Neutron Beam Research Articles

Proposal for a Ramsey Neutron-Beam Experiment to Search for Ultralight Axion Dark Matter at the European Spallation Source.

High-intensity neutron beams, such as those available at the European Spallation Source (ESS), provide new opportunities for fundamental discoveries. Here, we discuss a novel Ramsey neutron-beam experiment to search for ultralight axion dark matter through its coupling to neutron spins, which would cause the neutron spins to rotate about the velocity of the neutrons relative to the dark matter halo. We estimate that experiments at the HIBEAM beamline with a 50m free flight path at the ESS can improve the sensitivity to the axion-neutron coupling compared to the current best laboratory limits by up to 2-3 orders of magnitude over the axion mass range 10^{-22} eV-10^{-16} eV.

Preparation and characterization of a new Gd2O3-epoxy composite for neutron shielding applications

The current study aims to introduce a new polymeric composite consisting of epoxy resin as the matrix and gadolinium oxide (Gd2O3) as the neutron adsorption ingredient. The shielding performance of the composite was assessed by neutron attenuation experiments with an Am-Be source and polyethylene moderator. The results of these experiments showed an appreciable agreement with the Monte Carlo simulations. Other characteristics of the composite, including mechanical strength, thermal stability, microtexture, and its chemical compositions, were examined using standard tensile test, thermogravimetric analysis, X-ray diffraction, scanning electron microscopy, static light scattering analyses, and Fourier-transform infrared spectroscopy (FTIR). The results indicated that the new composites offer appreciable neutron absorption properties so that samples with 0.5%, 2%, 5%, and 10% Gd2O3 content could reduce the neutron beam intensity by 54%, 63%, 66%, and 70% at a thickness of 4 cm.

Measurement of γ-Rays Generated by Neutron Interaction with 16O at 30 MeV and 250 MeV

Abstract Deep understanding of $\gamma$-ray production from the fast neutron reaction in water is crucial for various physics studies at large-scale water Cherenkov detectors. We performed test experiments using quasi-mono energetic neutron beams ($E_n = 30$ and 250 MeV) at Osaka University’s Research Center for Nuclear Physics to measure $\gamma$-rays originating from the neutron–oxygen reaction with a high-purity germanium detector. Multiple $\gamma$-ray peaks which are expected to be from excited nuclei after the neutron–oxygen reaction were successfully observed. We measured the neutron beam flux using an organic liquid scintillator for the cross section measurement. With a spectral fitting analysis based on the tailored $\gamma$-ray signal and background templates, we measured cross sections for each observed $\gamma$-ray component. The results will be useful to validate neutron models employed in ongoing and future water Cherenkov experiments.

Neutron transmission measurements for silica glass at the KURNS-LINAC

Silica glass has been used as a base and host material in vitrified radioactive waste and lithium glass scintillators for neutron detection because of its superb transparency, high heat resistance, and excellent chemical inertness. Therefore, an accurate total cross section of the silica glass is crucial to evaluate the criticality safety of vitrified wastes and understand the neutron response for lithium glass scintillators. This study performed neutron transmission measurements for silica glass using a pulsed neutron beam with the time-of-flight method at the Kyoto University Institute for Integrated Radiation and Nuclear Science − Linear Accelerator to provide an accurate total cross section in the thermal and epithermal energy range. We obtained the neutron total cross section of the silica glass in the energy region from 0.002–25 eV. The results were compared and discussed with previous results and evaluated data.

Non-invasive characterization of the manufacturing process of a Nuragic bronze statuette: a Neutron Imaging study

The Nuragic civilization (Sardinia, Italy, XVIII–VIII Cen. B.C) developed a flourishing bronze metallurgy. The production of Nuragic bronze figurines from Sardinia represents a rich historical archive that provides key information about the iconography, the metal production and casting techniques, and on the development of metallurgy in the Mediterranean basin. Since the question about their manufacturing method remains without definitive answer, the understanding of the Sardinian bronze metallurgy is essential to determine which manufacturing techniques were employed to produce complex bronze artefacts. In the frame of a wider research project relating to Nuragic bronzes, four artefacts, three anthropomorphic statuettes (a warrior, a priestess, and an offering figure), and one miniature of a basket, were made available by Museo Nazionale Preistorico “L. Pigorini” (Roma, IT). In this work we present the results of the analyses conducted on a bronze figurine depicting an iconic type of Nuragic figure: the Priestess. The analysis was performed using White Beam Neutron Tomography (NT) and Bragg Edge Neutron Transmission (BENT) at the Paul Scherrer Institut (PSI) (Villigen, CH). Neutron techniques are nowadays the only available approach for revealing, non-destructively and with good spatial resolution, the morphological and microstructural properties within the whole volume of solid cast metallic artefacts such as this bronze statuette. This work presents the result of a non-invasive analytical investigation on an archaeological bronze artefact, providing outstanding results: from a quantitative analysis of the composition to an in-depth morphological and microstructural analysis capable of unveiling details on the ancient casting methods of the statuette.

