Digital Neutron Research Articles

Resolution enhancement of neutron radiography image using combined SRCNN-POCS method

In this work some post-acquisition image enhancement methods are applied to the case of under-sampled and low-resolution neutron radiography image. The digital neutron imaging system used in this study is a low resolved one; mainly based on an 8 bits CCD camera and a 250μm LiF-ZnS scintillator screen. In this paper, a software solution is proposed to improve the quality of digital neutron radiography images. The suggested method is a combination between the convolution neural network single-image and the projection onto convex sets multi-image super-resolution algorithms (SRCNN-POCS). The results obtained are compared to those by each algorithm separately. The proposed SRCNN-POCS method was found to be more reliable by producing higher quality image than the others tested methods in terms of modulation transfer function, peak-signal-to-noise ratio (PSNR), and structural similarity (SSIM). The visual analysis of the obtained image confirms too the obtained results.

Boron-Based Neutron Scintillator Screens for Neutron Imaging.

In digital neutron imaging, the neutron scintillator screen is a limiting factor of spatial resolution and neutron capture efficiency and must be improved to enhance the capabilities of digital neutron imaging systems. Commonly used neutron scintillators are based on 6LiF and gadolinium oxysulfide neutron converters. This work explores boron-based neutron scintillators because 10B has a neutron absorption cross-section four times greater than 6Li, less energetic daughter products than Gd and 6Li, and lower γ-ray sensitivity than Gd. These factors all suggest that, although borated neutron scintillators may not produce as much light as 6Li-based screens, they may offer improved neutron statistics and spatial resolution. This work conducts a parametric study to determine the effects of various boron neutron converters, scintillator and converter particle sizes, converter-to-scintillator mix ratio, substrate materials, and sensor construction on image quality. The best performing boron-based scintillator screens demonstrated an improvement in neutron detection efficiency when compared with a common 6LiF/ZnS scintillator, with a 125% increase in thermal neutron detection efficiency and 67% increase in epithermal neutron detection efficiency. The spatial resolution of high-resolution borated scintillators was measured, and the neutron tomography of a test object was successfully performed using some of the boron-based screens that exhibited the highest spatial resolution. For some applications, boron-based scintillators can be utilized to increase the performance of a digital neutron imaging system by reducing acquisition times and improving neutron statistics.

Development of neutron flux measurement module for research reactors based on FPGA. Part-1-Model development and generic verification

Many Instrumentation and Control (I&C) systems of Research reactors experience aging difficulties that necessitate optimizing operations and safety systems. Therefore, great attention paid to FPGA-based I&C use particularly in safety systems; this is due to the FPGAs' notable simplicity and advantages. This work aims to develop a prototype FPGA-Based Nuclear Instrumentation Module to use it as part of the nuclear instrumentation of Egypt Second Research Reactor (ETRR-2) based on Zynq-7000 FPGA. This phase focusses on using a model-based design approach for FPGA customization by the Xilinx system generator tool, with embedded MATLAB/Simulink to implement the neutron flux parameter algorithms. Thus, a model of Digital Neutron Flux Module (DNFM) developed to digitize the neutron flux as obtained from the processed signal of the Compensated Ionization Chamber (CIC). Substantial quantities of the neutron flux of the ETRR-2 operating power range calculated. The amounts include the neutron flux in the logarithmic range, the reactor power as a percentage of the full power in the linear range and the rate of change of the neutron flux. Generic verification got done using simulated current signals, followed by direct verification methods using real ETRR-2's CIC output current signal. The results of the FPGA-based DNFM model is consistent with the ETRR-2's neutron flux measurements.

Measuring Thickness-Dependent Relative Light Yield and Detection Efficiency of Scintillator Screens.

