Neutron Response Research Articles

Optimize neutron minimum detectable dose of OSL composite neutron detectors based on Al2O3:C and 6LiF

The n/γ ratio and minimum detectable dose (MDD) are two crucial parameters in the measurement of neutrons in a mixed field comprising both neutrons and gamma rays. In this study, four OSL neutron detectors based on 6LiF and very sensitive α-Al2O3:C powder with grain sizes of less than 100 and 38 μm, respectively, were developed. The first use of the detector pair comprised single-crystal α-Al2O3:C detectors and neutron-sensitive OSL composite detectors represents a pioneering advancement in the field. The n/γ ratio and MDD at various integration times were first investigated. For the four neutron detectors, the neutron response of the neutron detectors with a mixture of 1:1 mass ratio of the α-Al2O3:C powder with grain sizes of less than 100 μm and 6LiF exposed to the D2O moderated 252Cf source was the highest. When the integration time was 0.5 s, the MDD of the neutron detectors with a mixture of 1:1 mass ratio of the α-Al2O3:C powder with grain sizes of less than 100 μm and 6LiF was 6.1, 41.0, and 56.1 μSv, respectively, when irradiated by a calibrated D2O moderated 252Cf source, a bare 252Cf source and a bare 241Am–Be source. A comparison of the results indicates that the composite OSL neutron detectors based on Al2O3:C and 6LiF of this study have demonstrated a notable improvement in MDD.

Development of a compact active neutron monitor to measure the H∗(10): Design, simulation and validation

The rapid advancement of radiation technology underscores the need for effective radiation protection and monitoring. Neutron dose equivalent (NDE) meters play a crucial role in measuring the ambient dose equivalent, H∗(10), in neutron radiation environments. However, traditional NDE meters, while effective, tend to be bulky and less suitable for use in anisotropic neutron fields and confined spaces. This study presents the development of a compact, portable NDE meter designed to overcome these challenges. The device, featuring a cylindrical design with a diameter of 14.8 cm, a length of 30.5 cm, and weighing under 4 kg, is optimized for ease of use in constrained spaces. It incorporates a BF3 thermal neutron detector encased within a high-density polyethylene moderation assembly, calibrated to replicate ICRP-74 dose conversion coefficients. The design was optimized using Monte Carlo simulations using FLUKA, emphasizing neutron response uniformity and effective moderation. Experimental validation in standard neutron reference fields confirmed the accuracy of the simulated performance, with dose rate estimates deviating by less than 8% from reference values. The NDE meter's response was consistent with that of commercially available devices, showing relative energy response variations of less than 20% for both 241Am-Be and 252Cf sources. It demonstrated reliable energy response up to 5 MeV and consistent angular response (up to ∼ 60°), highlighting its potential for practical radiation protection applications in diverse and constrained environments.

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.

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.

Determination of neutron spectrum based on artificial neural network using liquid scintillation detector EJ-301.

This paper focuses on the neutron spectrum measurement using a liquid scintillation detector, where the neutron spectrum could be identified and unfolded from the light output distribution of the EJ-301 liquid scintillation detector through a linear artificial neural network (ANN). The response functions of the EJ-301 detector for monoenergetic neutron sources, as well as the light outputs, have been simulated and calculated by Monte Carlo procedure FLUKA. The linear ANN was trained and tested through the simulated data, where response functions were set as the input of ANN and the corresponding neutron spectra were output. Therefore, the neutron spectrum-unfolding model was created. This spectrum-unfolding model was tested through the light outputs induced by monoenergetic neutrons and the random superposition of them. Unfolding results show that this model could identify the information of the neutron spectrum accurately from the light outputs of a liquid scintillation detector. Moreover, the EJ-301 detector was used to measure the radioactivity of 252Cf, and the pulse height distribution induced by neutrons was derived through the charge-comparison method to remove the influence of gamma rays. The measured pulse height distribution was unfolded by the trained model, and measured results show that the unfolded neutron spectrum of 252Cf was consistent with the reference one. This paper presents the feasibility that the unknown neutron spectrum could be identified and confirmed through a linear neural network trained by simulated monoenergetic neutron response functions, which could be a candidate of choice for the determination of the neutron spectrum.