Development of an online neutron beam monitoring system for accelerator-based boron neutron capture therapy in a hospital.

Boron neutron capture therapy (BNCT) is a next-generation radiotherapy, utilizing both an external neutron beam and a -containing pharmaceutical. A compact accelerator for a high intensity neutron source was installed to conduct BNCT in a hospital. The dose administered to a patient was evaluated by measuring the proton beamcurrent. Neutron intensity should be monitored in real-time by measuring the neutrons emitted from the target during BNCT irradiation. This is crucial due to potential neutron target degradation. Online neutron beam monitoring systems are required for reliable measurements of the administered neutron dose. An online neutron beam monitoring system was developed to monitor neutron intensity irradiating on the patient at the National Cancer Center Hospital (NCCH). The neutron detector comprised a back-illuminated thin Si diode of 40- thickness and an ultrathin natural LiF neutron converter of 0.05- thickness. The neutron detector was installed on the neutron target unit, regardless of whether a patient was present, without any additional modifications to the setup. The response functions for high photon dose rates of upto 100 Gy/h were measured. The pulse heights were measured using the neutron beam monitor during BNCT neutron irradiation. Neutron temporal response measured using the online beam monitor was acquired and compared with the proton beam current and the measurements at a patient position. From this measurement at the patient position, the neutron fluence rate irradiating on a patient wasobtained. The neutron events were separated from the photon events. The neutron counting rates increased rapidly with the starting of proton beam irradiation and dropped to zero upon its termination. During intermittent drops and recoveries in the proton beam, the neutron beam monitor for counting rates responded quickly, synchronizing with the beam current. A scatter plot of the neutron counting rate and proton beam current indicated a good linear correlation. A direct relationship between the online neutron beam monitor's neutron counting rates and those of the patient neutron detector showed a good correlation coefficient of 0.84. A ratio of the both neutron counting rates showed a standard deviation of 6%. The correlation coefficient and standard deviation were improved to 0.94 and 1.5%, by re-binning the neutron temporal response with longer acquisition period than 1 s. Using the online neutron beam monitor, the neutron fluence rate was obtained from the direct relationship within 1.5%. Therefore, real-time monitoring of neutron intensity was achieved within the acceptable level as per the International Commission on Radiation Units and Measurementsreport. The online neutron beam monitoring system was developed to monitor the BNCT neutron beam intensity at NCCH. The temporal response of the neutron beam monitor was synchronized with the neutron counting rate at the patient position. Using the online neutron beam monitor, the neutron fluence rate irradiating on the patient can be monitored from the direct relationship. Fluctuation of the neutron beam intensity through BNCT irradiation and the degradation of the lithium target through the lifespan of the neutron target could be monitored using the neutron beammonitor.

Light Output Function and Pulse-Shape Discrimination Capability of p-Terphenyl Organic Scintillator in Wide Neutron Energy Range of 1.1 to 19 MeV

In this work, we studied the light-output properties, efficiency function, as well as the pulse-shape discrimination (PSD) capability of p-Terphenyl scintillator. The selected solid cylindrical scintillation detector has a thickness of 45 mm and a diameter of 45 mm. Recently presented studies of light-output functions have only been measured for low-neutron energies. Our motivation has been to determine the light output function for p-Terphenyl scintillator more accurately over a wider neutron energy range. The measurements have been carried out with mono-energetic neutron beams in the wide energy range from 1.1 to 19 MeV. The neutron–gamma spectrometric system which we developed has been used for the measurement. The input analog signal from the detector was digitized with a fast 12-bits analog to digital converter with a sampling frequency of 1 GHz. Measured data from the detector are processed into the gamma and neutron spectra. The accurate light output function for the p-Therphenyl scintillator has been calculated. The pulse-shape discrimination capability, as well as the detection efficiency, of a p-Terphenyl scintillator are lower in comparison with a NE-213 equivalent detector.

