Neutron Flux Research Articles

R-matrix analysis of 22Ne(α, n)25Mg and 22Ne(α, γ)26Mg reactions

Abstract The reaction rates of $^{22}$Ne($\alpha$, n)$^{25}$Mg and its competing channel $^{22}$Ne($\alpha$, $\gamma$)$^{26}$Mg control the production of neutron flux for weak $s-$process nucleosynthesis in low mass AGB stars and in massive stars with $M\geq 10M_\odot$. The temperature range of interest for these reactions lies between $0.2- 0.4$ GK. However, the rates of these reactions are poorly constrained at these temperatures due to uncertainties in the nuclear properties of a number of resonance states in the compound nucleus $^{26}$Mg, lying within the Gamow window. The present work reports a full $R$-matrix evaluation of the $^{22}$Ne($\alpha$, n)$^{25}$Mg and $^{22}$Ne($\alpha$, $\gamma$)$^{26}$Mg reaction rates using updated nuclear data of $^{26}$Mg states. Previous rate evaluation by Adsley~\textit{et al.} and $R-$matrix calculations of Wiescher~\textit{et al.} were limited by using narrow resonance approximations and omission of the resonances below $E_r=705$ keV, respectively. In this work, the $R-$matrix fit to the available $^{22}$Ne($\alpha$, n)$^{25}$Mg reaction data is performed by including the contributions of previously neglected resonances below E$_r$=705 keV and considering the interference effects. 
The ($\alpha$, n) reaction rate from the present $R$-matrix evaluations is noticeably higher than the narrow resonance approximation calculations in the temperature range $0.1-0.3$ GK. 
In particular, the present ($\alpha$, n) reaction rate is significantly higher ($8-3.7$ times) compared to Adsley~\textit{et al.} at $0.2-0.3$ GK and $\approx 2$ times greater than Wiescher~\textit{et al.} at $0.3$ GK. The estimated reaction rate ratio of ($\alpha$, n) to ($\alpha, \gamma$) in the relevant temperature window $0.2-0.8$ GK indicates that the production of neutrons for the $s$-process is more likely than the radiative alpha capture reaction, compared to the previous estimate by Adsley~\textit{et al.}

Clinical evaluation of performance, stability, and longevity of an accelerator system designed for boron neutron capture therapy utilising a beryllium target.

Boron neutron capture therapy for recurrent head and neck cancer has been approved as an insurance covered treatment in Japan since June 2020. The Kansai BNCT Medical Center has installed the NeuCure® BNCT system developed by Sumitomo Heavy Industries. This system utilises a beryllium target with a 30 MeV proton beam to generate neutrons. To date, the center has treated over 300 patients and the system is also used for quality control and fundamental research experiments. Information on the stability and longevity of the system is important for facilities and hospitals considering installation of a BNCT system in the future, particularly after an exchange of a major component, such as the target. The beam output (neutron and gamma ray) was measured before and after the exchange of the target. The thermal neutron, fast neutron and gamma ray distribution inside a water phantom was evaluated and the results showed no significant change in the beam quality after the exchange of the target. Furthermore, no significant change in the neutron flux as a function of increasing number of accumulated protons on the target was observed up to a value of just under 400 mAh. The NeuCure® BNCT system has been shown to be highly stable and the frequency of target replacement was much less than lithium target-based accelerators.

Neutron-induced backgrounds in coaxial cables on the South Pole neutron time-of-flight detector at the National Ignition Facility

As neutron yields increase at fusion facilities, a universal symptom the community must deal with is MeV neutron-induced backgrounds in cables running to diagnostics. On the first Gain >1 plasmas in the world, the National Ignition Facility (NIF) neutron time-of-flight (nToF) diagnostic registered significant cable backgrounds that compromised key performance measurements. The South Pole nToF is uniquely located inside the NIF Target Bay shield walls, ∼18 m from the fusion source, and consequently has long coaxial cable runs (>20 m) that see significant neutron fluence. The resulting neutron-driven current in the cable is comparable to the downscattered neutron signal, compromising the downscattered ratio (DSR) measurement. We have characterized this background with a series of on-shot tests and developed a background subtraction technique to mitigate these effects. The background subtracted DSR results are validated against zirconium activation measurements, indicating that we have successfully reclaimed high-quality data output. The ion temperature measurement is found to not be affected by this background. Alternative approaches to addressing neutron-induced cable backgrounds are presented for potential future hardware upgrades.