A Generic Implementation of the D1S Methodology for Dose Rate Assessment by Monte Carlo Codes in Fusion Applications

Dose rate assessment is of utmost importance in fusion reactors because of the maintenance operations that these facilities will require. The standard way to perform this assessment in ITER is use of the Direct 1-Step (D1S) methodology, which consists of performing neutron transport simulation, material activation, and decay photon transport simulation in one step and thus in one simulation only. Usually, implementation of the D1S methodology requires changing the source files in a reference Monte Carlo code. The purpose of the present work is to develop an easy-to-implement method for Monte Carlo codes to help calculate the dose rate in fusion reactors. To do so, the proposal is to act on nuclear data only and not on source files of the simulation codes. This is done by replacing prompt photon production in evaluation files by suitable decay photon production, taking into account radioactive decays of radionuclides and irradiation history. This study was to be applied first on the TRIPOLI-4® Monte Carlo code for the sake of simplicity. TRIPOLI-4 is the reference code for particle transport at CEA. The verification and validation process relies first on a comparison to the reference Rigorous 2-Step (R2S) methodology and then on an experiment, the so-called Frascati Neutron Generator (FNG) dose experiment. The nuclear database to be changed is of the ENDF-6 format, a recurrent format in neutronics studies. The analysis studied both neutron and photon responses to check if the simulation was performing normally in a physical way and to compare the results with references provided by simulations based on the R2S methodology or by the FNG dose experiment. The simulations have proven to be in good agreement with the experimental results.

Diamond based neutron detector in high temperature environments of 200 °C

The stability of the neutron response function and the detection efficiency of a sCVD diamond detector in the temperature range up to 200 °C are demonstrated in this paper. A CIVIDEC neutron detector was simultaneously heated with an AmBe neutron source. This enabled the measurement of the spectrum of the deposited energy in the diamond sensor, i.e. the neutron response function of the detector and the detection rate, which in combination represents the neutron flux as a function of temperature. The measured temperature stability of the neutron response function of the detector and detection rate demonstrates the suitability of sCVD diamond detectors for harsh environments, such as encountered in geodetic applications and nuclear fusion research.

Buoy‐Based Detection of Low‐Energy Cosmic‐Ray Neutrons to Monitor the Influence of Atmospheric, Geomagnetic, and Heliospheric Effects

AbstractCosmic radiation on Earth responds to heliospheric, geomagnetic, atmospheric, and lithospheric changes. In order to use its signal for soil hydrological monitoring, the signal of thermal and epithermal neutron detectors needs to be corrected for external influencing factors. However, theories about the neutron response to soil water, air pressure, air humidity, and incoming cosmic radiation are still under debate. To challenge these theories, we isolated the neutron response from almost any terrestrial changes by operating a bare and a moderated neutron detector in a buoy on a lake in Germany from July 15 to 02 December 2014. We found that the count rate over water has been better predicted by a theory from Köhli et al. (2021, https://doi.org/10.3389/frwa.2020.544847) compared to the traditional approach from Desilets et al. (2010, https://doi.org/10.1029/2009wr008726). We further found strong linear correlation parameters to air pressure (β = 0.0077 mb−1) and air humidity (α = 0.0054 m3/g) for epithermal neutrons, while thermal neutrons responded with α = 0.0023 m3/g. Both approaches, from Rosolem et al. (2013, https://doi.org/10.1175/jhm‐d‐12‐0120.1) and from Köhli et al. (2021, https://doi.org/10.3389/frwa.2020.544847), were similarly able to remove correlations of epithermal neutrons to air humidity. Correction for incoming radiation proved to be necessary for both thermal and epithermal neutrons, for which we tested different neutron monitor stations and correction methods. Here, the approach from Zreda et al. (2012, https://doi.org/10.5194/hess‐16‐4079‐2012) worked best with the Jungfraujoch monitor in Switzerland, while the approach from McJannet and Desilets (2023, https://doi.org/10.1029/2022wr033889) was able to adequately rescale data from more remote neutron monitors. However, no approach was able to sufficiently remove the signal from a major Forbush decrease event on 13 September, to which thermal and epithermal neutrons showed a comparatively strong response. The buoy detector experiment provided a unique data set for empirical testing of traditional and new theories on Cosmic‐Ray Neutron Sensing. It could serve as a local alternative to reference data from remote neutron monitors.

