Neutron Beam Research Articles

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 γ-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 (En = 30 and 250 MeV) at Osaka University’s Research Center for Nuclear Physics to measure γ-rays originating from the neutron-oxygen reaction with a high-purity germanium detector. Multiple γ-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 γ-ray signal and background templates, we measured cross sections for each observed γ-ray component. The results will be useful to validate neutron models employed in the on-going and future water Cherenkov experiments.

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.

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

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.

Characterization of the PGNAA neutron beam of Isfahan MNSR through the calculations and different measurement methods

In this work, the significant characterization of the neutron beam used in the PGNAA facility of the Isfahan MNSR research reactor has been determined using both calculation and measurement. The important parameters of the beam include flux and energy distribution of neutrons, cadmium ratio, thermal neutron content (TNC), and neutron-to-gamma ratio (n/γ). In this regard, Monte Carlo simulation and experiments were implemented to find these characteristics. The experiments have been carried out through the relative and absolute activation methods using irradiation of indium foil and mica in combination with uranium foil. Moreover, the condition of the neutron beam has been investigated in terms of contamination with fast neutrons using the germanium triangle method. The results showed that the MNSR has acceptable characteristics for performing PGNAA analyses despite its low power.

High repetition-rate 0.5 Hz broadband neutron source driven by the Advanced Laser Light Source

Neutron beams are an essential tool to investigate material structure and perform nondestructive analysis, as they give unique access to element composition, thus ideally complementing density analysis allowed by standard x-rays investigation. Laser-driven neutron sources, though compact and cost-effective, currently have lower average flux than conventional neutron sources, due to the limited repetition rate of the lasers used so far. However, advancements in laser technology allow nowadays to address this challenge. Here, we report results obtained at the Advanced Laser Light Source characterizing stable production of broadband (0.1–2 MeV) neutrons produced at a high repetition rate (0.5 Hz). The interaction of laser pulses of 22 fs duration and 3.2 J on-target energy with 2-μm-thick tantalum targets produced protons in the Target Normal Sheath Acceleration (TNSA) regime up to 7.3 MeV. These protons were subsequently converted into neutrons by (p,n) reactions in lithium fluoride (LiF). Activation measurements and bubble detectors were used to characterize neutron emissions, with a neutron fluence of up to ∼1.4×105 neutrons/shot/sr and energies mainly between a few hundred of kilo-electron volt and 2 MeV. The total neutron yield was ∼5×105 neutrons/shot. This paves the way for numerous applications, e.g., in homeland security, materials science, or cultural heritage.

Improving flame resistance and shielding properties for cross-linked polyethylene (XLPE) using nanoclay filler

Polymers play an essential role in both industry and medical fields due to their diverse and adaptable properties. In this work, prepared crosslinking polyethylene (XLPE) samples with hydrophilic bentonite nanoclay fillers (H2Al2O6Si) at concentrations of 0, 1, 2.5, 4, and 5 wt% enhance their flame-retardant and radiation shielding efficiency. The research investigated flame retardancy and thermal stability parameters. The XLPE/H2Al2O6Si nanocomposite polymer sheets were exposed to a collimated beam of fast neutrons using an Am/Be neutron source (5 Ci) and to gamma radiation using a 137Cs point source (5 μCi) to assess their radiation shielding properties. The study found that uniform dispersion of nanoclay particles enhanced the thermal properties of the composite, forming a char layer that acted as a barrier, slowing thermal decomposition and reducing the heat release rate. Limiting oxygen index (LOI) increased from 28% to 34%, and burning rate improved with higher nanoclay concentrations. Additionally, absorption and optical band gap calculations decreased with increasing filler concentrations. Radiation attenuation capabilities increased by approximately 40% for neutrons and 30% for gamma radiation compared to pure XLPE. The study concluded that incorporating nanoclay fillers into XLPE enhances its shielding capabilities and improves flame resistance properties, making the prepared samples suitable for various industrial applications.

Implementation of optically simulated luminescent dosimeter for quality control of gamma ray dose of an accelerator-based neutron source.

Neutron beams utilized for performing BNCT are composed of a mixture of neutrons and gamma rays. Although much of the dose delivered to the cancer cells comes from the high LET particles produced by the boron neutron capture reaction, the dose delivered to the healthy tissues from unwanted gamma rays cannot be ignored. With the increase in the number of accelerators for BNCT, a detector system that is capable of measuring gamma ray dose in a mixed neutron/gamma irradiation field is crucial. Currently, BeO TLDs encased in quartz glass are used to measure gamma ray dose in a BNCT irradiation field. However, this type of TLD is no longer commercially available. A replacement dosimetry system is required to perform the recommended ongoing quality assurance of gamma ray measurement for a clinical BNCT system. The purpose of this study is to evaluate the characteristics of a BeO OSLD detector system under a mixed neutron and gamma ray irradiation field and to assess the suitability of the system for routine quality assurance measurements of an accelerator-based BNCT facility. The myOSLD system by RadPro International GmbH was evaluated using the accelerator-based neutron source designed for clinical BNCT (NeuCure BNCT system). The readout constancy, linearity, dose rate effect, and fading effect of the OSLD were evaluated. Free-in-air and water phantom measurements were performed and compared with the TLD results and Monte Carlo simulation results. The PHITS Monte Carlo code was used for this study. The readout constancy was found to be stable over a month-long period and similar to the TLD results. The OSLD readout signal was found to be linear, with a high coefficient of determination (R2≥0.999) up to a proton charge of 3.6C. There was no significant signal fading or dose rate dependency. The central axis depth dose and off-axis dose profile measurements agreed with both the TLD and Monte Carlo simulation results, within one standard deviation. The myOSLD system was characterized using an accelerator system designed for clinical BNCT. The experimental measurements confirmed the OSLD achieved similar, if not superior to, the currently utilized dosimetry system for routine QA of an accelerator-based BNCT system. The OSLD system would be a suitable replacement for the current TLD system for performing routine QA of gamma ray dose measurement in a BNCT irradiation field.