A dual method using SMELS for neutron flux self-monitoring and quality control in k0-NAA

A dual method using SMELS for neutron flux self-monitoring and quality control in k0-NAA

Investigation of Borated Polyethylene as a Neutron Shielding Material using Monte Carlo Simulation Code

The use of neutron shielding materials is essential in nuclear reactors to ensure safety and minimize radiation exposure. Conventional materials like lead and concrete pose environmental concerns, leading to the need for more sustainable alternatives. In this study, we have investigated the usage of borated polyethylene as an eco-friendly neutron shielding material using the Monte Carlo simulation code, FLUKA. Borated polyethylene slabs of varying thicknesses were evaluated for neutron attenuation, focusing on key parameters such as mass attenuation coefficient, mean free path, and half-value layer (HVL). Simulation results showed that borated polyethylene achieved a mass attenuation coefficient of 0.150 cm²/g at 0.025 MeV, which is 25% higher than that of conventional concrete. The mean free path for thermal neutrons was 6.67 cm, while the half-value layer was 3.34 cm, indicating effective neutron shielding, especially for thermal neutrons. Increased material thickness led to significant reductions in neutron fluence and transmission, with neutron fluence reduced to 5.0E+5 n/cm² at 15 cm thickness. These findings suggest that borated polyethylene is highly effective in attenuating neutron radiation and offers a more sustainable alternative to traditional materials.

A matrix effect correction method for fissile nuclear material mass measurement by delayed neutrons

Active neutron interrogation (ANI), extensively employed in nuclear safeguards, is sensitive to the presence of special nuclear materials (SNMs). However, the interaction of the matrix material with neutrons weakens the precision of fissile mass measurements by the ANI system. Therefore, it is paramount to ensure that the evaluation of the fissile mass remains unaffected by the matrix and enhances the assay performance of the ANI system. The present work proposes a matrix correction method based on the neutron flux monitor group (NFMG) response to tackle this issue. Based on the varying influence of different matrices on the neutron energies and fluxes, the NFMG response can be used to quantify the matrix effects. This allows the method to enable the identification of matrices already present in the database and has the potential to compensate for unknown matrix effects. To validate the applicability and accuracy of this method, a Shuffler system model and various matrix compositions were developed using the Geant4 toolkit. The results demonstrate that the improved simulated annealing (SA) algorithm exhibits excellent stability when confronted with varying enrichments and distributions of U3O8 materials. When the threshold is set at α ≥ 0.90, the improved SA algorithm achieves a successful identification rate of 90.4% for matrices. Simultaneously, using the response equation, the average relative deviation of the corrected 235U mass is no more than 7%. For the unknown matrix, the correction capability of this method relies primarily on the construction of the reference database and the response equation. In most cases, the average relative deviation of the corrected 235U mass is less than 28%.

Built-in physics models and proton-induced nuclear data validation using MCNP, PHITS, and FLUKA – Impact on the shielding design for proton accelerator facilities