Digital camera-based neutron imaging systems consisting of a neutron scintillator screen optically coupled to a digital camera are the most common digital neutron imaging system used in the neutron imaging community and are available at any state-of-the-art imaging facility world-wide. Neutron scintillator screens are the integral component of these imaging system that directly interacts with the neutron beam and dictates the neutron capture efficiency and image quality limitations of the imaging system. This work describes a novel approach for testing neutron scintillators that provides a simple and efficient way to measure relative light yield and detection efficiency over a range of scintillator thicknesses using a single scintillator screen and only a few radiographs. Additionally, two methods for correlating the screen thickness to the measured data were implemented and compared. An example 6LiF:ZnS scintillator screen with nominal thicknesses ranging from 0–300 μm was used to demonstrate this approach. The multi-thickness screen and image and data processing methods are not exclusive to neutron scintillator screens but could be applied to X-ray imaging as well. This approach has the potential to benefit the entire radiographic imaging community by offering an efficient path forward for manufacturers to develop higher-performance scintillators and for imaging facilities and service providers to determine the optimal screen parameters for their particular beam and imaging system.

An effective gamma white spots removal method for CCD-based neutron images denoising

In CCD-based neutron radiography and tomography, resulting digital neutron images are usually degraded by gamma white spots caused by the interaction between high energy gamma radiation and the charge-coupled device (CCD) detector. The gamma white spots have a major impact on the accuracy of quantitative data analysis and are difficult to remove due to their high magnitudes and non-uniform distribution. In this paper, we put forward a novel iterative filtering method based on a “detection-corrosion” strategy for degraded neutron image denoising. A collaborative filtering method is adopted via integrating a spatially adaptive filter and a median filter for gamma white spots detection and removal. The experimental results show that the proposed method can achieve satisfactory denoising effect with preferable preservation of image details.

Neutron–gamma discrimination based on quantum clustering technique

Abstract In this study, the digital neutron–gamma discrimination (DNGD) has been undertaken based on the pulse-shape discrimination on the anode pulse of a 2” by 2” right cylinder BC501A liquid scintillator with a fast data acquisition card (14-bit resolution, 500 MS/s). Seven different features of the anode pulse have been extracted and the discrimination based on quantum clustering (DBQC) has been studied. The influence of three different parameters (i.e., η , σ and K) on the discrimination figure-of-merit (FoM) has been investigated. In addition, the FoM dependencies on different choices of DBQC features have been determined. A 100 mCi 241Am-Be neutron source has been used for the DNGD and the calibration has been made with a 1.1 μ Ci 22Na gamma-ray source. The results show that the FoM value is around 1 at 100 keVee bias energy, whilst this value can be improved up to 50% by choosing and setting appropriate features.

On the possibility to investigate irradiated fuel pins non-destructively by digital neutron radiography with a neutron-sensitive microchannel plate detector with Timepix readout

Post irradiation examination (PIE) of nuclear fuels provides valuable information on their overall performance and is important for the assessment of safety and integrity of nuclear fuels. PIE is also very useful for the optimization of the spent fuel processing and for the development of future advanced fuels. Neutron radiography (NR) provides more comprehensive information about the internal condition of irradiated nuclear fuels than any other non-destructive PIE examination, and NR informs subsequent characterization by destructive techniques. At the present time most NR of irradiated fuel is conducted through image transfer methods using radiographic films or imaging plates. Digital neutron radiography of irradiated fuel poses significant challenges due to the extreme levels of radiation generated by the irradiated fuel itself. In this paper, we present the results of neutron imaging of irradiated fuel by a digital detector installed in a radial neutron beam with direct line of sight to the reactor core. The results of these experiments demonstrate the possibility to perform nearly real-time initial screening of the internal structure of an experimental transmutation fuel pin at the Neutron Radiography reactor at Idaho National Laboratory. The detector used in these experiments, based on neutron sensitive microchannel plates and Timepix readout. It sustained operation at high radiation environment with gamma dose rates of ∼5.5 Sv/hr from the sample itself and ∼2.0 Sv/hr in the neutron beam The influence of various experimental setup characteristics (beam collimation, gamma and neutron background fluxes, image integration time) on the quality of resulting neutron radiographic images was studied for the optimization of the future imaging experiments. These measurements also demonstrate the possibility to reduce the fraction of unwanted gamma events by lead filter installed between the detector and the irradiated fuel. At the optimal configuration the fraction of neuron counts was measured to exceed 90% of all registered events in the detector. The results indicate that non-destructive post irradiation examination of nuclear fuels can be extended beyond conventional neutron radiography to more advanced imaging techniques, which include tomographic, dark field, energy resolved neuron imaging and many others, enabling more detailed quantitative characterization beyond preliminary screening.