Unfolding of mono-energy neutron spectra using artificial neural network based on LMBP training algorithm

In this work, neutron spectra are unfolded using artificial neural networks (ANNs). The neutron response of the NE213 scintillator detector, characterized by the pulse height distribution, is calculated to obtain the necessary data for unfolding the energy spectrum. This is achieved using both analytical response functions and response functions generated by the MCNPX-PHOTRACK code. In this query, the Levenberg-Marquardt method (LMM), which has a high computational speed in the learning method, is used to train the network. The performance of the ANN for unfolding the neutron energy spectrum of the NE213 scintillation detector was evaluated by comparing its results to the established Gravel method. The ANN method consistently produced spectra with a single peak closely matching the incident energy, while the Gravel method showed additional peaks and distortions. Quantitative analysis revealed a lower relative energy peak difference (indicating higher accuracy) for the ANN method compared to Gravel, particularly when noise was introduced into the data. These results suggest that ANNs offer a more robust and accurate approach for neutron spectrum unfolding.

Development of trench-shaped microstructure thermal neutron detector using α-Al2O3:C based on optically stimulated luminescence

Optical stimulated luminescence technique is increasingly used in neutron personal dosimetry, due to its capability of fast readout and potential for re-estimation of absorbed dose. The design configuration and performance of the microstructure optically stimulated luminescence neutron detector (MOSLND) using α-Al2O3:C was reported. To improve thermal neutron response, the MOSLND is structured with a matrix of microstructure trenches into the α-Al2O3:C substrate and backfilled with 6LiF. To obtain an optimum trench geometric parameter, several Monte Carlo simulations were performed using Geant4 code. The simulation results show that relatively high energy deposition can be realized for small cell dimensions with deep trenches, and is significantly increased compared with that of simple planar designs. Based on the results of calculation and practical limitation, the device was produced and tested using D2O-modrated 252Cf. The optimized minimal detectable dose (MDD) corresponding to signal integration time of 2 s is 7.2 μSv. The experimental results indicate that MOSLND can be a potential solution for personal neutron dosimetry.

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

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

Transparent Glass Composite Scintillator with High Crystallinity for Efficient Thermal Neutron Detection

AbstractThe sensitive and rapid detection of thermal neutrons holds significant importance in various fields such as energy utilization, medical treatment, and national defense. However, the available thermal neutron scintillator is difficult to reach this target, mainly limited by the optical and scintillating performance. Herein, the in situ crystallization strategy of glass to construct scintillation glass composites for efficient thermal neutron detection is proposed. The congruent crystallization of the hybridized alkali earth silicate glass system may not only achieve high crystallinity, but also will keep the refractive indexing matching between the precipitated crystal and precursor glass phases. The prepared glass composites feature high luminescence efficiency, optical transmission and excellent neutron response properties. These factors collectively contribute to the robust neutron scintillation performance with a light output of ≈36 000 photons/neutron, which represents the highest value among glass‐based scintillators. Moreover, the composite configuration brings about the additional function of thermal neutron and γ‐ray discrimination. By using this composite scintillator, the neutron detector is constructed and demonstrates its application for detecting thermal neutron and distinguishing it from γ‐ray in an online way. The studies prove that glass composites with high crystallinity are expected to be promising candidates for a new generation of multifunctional neutron detectors.