Neutron phase filtering for separating phase- and attenuation signal in aluminium and anodic aluminium oxide

Neutron imaging has gained significant importance as a material characterisation technique and is particularly useful to visualise hydrogenous materials in objects opaque to other radiations. Fields of application include investigations of hydrogen in metals as well as metal corrosion, thanks to the fact that neutrons can penetrate metals better than e.g. X-rays and are highly sensitive to hydrogen. However, at interfaces refraction effects sometimes obscure the attenuation image, which is used for hydrogen quantification. Refraction, a differential phase effect, diverts the neutron beam away from the interface in the image leading to intensity gain and intensity loss regions, which are superimposed to the attenuation image, thus obscuring the interface region and hindering quantitative analyses of e.g. hydrogen content in the vicinity of the interface. For corresponding effects in X-ray imaging, a phase filter approach was developed and is generally based on transport-of-intensity considerations. Here, we compare such an approach, that has been adapted to neutrons, with another simulation-based assessment using the ray-tracing software McStas. The latter appears superior and promising for future extensions which enable fitting forward models via simulations in order to separate phase and attenuation effects and thus pave the way for overcoming quantitative limitations at refracting interfaces.

Effect of neutron beam properties on dose distributions in a water phantom for boron neutron capture therapy.

From the viewpoints of the advantage depths (ADs), peak tumor dose and skin dose, we evaluated the effect on the dose distribution of neutron beam properties, namely the ratio between thermal and epithermal neutron fluxes (thermal/epithermal ratio), fast neutron component and γ-ray component. Several parameter surveys were conducted with respect to the beam properties of neutron sources for boron neutron capture therapy assuming boronophenylalanine as the boron agent using our dose calculation tool, called SiDE. The ADs decreased by 3% at a thermal/epithermal ratio of 20-30% compared with the current recommendation of 5%. The skin dose increased with the increasing thermal/epithermal ratio, reaching a restricted value of 14 Gyeq at a thermal/epithermal ratio of 48%. The fast neutron component was modified using two different models, namely the 'linear model', in which the fast neutron intensity decreases log-linearly with the increasing neutron energy, and the 'moderator thickness (MT) model', in which the fast neutron component is varied by adjusting the MT in a virtual beam shaping assembly. Although a higher fast neutron component indicated a higher skin dose, the increment was <10% at a fast neutron component of <1 × 10-12Gycm2 for both models. Furthermore, in the MT model, the epithermal neutron intensity at a fast neutron component of 6.8 × 10-13Gycm2 was 41% higher compared with that of 2 × 10-13Gycm2. The γ-ray component also caused no significant disadvantages up to several times larger compared with the current recommendation.

Accurate measurements of delayed neutron data for reactor applications: methodology and application to 235U(nth,f)

Large inconsistencies still exist in nuclear data libraries regarding the kinetic parameters of delayed neutron (DN) precursors. As an example, there is a 17% gap between ENDF-B/VIII.0 and JEFF-3.3 on the average lifetime T1/2 of DN precursors from thermal fission of 235U. This parameter is of major importance for reactivity predictions of nuclear reactors in nominal or accidental configurations. In this context, CEA is actively participating to the ALDEN project (Average number and Lifetime of DElayed Neutrons) which aims at providing the nuclear data community with new data sets of DN from thermal and fast neutron induced fission of various actinides. A dedicated experimental setup was designed and optimized for that purpose and is presented in this paper. It consists of a “long counter” detector containing 16 proportional counters filled with 3He, embedded in a high density polyethylene matrix. The detector surrounds a fissile target prepared in the form of a miniature fission chamber, containing a few hundreds of micro-grams of fissile material. This set-up is connected to fast and efficient neutron shutters that can produce step-irradiations of variable durations. The equations driving the DN counting following step-irradiations of the fissile target are established and discussed in the perspective of DN yield or group parameter measurement. A comprehensive analysis of the different steps of data reduction is detailed: dead time characterization, Region of Interest (ROI) determination, absolute and relative efficiency calibration, fission rate estimation, irradiation time and background determination, DN decay curve production and physical parameter fitting. Following a prototype experiment performed in 2018 at the PF1b cold neutron beam line of Institute Laue Langevin (ILL, Grenoble, France), we discuss here the analysis of two campaigns occurring in 2019 and 2021 in which significant improvements were achieved in terms of background minimization, counting statistics and fission rate determination. The achievements of this work are the measurement of the delayed neutron emission per fission for the thermal neutron induced fission of 235U, estimated at (1.625 ± 0.010) % and the group parameters leading to an estimated lifetime of T1/2\\documentclass[12pt]{minimal} \\usepackage{amsmath} \\usepackage{wasysym} \\usepackage{amsfonts} \\usepackage{amssymb} \\usepackage{amsbsy} \\usepackage{mathrsfs} \\usepackage{upgreek} \\setlength{\\oddsidemargin}{-69pt} \\begin{document}$${T}_{1/2}$$\\end{document} = (8.87 ± 0.10) s. Those results are consistent with the values recommended by the IAEA/CRP work and they come with reduced uncertainties compared with previously published results.