A Novel Technique for Removal of High Density White Spot Noise from Digital Neutron Radiographic Images

This paper proposes a novel technique of a daptive s witching a lternative m edian (ASAM) filter for high-density white spot noise removal. The ASAM filter is composed of two blocks filtering, namely main and secondary block filtering, respectively. The proposed secondary block filtering is a new technique in high-density impulse noise removal and the main contribution of this research. The ASAM algorithm was tested on the standard 8-bit gray-scale, 512×512 pixel Lena image and a real neutron radiographic image. The results showed significant reduction of white spot noise in both types of images through visual inspection. To measure the performance of noise removal in simulation test we measured the p eak si gnal - to -n oise r atio (PSNR), s tructural s imilarity (SSIM) index , and denoising time, and in real application tests we measured s ignal - to -n oise r atio (SNR). From the experiments of simulation test, at the highest level noise of 95 % th e obtained PSNR and SSIM are 23.584 dB and 0.696 respectively. These are higher than the results of other algorithms that are 16.697 dB and 0.475 , respectively , for DBA, 16.696 dB and 0.408 for NAFSM, and 18.860 dB and 0.568 for NASNLM. The denoising time s for DBA, NAFSM, NASNLM , and ASAM we re obtained as 6.469 s, 5.186 s, 36.735 s, and 5.197 s respectively. From the experiments of real application test we obtained the SNR for DBA, NAFSM, NASNLM , and ASAM as 32.42 dB, 6.01 dB, 18.77 dB , and 32.96 dB , respectively. In general, these results show that ASAM filter is superior to the existing filtering methods. The ASAM filter improved the image restoration quality, especially in removing the high-density white spot noise, and was able to yield good filtering result w hic h exhibits better PSNR, SSIM, denoising time , and qualitative visual inspection .

Optimization of the n/[formula omitted] separation algorithm for a digital neutron spectrometer

An analysis of various n/γ separation algorithms for a digital neutron spectrometer based on a stilbene crystal is performed. The same set of digital signals is used to test the following n/γ separation algorithms: the algorithm based on the correlation analysis; the algorithms based on the direct comparison of the pulse fast component and the pulse tail; SDCC (Simplified Digital Charge Collection) algorithm; PGA (Pulse Gradient Analysis) and FGA (Frequency Gradient Analysis) algorithms. It is shown that the algorithm based on the correlation analysis makes it possible to carry out the most effective n/γ separation for the entire pulse areas range.The shape of the distributions of the separation parameter for electrons and protons is analyzed. The correctness of these distributions approximation by the Gaussian curve is justified.An algorithm for the selection of the recoil protons resulting from the neutron interaction with the scintillator is proposed. It is based on the dynamic change of the separation parameter threshold depending on the pulse area. The distributions of the separation parameter for electrons and recoil protons strongly overlap for small pulse areas. In this case, the threshold for the separation parameter is chosen on the basis of the fixed probability of the event false identification. In our work, for example, the contribution of false events was fixed at 1%. With the increase of the pulse area, the distributions for electrons and protons begin to separate well. In this case it is convenient to set the threshold on the separation parameter at the point of their intersection. The use of this algorithm makes it possible to control the fraction of the falsely identified events for the entire range of the pulse areas. This is especially important for low-energy recoil protons, for which the probability of false event identification reaches its maximum. This algorithm was applied to the measurements performed by the time-of-flight method with a 252Cf source. The suppression degree of the prompt gamma-ray peak was 1100 at the neutron detection threshold of 275 keV. The contribution of the falsely identified events for the different parts of the time-of-flight spectrum using the proposed event selection algorithm was studied. The largest contribution to the false pulses was found for the neutron distribution edges — for the neutrons with the energies above 8 MeV and below 0.5 MeV, and it did not exceed 0.23% in the entire range of the pulse areas.