The MCNP, PHITS, and FLUKA are general-purpose Monte Carlo (MC) radiation transport codes that are widely used for many real-world shielding problems at accelerator facilities around the world. Nuclear interactions are described in these codes by either built-in physics models, or tables with evaluated cross sections and secondary energy-angular distributions, or a combination of both. Over the decades, many code validation efforts have been made, owing to the availability of shielding benchmarks to test the physics models and nuclear data and to verify the accuracy of simulation codes.For high beam energy and high beam current accelerator applications, neutron emission through the vacuum pipe along the reverse direction of incident proton beam is an important factor for a shielding design in order to correctly assess the dose rates for workers and the structural materials of the accelerator and handle with the waste activated by the backscattered neutron fluxes. In this work, neutron-production cross sections and thick-target yield predictions from MC codes relying on physics models and nuclear data libraries are benchmarked against the experimental data, in order to assess their accuracy in predicting neutron emission and furthermore to assess the corresponding impact on shielding design.The results of this study demonstrate that the nuclear data libraries and physics models, which are not expected to give good results at lower energies (< 150 MeV) but are used anyhow when there is no nuclear data available or above the energy range where the data tables end in the so-called “mix-and-match” strategy, need further improvements. Among the investigated proton induced nuclear data libraries, JENDL-4.0/HE produces the most satisfactory agreement to experimental data for all target materials, but may still benefit from refinement. Concerning the physics models of the codes, FLUKA V4-4.0 has the best performance in terms of reproducibility of the experimental values. It is also shown that all discrepancies between the calculations and the experiments for the energy range < 10 MeV (which is dominant on the dose rates through the shield thickness), are up to factor of two. This might be considered as an acceptable figure as it is equivalent to a normal safety margin (x2) considered in shielding calculations of accelerator facilities around the world.

Investigation of neutron irradiated W/CuCrZr joints

This study investigates the effects of neutron irradiation on tungsten (W) and copper-chromium-zirconium (CuCrZr) joints under conditions mimicking the high neutron flux environment of a tokamak fusion reactor. Samples of W/CuCrZr joints were subjected to irradiation in the Belgian Reactor 2 (BR2) nuclear reactor at SCK CEN (Belgian Nuclear Research Centre) to simulate the intense neutron exposure characteristic for International Thermonuclear Experimental Reactor (ITER) and DEMOnstration power plant reactor (DEMO) operations. The primary objective was to evaluate changes in the mechanical properties and microstructure of these materials, which are critical for their potential use in plasma-facing components.It is revealed that a significant reduction in tensile elongation of the joint, indicating some degree of embrittlement, is observed after the irradiation. Importantly, this effect is independent of the irradiation temperature. Possible physical reasons for the observed phenomenon are discussed.

Collimated beams of fast neutrons at the NPI CAS

The Nuclear Physics Institute of the Czech Academy of Sciences operates multiple neutron sources that can produce neutrons with energies up to 33 MeV. Recently, a segmented collimator was constructed to facilitate research on collimated beams of fast neutrons. In front of the collimator, a new quasi-monoenergetic neutron source was built using accelerated protons interacting with a 2.5 mm thick beryllium target. The collimated beam provides a neutron flux of approximately 106 n/cm2/s at the standard measurement position.To determine the parameters of the collimated neutron beam, various measurement techniques were employed, including scintillator-based time-of-flight mode, proton recoil telescope, and activation detection through the (n,2-3n) reaction on a yttrium sample. Furthermore, Monte Carlo simulations were conducted to model the neutron transport through the collimator, and the results were subsequently compared to the experimental data obtained.

KATANA - water activation facility at JSI TRIGA, Part II: First experiments

Water as a primary coolant will play an important role in the performance of fusion reactors, as after being irradiated and activated, it causes an ionising radiation field throughout the facility. In direct support of ITER, the KATANA irradiation facility, which utilises a closed-water activation loop and serves as a well-defined and stable high-energy γ and neutron source, was commissioned in 2024 at the TRIGA Mark II research reactor at the Jožef Stefan Institute in Slovenia.A series of first experiments using the KATANA irradiation facility were performed to determine the operational characteristics of the KATANA, i.e. characterisation of the neutron flux within the irradiation component, dose rate measurements, spectrum characterisation of the activated water, and response to a change in reactor power & flow rate. The KATANA facility demonstrated the desired operational characteristics in terms of high and stable water flow rates and high activity values of the observed isotopes 16N, 17N and 19O, which is essential for minimising the experimental uncertainties.The first experiments performed at KATANA provided in-depth knowledge into the operation and capabilities of the facility and essential data to serve as a basis for further, more detailed/benchmark experiments.