Research on burnup depth measurement of spent fuel board based on thermal neutron imaging technology

The detection equipment for measuring the burnup depth in spent fuel rods is a vital component of the spent fuel management system. which can determine the burnup depth of spent fuel board and plays a crucial role in safeguarding the safety, economic efficiency, and structural integrity of the fuel assembly. This study introduces an innovative technical approach for assessing the burnup depth of spent fuel veneers, utilizing transmitted thermal neutron imaging technology. We have significantly enhanced the design of a thermal neutron moderator collimator, leading to remarkable improvements in the quality of the thermal neutron beam. Following moderation by the collimator, the ultimate thermal neutron injection rate at the designated sample location exceeds 103 n/cm2, with thermal neutrons comprising over 74% of the collimated neutron beam. This advanced measurement system enables us to obtain a detailed two-dimensional distribution map of thermal neutrons transmitted through spent fuel boards with varying burnup depths. By analyzing the grayscale intensity patterns in these maps, we can accurately evaluate the burnup degree within the simulated spent fuel plate. Furthermore, we establish a correlation between the transmitted thermal neutron count in the imaging field and the burnup depth of the spent fuel veneer. This allows for precise determination of burnup depth through the analysis of the two-dimensional distribution of transmission thermal neutron intensity. Our findings demonstrate the feasibility of a scheme for detecting the burnup depth in spent fuel boards based on transmission thermal neutron imaging technology, and obtained a linear relationship between the neutron transmission count and the burnup depth of the spent fuel plate, laying a solid theoretical foundation for future research and development of testing equipment to assess burnup in spent fuel veneers.

Retrospective analysis of treatment-positioning accuracy and dose error in boron neutron capture therapy using a sitting-position treatment system for head and neck cancer

The neutron beam in boron neutron capture therapy (BNCT) exhibits poor directionality and significantly decreasing neutron flux with increasing distance. Therefore, the treatment site must be close to the irradiation aperture. Some patients with head and neck cancer may benefit from a sitting-position setup. The study aim was to evaluate the treatment-positioning accuracy and dose error in sitting patients receiving BNCT. Thirty-two patients with head and neck cancer who underwent sitting-position BNCT at Southern Tohoku BNCT Research Center were included in the study. Horizontal (ΔX) and vertical (ΔY) errors were defined as the displacement between the treatment planning system (TPS) digital reconstructed radiograph and the pre-treatment X-ray image. Using in-house software, image matching was performed. The beam-axial directional (ΔZ) error was compared with the parameters entered into the TPS and the actual pre-treatment measured values. The translational-position error was reflected in the TPS’s patient coordinate system with respect to the reference plan. Re-dose calculations were performed to evaluate the effect of positional error on tumor and normal-tissue doses. The [ΔX, ΔY, ΔZ] DRR-CR mean ± 1SD were − 0.40 ± 2.0, 0.30 ± 2.3, and − 1.4 ± 1.5 mm, respectively. The Dmean and D98% tumor-dose errors were 1.22 % ± 1.44 % and 0.99 % ± 1.63 %, respectively. The D2% pharyngeal and oral mucosal-dose errors were 0.98 % ± 1.91 % and 1.21 % ± 1.78 %, respectively. The tumor- and normal-tissue dose errors were typically < 5 %. High-precision treatment was feasible in sitting-positioned BNCT.