Digital neutron and gamma-ray radiography in high radiation environments with an MCP/Timepix detector

The radiation levels from used nuclear fuel make their post irradiation examination very difficult. Handling material removed from assemblies is almost exclusively performed inside shielded hot cells. Neutron radiography provides a non-destructive option for examining the internal structure, location and integrity of irradiated fuel. This paper addresses the performance and stability of a neutron imaging detector comprised of microchannel plates (MCPs) and a Timepix readout in gamma and neutron radiation environments. The detector is being considered as an alternative to the currently used activation foils and photographic film measurement techniques. The MCP/Timepix detector was used in neutron radiography experiment on a beam line at the Neutron Radiography Reactor (NRAD) at Idaho National Laboratory (INL). The detector was installed ∼17.5 m from the reactor core with a line of sight view to the core, which results in a fast and epithermal neutron spectrum as well as high intensity gamma radiation. The gamma-ray dose rate in the beam line was 812 mSv/hr. Complementary measurements were made using 137Cs and 60Co isotopic sources at gamma dose rates in excess of 6 Sv/hr. The detector functioned normally over several hours in this environment. These conditions are comparable to planned radiography of irradiated nuclear fuel assemblies. Experimental results using ASTM sensitivity and beam purity indicators suggest that a spatial resolution of <100 μm may be possible in neutron radiographic measurements using this detector for studies of irradiated nuclear fuel. Potential radiographic and tomographic applications are discussed. The results of present experiments are not limited to the studies of irradiated fuel assemblies, but rather are relevant to any experiment where imaging needs to be performed in a high radiation environment.

A novel digital neutron flux monitor for international thermonuclear experimental reactor

A novel full-digital real-time neutron flux monitor (NFM) has been developed for the International Thermonuclear Experimental Reactor. A measurement range of 109 counts per second is achieved with 3 different sensitive fission chambers. The Counting mode and Campbelling mode have been combined as a means to achieve higher measurement range. The system is based on high speed as well as parallel and pipeline processing of the field programmable gate array and has the ability to upload raw-data of analog-to-digital converter in real-time through the PXIe platform. With the advantages of the measurement range, real time performance and the ability of raw-data uploading, the digital NFM has been tested in HL-2A experiments and reflected good experimental performance.

Novel Neutron Detector Material: Microcolumnar Li&lt;italic&gt;x&lt;/italic&gt;Na1–&lt;italic&gt;x&lt;/italic&gt;I:Eu

Current neutron detection relies on 3He detectors, which are prone to material shortages, and suffer from poor spatial resolution. Such inadequacies make 3He unsuitable for new applications in state-of-the-art spallation sources such as the Spallation Neutron Source at Oak Ridge National Laboratory. Here, we report on the synthesis and application of a novel scintillator material for neutron detection: a mixed halide compound, Eu-doped (0.05 mol. %) 6Li x Na1– x (6LNI:Eu). 6LNI:Eu is a bright and efficient scintillator that can be grown in large format, thin films that exhibit microcolumnar structure in their cross section. The ability to grow in a microcolumnar structure means that a resolution of approximately 1 lp/mm can be achieved with neutrons using an electron-multiplying charge coupled device camera, thereby conferring an advantage over other solid (not gaseous) neutron scintillators, such as Cs26LiYCl6:Ce (CLYC) and GS20. The large area format of 6LNI:Eu is a significant advantage for detecting specular reflections in neutron scattering experiments. Moreover, the microcolumnar nature of films permits high-resolution imaging when films are coupled to a-Si:H flat panels, CCD, or CMOS readouts. While not the primary focus, this also makes them suitable for digital neutron radiography. Enriched 6Li was used to increase the neutron absorption cross section of the thin films, while preserving the brightness of the scintillation response. Here, we describe the vapor deposition techniques used to fabricate the 6LNI:Eu scintillator screens and report on the test results. The films were tested at the High Flux Isotope Reactor (HFIR) at ORNL. An efficiency of approximately 43% for 4.1-A neutrons with a resolution of 650 $\mu \text{m}$ was measured for a 375- $\mu \text{m}$ thick film of 6LNI:Eu using a neutron sensitive Anger camera.