Impact of fuel temperature on nuclear core design calculations

The operation of a nuclear power plant relies on precalculated nuclear design predictions based on core calculations of various reactor states. The fuel temperature is a crucial factor in determining the reactor fuel behavior, but assessing the temperature variation in a fuel pellet taking into account neutron transport is challenging. Detailed simulation of the temperature behavior within the fuel pellet can be obtained by coupling of Monte Carlo neutron transport codes with thermal-hydraulics solvers. However, this approach is not practical for standard nuclear design calculations, and computationally cheaper and faster methods must be used. In nuclear core simulators, a concept of a single "effective temperature" that yields the same neutron response as in the case of the actual temperature shape is mainly applied. This paper evaluates various fuel temperature models used in nuclear core simulation calculations, ultimately recommending a new effective temperature model that considers the burnup correction.

A comparison of the neutron detection efficiency and response characteristics of two pixelated PSD-capable organic scintillator detectors with different photo-detection readout methods

We characterize the performance of two pixelated neutron detectors: a PMT-based array that utilizes Anger logic for pixel identification and a SiPM-based array that employs individual pixel readout. The SiPM-based array offers improved performance over the previously developed PMT-based detector both in terms of uniformity and neutron detection efficiency. Each detector array uses PSD-capable plastic scintillator as a detection medium. We describe the calibration and neutron efficiency measurement of both detectors using a 137Cs source for energy calibration and a 252Cf source for calibration of the neutron response. We find that the intrinsic neutron detection efficiency of the SiPM-based array is (30.2 ± 0.9)%, which is almost twice that of the PMT-based array, which we measure to be (16.9 ± 0.1)%.

Response of LaBr2.85Cl0.15:Ce, LaBr[formula omitted]:Ce and NaI:Tl crystals to fast and thermal neutrons

This communication reports the measurement of direct response for fast and thermal neutrons of a newly marketed scintillation crystal LaBr2.85Cl0.15:Ce, along with the widely used LaBr3:Ce and NaI:Tl crystals. It continues our investigation on the possible utility of LaBr2.85Cl0.15:Ce crystal for the detection of high-energy γ-rays and neutrons. This is the first report of the neutron response of the LaBr2.85Cl0.15:Ce crystal. The neutron response was measured for the three detectors using a calibrated Am-Be source, which emits both neutrons and γ-rays. To eliminate the events due to the interaction of the 4.4 MeV γ-ray emitted from Am-Be, the γ-ray response of all three crystals was simulated using the Monte Carlo simulation toolkit GEANT4. Events due to 2.22 MeV γ-ray produced by the capture of thermal neutrons in the paraffin moderator were also eliminated with the aid of simulations. In addition, the response due to fast neutrons was also simulated using GEANT4 and compared with the measured spectrum. This is the first report of the simulation of fast neutron response of LaBr2.85Cl0.15:Ce using GEANT4. The fast neutron detection efficiency of the three detectors has been obtained from the measured spectra.

Neutron spectrum response of ST401 scintillators with different thicknesses

In the measurement of pulsed neutrons in the MeV energy range, plastic scintillators are one of the most widely used materials, and their neutron energy spectrum response is the key data required for pulsed neutron energy spectrum measurement. Base on the time of flight (TOF) method, the neutron energy spectrum response of ST401 plastic scintillator with 5 different thicknesses from 0.5 to 10 mm were measured for the 0.5 to 100 MeV energy range on the white neutron source (WNS) beamline of the China Spallation Neutron Source (CSNS). The effects of in-beam gamma rays, the slow component of scintillators produced by the gamma flash and the pulse width of the neutron source on the measurement of neutron spectrum response were analyzed. Due to the boundary effect of the finite volume of the scintillator, the neutron energy spectrum response curves of ST401 with different thicknesses are approximately logarithmic, and proton escape is the main reason for the deviation of the curve from linearity. The thicker the scintillator, the higher neutron energy that deviates from linearity.