Design and implementation of HDF5-format neutron nuclear database in RMC code

This study developed a new structured nuclear database to improve the readability and extensibility of the ACE (A Compact ENDF) nuclear database, save disk space, reduce memory usage, and enhance the computational efficiency of the Reactor Monte Carlo (RMC) code. A Python package was developed to store nuclear data in HDF5 format. Compared to the ACE database, the HDF5-format database shows significant improvements: an 80% reduction in disk space for continuous energy neutron data and a 60% reduction for neutron thermal scattering data. The HDF5-format database was implemented in RMC’s criticality calculation mode and validated through VERA benchmark problem 2B, demonstrating perfect agreement with the ACE results in keff and neutron flux counts. The Computational results indicate a 7.7% reduction in memory usage and a 20.1% improvement in computational efficiency with the HDF5-format database. Additional tests show that using the database at a single temperature point reduces memory usage by 5.4% and running time by 13.2%. At two temperature points, memory usage decreases by 35.2% and running time by 18.0%. The new data structure reduces temperature-independent redundant data and improves indexing efficiency, leading to greater savings with more temperature points. This development enhances performance of criticality calculation of RMC and addresses ACE database limitations.

Optimization study on the compact thermal neutron imaging system based on portable D-T neutron generator

The distribution of lithium in lithium-ion batteries is critical to their performance and lifespan. Studies have shown that thermal neutron imaging can effectively perform non-destructive testing of this distribution. However, most current thermal neutron imaging systems rely on expensive, bulky, and immobile nuclear reactors or spallation neutron sources, significantly limiting their application scenarios. Therefore, this study adopts a portable D-T neutron generator as the neutron source and uses Monte Carlo simulation methods to optimize the design of key components in the thermal neutron imaging system. Simulation results indicate that the optimized thermal neutron imaging system achieves a thermal neutron flux of 8.36 × 104 n·cm−2·s−1 and a thermal neutron content (TNC) of 14.4% at a collimation ratio of 50. Within an imaging range of Φ100 mm, the non-uniformity of the thermal neutron fluence is approximately 2.07%. Simulation imaging results of lithium and iron step wedges demonstrate that the compact thermal neutron imaging system designed in this study can effectively distinguish between iron and lithium of the same thickness. This finding provides an important reference for subsequent detection of lithium distribution within lithium-ion batteries.

Sensitivity improvement of a deuterium-deuterium neutron generator based in vivo neutron activation analysis (IVNAA) system.

Our lab has been developing a deuterium-deuterium (DD) neutron generator-based neutron activation analysis (NAA) system to quantify metals and elements in the human body in vivo. The system has been used to quantify metals such as manganese, aluminum, sodium in bones of a living human. The technology provides a useful way to assess metal exposure and to estimate elemental deposition, storage and biokinetics. It has great potential to be applied in the occupational and environmental health fields to study the association of metal exposure and various health outcomes, as well as in the nutrition field to study the intake of essential elements and human health. However, the relatively low sensitivity of the system has greatly limited its applications. Neutron moderation plays an important role in designing an IVNAA facility, as it affects thermal neutron flux in irradiation cave and radiation exposure to the human subject. This study aims to develop a novel thermal neutron enhancement method to improve the sensitivity of the in vivo neutron activation analysis (IVNAA) system for elemental measurement but still maintain radiation dose. Utilizing a compact DD neutron source, we propose a new and practical moderator design that combines high density polyethylene with heavy water to enhance thermal neutrons by reducing thermal neutron absorption. All material dimensions are calculated by PHITS, a general-purpose Monte Carlo simulation program. The improvement of the new design predicted by the Monte Carlo simulation for the quantification of one of the elements, manganese was verified by experimental irradiation of manganese-doped bone equivalent phantoms. For the same radiation dose, a 67.9% thermal neutron flux enhancement is reached. With only 4.2% increase of radiation dose, the simulated thermal neutron flux and activation can be further increased by 84.2%. A 100% thermal neutron enhancement ratio is also achievable with a 20% dose increase. The experimental results clearly show higher manganese activation gamma ray counts for each specific phantom, with a significantly reduced minimum detection limit. Additionally, the photon dose was suppressed. The thermal neutron enhancement method can increase the number of useful neutrons significantly but maintain the radiation dose. This greatly decreased the detection limit of the system for elemental quantification at an acceptable dose, which will broadly expand the application of the technology in research and clinical use. The method can also be applied to other neutron medical applications, including neutron imaging and radiotherapy.

