level interface in the oil storage tank. Various advanced-level indicators almost meet the demand, but these may not be in online condition and are less accurate due to sedimentation in the bottom tank. Sedimentation or sludge known as a petroleum hydrocarbons complex emulsion, contained water and solid particles generated during distribution, storage, or even production. The neutron backscatter technique can help the maintenance system of the tank or vessel by measuring sludge levels from outside the wall. This paper presents a non-intrusive method for identifying the interface between two materials in the crude oil storage system. A fast neutron source (241Am–Be) with activity 1 Ci (Ci) and a portable backscattering neutron detector are used for storage tank measurement. The storage tank has an outer diameter (OD) of 34 m and a height of 11 m. Time measurement is 2 s for each position, starting from 0 m to 11 m. The slow neutron intensity can be correlated to the hydrogen concentration inside chemical compounds. The interface between sludge and oil is identified at 27 cm from the bottom plate. The method has the potential for non-destructive measurement of sludge in a refinery facility.

Measurement of the applicability of various experimental materials in a medically relevant reactor neutron source Part One: Material characteristics acting as a carrier for boron compounds during neutron irradiation

A 100 kW thermal power pool-type light water reactor and Pu(Be) as a fast neutron source were used to determine the appropriate carrier for irradiating boron-containing samples with neutron beams. The tested materials (carriers) were subjected to neutron beams in the reactor's tangential channel. The geometrical arrangement of experimental facilities relative to the neutron beam trajectory, as well as the effect of sample thickness on the count rate, were investigated. The majority of the detectable charged particles emitted by the neutron beam's interaction with tested materials and the detector's detecting layer are protons (recoiled hydrogen) and particles generated in nuclear reactions (protons and alpha particles), respectively. Stopping and Range of Ions in Matter (SRIM) software was used to do theoretical calculations for the range of expected released particles in various materials, including human tissue. The results of measurement and calculation are in good agreement. According to experiments and theoretical calculations, the number of protons emitted by tissue-like materials may commit a dose comparable to that of boron capture reactions. Furthermore, the range of protons is significantly larger than that of alpha particles, which most probably changes dose distribution in healthy cells surrounding the tumor, which is undesirable in the BNCT approach.

Measurement of the applicability of various experimental materials in a medically relevant reactor neutron source part two: Study of H3BO3 and B-DTPA under neutron irradiation

Experiments related to Boron Neutron Capture Therapy (BNCT) accomplished at the Institute of Nuclear Techniques (INT), Budapest University of Technology and Economics (TUB) are presented. Relevant investigations are required before designing BNCT for vivo applications. Samples of relevant boron compounds (H3BO3, BDTPA) usually employed in BNCT were investigated with neutron beam. Channel #5 in the research reactor (100 kW) of INT-TUB provides the neutron beam. Boron samples are mounted on a carrier for neutron irradiation. The particle attenuation of several carrier materials was investigated, and the one with the lowest attenuation was selected. The effects of boron compound type, mass, and compound phase state were also investigated. To detect the emitted charged particles, a traditional ZnS(Ag) detector was employed. The neutron beam's interaction with the detector-detecting layer is investigated. Graphite (as a moderator) was employed to change the neutron beam's characteristics. The fast neutron beam was also thermalized by placing a portable fast neutron source in a paraffin container and irradiating the H3BO3. The obtained results suggest that the direct measurement approach appears to be insufficiently sensitive for determining the radiation dose committed by the Alpha particles from the 10B (n,α) reaction. As a result, a new approach must be used.

Li2B4O7–Bi2O3–ZrO4–CaWO4 glass system for neutron protection in neutron applications

In this study, new glass samples with a chemical composition of 40Li2B4O7+xBi2O3+(55–x)ZrO4+5CaWO4 (x = 10, 20, 30, 40, and 50 wt%) were fabricated using the melt quenching-annealing technique. Important neutron shielding parameters, such as the effective neutron removal cross-section, half-value layer, and neutron transmission factor, of all the new glass samples were calculated using the GEANT4 Monte Carlo simulation code. In addition, neutron dose measurements were performed using an 241Am–Be fast neutron source with an activity of 74 GBq and an average neutron energy of 4.5 MeV and a BF3 gaseous neutron detector. The results were compared with those of conventional shielding materials to determine the neutron absorption capabilities of the new glass materials. All the new glass samples were found to have excellent neutron shielding abilities. In particular, the GB5 sample had a high absorption capacity compared to the other samples. These new glass samples may be used as shielding materials in various radiation applications, such as generation of nuclear energy, radiation medicine, and the transport and storage of radioactive wastes.