- Beam Neutron Optimization for Boron Neutron Capture Therapy (BNCT) facility

- Beam Neutron Optimization for Boron Neutron Capture Therapy (BNCT) facility

Beams of electrons, gammas and fast neutrons at BINP

An overview of test beam facilities at Budker Institute of Nuclear Physics (BINP) is presented. Facility for generation of extracted beams of electrons was designed at BINP for testing of detector prototypes in HEP field. The electron beam energy is from 100[Formula: see text]MeV to 3.5[Formula: see text]GeV. Since 2011, the facility has been used successfully for testing various detector prototypes. On the base of VEPP–4M collider, it is possible to obtain gamma beams too. The energy scale and resolution of calorimeters can be calibrated using edges of Compton and bremsstrahlung spectra. A facility for boron–neutron capture therapy (BNCT) was upgraded to perform radiation tests with fast neutrons with the integral flux up to [Formula: see text][Formula: see text]n/cm2. In 2022, the experiment on the study of the radiation hardness of optical fibers for the laser calibration system of electromagnetic calorimeter and the CMS detector was carried out. It has been demonstrated for the first time that at the BINP it is possible to operate with such doses. It could be further used for the wide range of radiation test tasks related with the development of facilities for HEP.

Boron neutron capture therapy and environmental health

Boron Neutron Capture Therapy (BNCT) is an innovative cancer treatment modality that exploits neutron beams to specifically destroy malignant cells. Despite its assuring clinical benefits, BNCT familiarizes itself with unique environmental health considerations, including radiation safety, waste supervision, and potential ecological ramifications. This review investigates these aspects by appraising radiation safety protocols, waste disposal procedures, and environmental monitoring implementations associated with BNCT facilities. Effective radiation shielding and demanding safety protocols are necessary to protect occupational and public health from potential neutron radiation exposure. Waste management implementations must inscribe the treatment, handling, and disposal of radioactive and biological waste to anticipate environmental contamination. Long-term ecological monitoring is decisive for detecting and mitigating any adverse influences on surrounding ecosystems. The review also emphasizes the significance of public awareness and education in addressing environmental health concerns. By integrating case studies from various international perspectives and recognizing research gaps, this review aims to contribute comprehensive insights into the ecological health suggestions of BNCT and propose strategies for enhancing its sustainability. The findings accentuate the need for continuous improvement in safety measures, waste management, and public engagement to corroborate that BNCT can be safely and effectively implemented in a technique that protects both human health and the environment.

Measurement of fast neutron induced (n,γ) reaction cross-section of 68Zn, 96Zr, 121Sb and 123Sb in the energy range of 1 to 2 MeV

The (n,γ) reaction cross-section for the elements 68Zn, 96Zr, 121Sb and 123Sb, present in the reactor structural/shielding materials, was measured by neutron activation technique in the neutron energy region of 1–2 MeV as very limited data is available in this energy range. Further, the neutron spectrum peaks in this energy region for the fast breeder reactors and proposed accelerator driven sub-critical systems. The natural strontium (natSr) element was used as a neutron flux monitor by considering effective combined reaction cross-section for 86Sr(n,γ)87Srm and 87Sr(n,n′)87Srm reactions. The pellets of mixture of sample and monitor were irradiated by a quasi-mono energetic fast neutron beam, generated by 7Li(p,n)7Be reaction at FOTIA, Bhabha Atomic Research Centre, Mumbai, India. The activity of activation products was measured by off-line gamma-ray spectrometry using High Purity Germanium Detector (HPGe). The present data with improved uncertainty and covariance analysis enhance the cross-section data base for better constraining the evaluated data and theoretical models. The theoretical (n,γ) reaction cross-sections were calculated using TALYS 1.96, which could reasonably explain the present data with the Fermi gas level density prescription.

On neutron generator applications in Brazil: current panorama and perspectives

Techniques based on neutron beams are used in research, industry and medicine being especially suitable for the characterization of a wide variety of materials. Neutron radiation can be produced using nuclear reactors, isotopic sources, or particle accelerators. Since the number of reactors is in decline and isotopic sources are expensive and difficult to handle, neutron generator (NG) technology has experienced significant development. Neutron generators are safer than nuclear reactors and isotopic sources, and the use of NGs is increasing worldwide, including in Brazil. This article reviews the main applications of neutron radiation, neutron production techniques, and specifically the technologies used in neutron generators. An overview of the utilization of these techniques in Brazil is presented, along with studies on acquisition and start-up costs of neutron-generating equipment and perspectives for future utilization in various areas.