Spectral analysis techniques in measuring neutron-induced gamma production cross-section

In neutron reaction cross-section measurements, the prompt gamma ray method is a method of obtaining cross-section data by measuring the characteristic gamma rays emitted by a nuclear reaction, thereby avoiding the interference generated by competing reaction channels. However, the prompt gamma ray method is an on-line experiment with abundant background sources, high background counts of the obtained experimental spectra, and numerous interferences such as weak peaks, overlapping peaks, Compton scattering peaks, and neutron effect peaks of Ge in HPGe, which cause the difficulty in analysing the on-line experimental spectra and the high uncertainty in the results. In this work, we study and summarise the spectrum analysis techniques of the prompt gamma ray method that can be used for measuring the neutron cross-section, and comprehensively consider the physical processes of the formation of different characteristic peaks of the prompt gamma ray method, so as to reduce the uncertainty of calculating the net area of the effect peaks in the process of on-line experimental spectrum processing. The Compton edge, weak peaks, overlapping peaks, and the neutron response peaks of the HPGe detector on-line experiment are discussed and analysed, and the net area of the effect peaks is accurately extracted by combining several reasonable functions to fit the total energy peak, the background, and the interferences. For the net area of weak peaks, this method can reduce the peak area selection caused fluctuation from 30% to less than 1%, and the difference between the fitted value of the net area and the theoretical value is comparable to the statistical uncertainty; for the overlapping peaks’ decomposition, the difference between the results obtained by this method and the theoretical value is significantly lower than 1%. The reliability of the spectral analysis method is simultaneously verified by efficiency curve analysis and goodness-of-fit calculation.

Simulation research on the characteristics of a minitype neutron reference radiation field

In order to research the characteristics of minitype neutron reference radiation field regulated by ISO 8529 series standards, Monte Carlo simulation method was applied to calculate the ratio of room-scattered neutron response to the direct neutron response, neutron fluence and energy spectra in the radiation field. Firstly, the characteristics of the field were investigated from the four physical factors as shielding materials, field sizes, shielding thicknesses and borated polyethylene with different boron contents. Secondly, the response of sphere instrument with different diameters in the field and the influence of instrument on the field were analyzed. The results showed that minitype neutron reference radiation field with a size of 2 m × 2 m × 2 m, a shielding material of 5 % borated polyethylene and a shielding thickness of 10 cm was expected to be used for the calibration of neutron measuring instrument.

Investigation of the phase diagram of the CsI-LiBr system and fabrication of the eutectic scintillator for thermal neutron detection

3He gas detectors have been commonly used in applied technology related to thermal neutron detectors; however, owing to the depletion of 3He, applications using solid-state scintillators are expanding. In recent years, our group has reported a novel Tl:CsI/6LiBr eutectic scintillator for thermal neutron detection, which has excellent light yield and (n,γ) discrimination ability. In practical applications of eutectic scintillators, achieving eutectic production precisely at the eutectic point is crucial to ensure a high production yield and the development of a uniform eutectic structure. However, the phase diagram of the CsI-LiBr system has not yet been reported. In this study, the phase diagram obtained from calorimetry was investigated. Furthermore, eutectic fabrication was performed at the eutectic point, and the neutron response was evaluated.

R&D of glass scintillator for nuclear radiation detection

In 2021, the Institute of High Energy Physics proposed a design of glass scintillator coupled with SiPM as a new solution for the next generation calorimeter, to explore the application of glass scintillators in high energy physics and nuclear radiation detection. The Large Area Glass Scintillator Collaboration Group was established to research and develop a glass scintillator with high density, high light yields and fast decay time. Through continuous optimization, the glasses have excellent scintillation performance with a light yield of 1000 ph/MeV and a density of 6 g/cm3. Moreover, the neutron response of the glasses was investigated, and different high-energy particles can be distinguished by signal amplitude. In addition, the radiation resistance of different glasses was tested under proton beam. All the glasses appeared opaque and produced a high radioactive background, because Gd element interacts with proton to produce radionuclides with high activity and long half-life.