Investigation of the radioprotective properties of Zingerone, Ibuprofen, and Sulfasalazine against neutron radiation damage

Nuclear radiation (neutron, gamma X rays, etc.) is widely used in power generation, industry, medicine, research, and agriculture. Exposure to ionizing or indirect ionizing radiation can cause mutagenetic cell death depending on the dose and duration. It is important to reduce the carcinogenesis of cancer patients undergoing radiotherapy treatment and people working in radiation fields. Neutrons have wide applications in neutron activation analysis (NAA), neutron capture therapy (NCT), neutron radiography (NR), neutron capture, neutron conversion doping (NTD) sample irradiation, and neutron diffraction (ND). In these applications, protective substances are needed to stop the neutrons from entering the body and to reduce their effects. In this work, the use of zingerone, ibuprofen, and sulfasalazine compounds as active ingredients in a neutron protective drug was investigated. Some important neutron attenuation parameters such as an effective removal cross section, mean free path, half-value layer, and transmission number were theoretically calculated with the Monte Carlo simulation GEANT4 code for fast and epithermal neutrons.The 241Am-Be fast neutron source and a BF3 gaseous detector were used to determine the neutron absorption ability of all of the samples. All of the results were compared with paraffin, which has a good shielding capacity. According to the results, it was found that every examined sample had a fast and epithermal neutron absorption capacity. It reported that the dose amount from the source was absorbed by the samples at rates of 33.37% (DAI1), 34.85% (DAI2), and 34.21% (DAI3). It was determined that the DAI2 sample in particular had a high absorption ability ratio compared to other samples. Because this sample have high hydrogen content than others samples. So, it was shown that these compounds can be used as active substances to make a radioprotective drug.

Determination of toxicity and radioprotective properties of bacterial and fungal eumelanin pigments

Purpose Determination of the protective property of melanin, an organic polymer class consisting of phenolic and/or indolic compounds isolated from bacteria and fungi, against fast neutron radiation. To show that these melanin samples, which also have antioxidant and metal chelating properties, can be used as an active ingredient for a drug to be developed against neutrons used in nuclear research and medicine. Materials and methods Bacterial and fungal media were prepared, and melanin pigments were produced and isolated. For molecular characterization of pigments, bacterial genomic DNA extraction, 16S rDNA gene amplification processes, and fungal genomic DNA extraction, ITS1, and ITS4 Gene Regions amplification were performed. The DEL assay was implemented to determine the genotoxicity properties of bacterial and fungal melanin pigments. Samples were prepared in a pad measuring 10 ml volume (60 × 15 mm) at a concentration of 0.2–1 microgram in 1% agarose gel for radiation-absorbed dose measurements. Absorption measurements were made using 241Am-Be fast neutron source and Canberra brand NP series BF3 gaseous detector to determine the neutron radiation absorption capacity of all samples. The results obtained to determine the absorption degrees of melanin samples were compared with paraffin and normal concrete, which are widely used in neutron radiation shielding studies. Results Melanin pigments were obtained using different bacteria and fungi strains. Afterwards, the fast neutron radiation absorption capacity of these purified pigments were determined. Compared to reference samples, these pigments were found to have slightly lower radiation absorbing ability. In addition to these experiments, cytotoxicity tests were carried out using the Yeast DEL assay technique to evaluate the potential for use of these organic pigments in fields such as medicine and pharmacology. According to the results obtained from the tests, it was determined that these melanin samples did not have any toxic effects. Conclusion It was determined that these melanin samples have the potential to be used as a radioprotective drug active substance to protect the tissues and cells of people exposed to neutron radiation after a nuclear accident or nuclear war. Giving a drug that will be developed by using these active ingredients before or after people are exposed to a radiation environment can provide great benefits.

Investigation of the neutron radiation protective properties of chlorophyll and carotenoid

In this study, we have investigated neutron radiation absorption properties of chlorophyll and carotenoid. Our aim is to show that chlorophyll and carotenoid, which are naturally found in every green plant, can be used to make a cream, medicine, paint, or shield material against neutrons. Neutron radiation is generally used in many applications such as diffraction and scattering experiments, material crystallography, diagnosis and therapy in medicine, cell DNA studies in biology, condensed matter physics, materials research and development, solid-state chemistry, mineralogy, geology, and the study and improvement of the structural properties of materials. In order to reduce or completely eliminate the damages that can be caused by direct or scattered neutrons by the people working in these studies or other people in the environment, materials with high shielding are needed to absorb neutrons. In order to understand that chlorophyll and carotenoid can be used in the production of neutron shielding materials, the neutron radiation absorption parameters of these materials were investigated. The parameters, which are widely used in neutron shielding studies such as an effective removal cross section, half value layer, mean free path, and neutron transmission factor (NTF) were estimated using the Monte Carlo simulation GEANT4 code for both fast and epithermal neutrons. In addition, the dose rates of all samples were determined by using an 241Am–Be fast neutron source. All theoretical and experimental results obtained to understand the absorption capacities of chlorophyll and carotenoid were compared with paraffin, which is widely used in neutron shielding applications.It is to be noted that determined that chlorophyll showed an effective shielding against fast neutrons and carotenoid showed an effective shielding against epithermal neutrons. Investigation revealed that these materials are alternate shielding materials against neutron radiation for the production of medicine, composite material, and paint to be developed for neutron radiation.

Reconstruction of fast neutron direction in segmented organic detectors using deep learning

A method for reconstructing the direction of a fast neutron source using a segmented organic scintillator-based detector and deep learning model is proposed and analyzed. The model is based on recurrent neural network, which can be trained by a sequence of data obtained from an event recorded in the detector and suitably pre-processed. The performance of deep learning-based model is compared with the conventional double-scatter detection algorithm in reconstructing the direction of a fast neutron source. With the deep learning model, the uncertainty in source direction of 0.301 rad is achieved with 100 neutron detection events in a segmented cubic organic scintillator detector with a side length of 46 mm. To reconstruct the source direction with the same angular resolution as the double-scatter algorithm, the deep learning method requires 75% fewer events. Application of this method could augment the operation of segmented detectors operated in the neutron scatter camera configuration for applications such as special nuclear material detection.