New physics at a neutron beam dump

We find a new utility of neutrons, usually treated as an experimental nuisance causing unwanted background, in probing new physics signals. The new physics signals can either be radiated from neutrons (neutron bremsstrahlung) or appear through secondary particles from neutron-on-target interactions, dubbed “neutron beam dump.” As a concrete example, we take the FASER/FASER2 experiment as a “factory” of high-energy neutrons that interact with the iron dump. We find that if new physics particles had the same interaction strength to protons and neutrons, their production rate via neutron-initiated bremsstrahlung would be comparable to that via proton-initiated ones, in terms of the resulting flux and the range of couplings that can be probed. The neutron bremsstrahlung can be used, for instance, to probe dark gauge bosons with nonzero neutron coupling. In particular, we investigate protophobic gauge bosons and find that FASER/FASER2 can probe new parameter space. We also illustrate the possibility of neutron-induced secondary particles by considering axionlike particles with electron couplings. We conclude that the physics potential of FASER/FASER2 in terms of new physics searches can be greatly extended and improved with the inclusion of neutron interactions. Published by the American Physical Society 2024

Demonstration of shape analysis of neutron resonance transmission spectrum measured with a laser-driven neutron source

Laser-driven neutron sources (LDNSs) can generate strong short-pulse neutron beams, which are valuable for scientific studies and engineering applications. Neutron resonance transmission analysis (NRTA) is a nondestructive technique used for determining the areal density of each nuclide in a material sample using pulsed thermal and epithermal neutrons. Herein, we report the first successful NRTA performed using an LDNS driven by the Laser for Fast Ignition Experiment at the Institute of Laser Engineering, Osaka University. The key challenge was achieving a well-resolved resonance transmission spectrum for material analysis using an LDNS with a limited number of laser shots in the presence of strong background noise. We addressed this by employing a time-gated 6Li\\documentclass[12pt]{minimal} \\usepackage{amsmath} \\usepackage{wasysym} \\usepackage{amsfonts} \\usepackage{amssymb} \\usepackage{amsbsy} \\usepackage{mathrsfs} \\usepackage{upgreek} \\setlength{\\oddsidemargin}{-69pt} \\begin{document}$$^{6}{\ extrm{Li}}$$\\end{document}-glass scintillation neutron detector to measure the transmission spectra, reducing the impact of electromagnetic noise and neutron and gamma-ray flashes. Output waveforms were recorded for each laser shot and analyzed offline using a counting method. This approach yielded a spectrum with distinct resonances, which were attributed to 115In\\documentclass[12pt]{minimal} \\usepackage{amsmath} \\usepackage{wasysym} \\usepackage{amsfonts} \\usepackage{amssymb} \\usepackage{amsbsy} \\usepackage{mathrsfs} \\usepackage{upgreek} \\setlength{\\oddsidemargin}{-69pt} \\begin{document}$$^{115}\\,{\ extrm{In}}$$\\end{document} and 109Ag\\documentclass[12pt]{minimal} \\usepackage{amsmath} \\usepackage{wasysym} \\usepackage{amsfonts} \\usepackage{amssymb} \\usepackage{amsbsy} \\usepackage{mathrsfs} \\usepackage{upgreek} \\setlength{\\oddsidemargin}{-69pt} \\begin{document}$$^{109}{\ extrm{Ag}}$$\\end{document}, as confirmed through neutron transmission simulation. The spectrum was analyzed using the least-square nuclear-resonance fitting program, REFIT, demonstrating the possibility of using an LDNS for nondestructive areal-density material characterization.

Design of a mixed material moderator in a beam-shaping assembly for proton accelerator-based boron neutron capture therapy

Boron Neutron Capture Therapy is being promoted with the development of accelerator neutron sources, and many new accelerator-based BNCT facilities are being built. In Particle Accelerator Facility project of Sun Yat-sen University, we plan to build a terminal for BNCT research based on an 8 MeV, CW 3 mA proton accelerator. In this paper, we present a beam-shaping assembly for this proton accelerator with such low 24 kW beam power, using composite moderator materials composed of five elements: Mg, Al, F, O, and Li. The calculation result of FLUKA with ENDF/B and JENDL libraries shows that the epithermal neutron beam flux is 1.57×109n/cm2/s with the CW 3 mA proton beam. The fast neutron component and the gamma ray component under free-air condition are 1.49×10−13Gy∙cm2 and 8.12×10−14Gy∙cm2 respectively, in line with IAEA-TECDOC-1223 design recommendations. The thermal analysis shows that the maximum temperature of beryllium target is 706.5 K, and the structure materials of BSA do not melt.