Software-Based automation of burnup calculations for the NUR research reactor using SCALE/TRITON T6-DEPL sequence

This work represents a key milestone in the development of a Monte Carlo Burnup Calculation System (MCBCS) specially tailored for the NUR research reactor. Developed using Python and leveraging the 3D Monte Carlo TRITON depletion sequence (T6-DEPL) within the SCALE code, MCBCS accurately simulates the reactor’s operating history. The paper provides an overview of MCBCS, focusing on its components, verification, and validation.The verification and validation process cover both fresh and burnt core conditions. For the fresh core, comparisons of excess reactivity, control rods worth, and critical configurations against experimental data and MCNP5 calculations showed good agreement. Burnup calculations were validated against measured core excess reactivity, reactivity worths of fuel assemblies, and neutron flux distribution. The system slightly underpredicted core excess reactivity by −2.95%, and discrepancies in reactivity worths remained within the 7% uncertainty range. Neutron flux distribution showed good consistency with minor location-specific deviations.Overall, these findings confirm MCBCS as a reliable and accurate tool for burnup calculations of the NUR research reactor.

Solid, structured composite neutron detectors with high dynamic range capability

Neutron detectors are essential across disciplines such as fundamental science, nuclear security, safeguards, and civilian applications. While 3He-filled gas proportional counters have long been revered for their efficacy in detecting thermal neutrons and praised for their efficiency, neutron/gamma discrimination, and stability, the scarcity of 3He has spurred a search for alternatives. Here, we explore a solid structured scintillating particle composite (SPC) consisting of 6Li-containing scintillating glass particles within an acrylic matrix as a neutron detector for high dynamic range applications. We show for the first time that an SPC neutron detector can boast an intrinsic detection efficiency of 0.261% for pure 252Cf fission neutrons and an overall neutron detection efficiency of (0.546 ± 0.003)% at the Neutron Free-in-Air facility while being able to function in an intense gamma-ray environment. We also show that the SPC neutron detector supports fast neutron capture times and enables a dual-readout scheme that extends the detector dynamic range to high incident neutron fluxes. A scalable fabrication process allows for tailoring the SPC detector properties to the requirements of specific applications. Good agreement is found between the experimental results taken with a National Institute of Standards and Technology traceable 252Cf source and the coupled MCNP6 and optical-ray-tracing simulations.

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.

External Moderation of Reactor Core Neutrons for Optimized Production of Ultra-Cold Neutrons

The ultra-cold neutron (UCN) source being commissioned at North Carolina State University’s PULSTAR reactor is uniquely optimized for UCN production in the former graphite-filled thermal column outside of the reactor pool. The source utilizes a remote moderation design, which is particularly well suited to the PULSTAR reactor because of its high thermal and epithermal neutron leakage from the core face. This large non-equilibrium flux from the core is efficiently transported to the UCN source through the specially designed beam port in order to optimize UCN production at any given reactor power. The increased distance to the source from the core also greatly limits the heat load on the cryogenic system. A MCNP (Monte Carlo N-Particle) model of this system was developed and is in good agreement with gold foil activation measurements using a test configuration as well as with the real UCN source’s heavy water moderator. These results established a firm baseline for estimates of the cold neutron flux available for UCN production and prove that remote moderation in a thermal column port is a valuable option for future designs of cryogenic UCN sources.

Development of a silicon carbide radiation detection system and experimentation of the system performance

Silicon carbide (SiC) detectors have excellent radiation detection capabilities for various radiation particles, including high energy resolution, fast response times, and good radiation resistance. A SiC radiation detection system was developed to measure the neutron fluence rate and the γ-ray dose rate in high intensity radiation fields. The system was composed of two SiC detectors, a temperature monitor, two preamplifiers for each SiC detector, a data acquisition unit with two signal channels, three pairs of communication devices, and an application software to analyze and visualize the measurement data. The two SiC detectors were fabricated based on two kinds of 4H-SiC diodes and used to respectively respond to neutrons and γ-rays. Repeated experiments showed that the two SiC detectors of the system can respond to α-particles, neutrons, and γ-rays. To verify the performance of the SiC detection system, including the response linearity of the neutron fluence rate, the measurement range of the γ-ray dose rate, and the radiation resistance of the SiC radiation detectors, the system was tested in multiple neutron and γ-ray fields. The tests results show the system can measure the neutron fluence rate from 5.64 × 10 2 cm−2 s−1 to 1.03 × 10 5 cm−2 s−1 with excellent linearity response, and the γ-ray dose rate from 0.005 Gy/h to 20 Gy/h. Furthermore, the SiC detectors demonstrated good radiation resistance. The neutron and γ-ray radiation field can still be measured stably by the system after exposure to neutron fluence of 1.07 × 10 14 cm−2 and γ-ray dose of 3.52 × 10 4 Gy. This work is the preliminary research to continue the exploration how to measure the n/γ hybrid fields accurately using SiC detectors considering the different energy of neutrons.

Examination of the use of thorium-based fuel for burning minor actinides in European sodium cooled fast reactor

Abstract To reduce the potential and volume of radiotoxicity of radioactive wastes, minor actinides (MAs) elements recovered from discharged fuel are to be recycled. In this context, the main aim of this work is to examine the possibility of using thorium-based fuel for burning extra MAs in the fourth generation European Sodium cooled Fast Reactor (ESFR). For that MCNPX transport code was used to design a representative fuel assembly of this concept. The reference uranium-based fuel was uniformly replaced by thorium mixed with different fractions of MAs and burned for 2050 effective full power days. The influence of MAs on the evolution of reactivity and fuel transmutation is being investigated to ensure the performance of the fuel. Also, different neutronic and safety parameters such as neutron yield, neutron spectrum, neutron flux, effective delayed neutron fraction, Doppler constant, sodium void reactivity coefficient, were computed and their change with MAs content was investigated. Results indicate that recycling of MAs in ESFR using thorium-based fuel has a positive effect on reactivity and transmutation evolution but on other hand has a negative effect on safety parameters.

Development of a multi-layer high-efficiency GEM-based neutron detector for spallation sources

Neutron detection is nowadays mostly based on 3He gas detectors, but its shortage and the continuous upgrades of the neutron facilities require new devices to perform experiments with maximum performances. This work presents a new detector based on the Gas Electron Multiplier (GEM) combined with several boron layers. This detector combines the features of GEM technology with the properties of boron as a neutron converter and the device is produced to sustain high neutron fluxes with high detection efficiency. The detector has been characterised at the ISIS Pulsed Neutron and Muon Source (UK). Based on the analysis of our results, the detector has shown a good response to thermal and epithermal neutrons reaching a detection efficiency of 16% at 1.8 Å (25 meV). The good detection efficiency (even increasable with the addition of further boron GEM foils) and the good time resolution, make the detector a unique device for the neutron techniques. In particular, its use can easily be envisaged in techniques involving neutron transmission measurements, that require high fluxes impinging on the detectors, with the added bonus of a 2D-resolved capability due to the padded anode.