A digital wide range neutron flux measuring system for HL-2A**Contributed paper, published as part of the Proceedings of the 3rd Domestic Electromagnetic Plasma Diagnostics Workshop, September 2016, Hefei, China.

To achieve wide-range, high-integration, and real-time performance on the neutron flux measurement on the HL-2A tokamak, a digital neutron flux measuring (DNFM) system based on the peripheral component interconnection (PCI) eXtension for Instrumentation express (PXIe) bus was designed. This system comprises a charge-sensitive preamplifier and a field programmable gate array (FPGA)-based main electronics plug-in. The DNFM totally covers source-range and intermediate-range neutron flux measurements, and increases system integration by a large margin through joining the pulse-counting mode and Campbell mode. Meanwhile, the neutron flux estimation method based on pulse piling proportions is able to choose and switch measuring modes in accordance with current flux, and this ensures the accuracy of measurements when the neutron flux changes suddenly. It has been demonstrated by simulated signals that the DNFM enhances the full-scale measuring range up to 1.9 × 108 cm−2 s−1, with relative error below 6.1%. The DNFM has been verified to provide a high temporal sensitivity at 10 ms time intervals on a single fission chamber on HL-2A.

Digital spectrometer for prompt fission neutron spectrum measurements

This paper presents a digital neutron spectrometer based on simultaneous digitizing of the signals from the fission chamber and a scintillation detector. The scintillation detector is based on stilbene crystal. The intrinsic detection efficiency of the used stilbene crystal and the energy dependence of the light output for the recoil protons were measured. It is shown, that the method allowed us to achieve time resolution of 1.5 ns and an good n/γ separation down to neutron energies of 400 keV.

Digital neutron–gamma discrimination with scintillators: An innovative approach

In this paper, a digital neutron–gamma discrimination (DNGD) method with an NE213 scintillator has been proposed in which the anode pulse is divided into two different pulses, one representing the amplitude and the other characterizes the DNGD. Then the two pulses are summed up after travelling through delay and mixer circuits to form an input pulse for sampling in analog–to–digital converter (ADC). The discrimination tests have been performed with an 8–bit digital storage oscilloscope (DSO) as ADC and 241Am–Be neutron source, whereas the Fourier method has been used to derive the discrimination characteristic. The results confirm the fast performance and efficiency of proposed method.

Microstructured boron foil scintillating G-GEM detector for neutron imaging

In this study, a new simple neutron imaging gaseous detector was successfully developed by combining a micro-structured 10B foil, a glass gas electron multiplier (G-GEM), and a mirror–lens–charge-coupled device (CCD)–camera system. The neutron imaging system consists of a chamber filled with Ar/CF4 scintillating gas mixture. Inside this system, the G-GEM is mounted for gas multiplication. The neutron detection in this system is based on the reaction between 10B and neutrons. A micro-structured 10B is developed to overcome the issue of low detection efficiency. Secondary electrons excite Ar/CF4 gas molecules, and high-yield visible photons are emitted from those excited gas molecules during the gas electron multiplication process in the G-GEM holes. These photons are easily detected by a mirror–lens–CCD–camera system. A neutron radiograph is then simply formed. We obtain the neutron images of different materials with a compact accelerator-driven neutron source. We confirm that the new scintillating G-GEM-based neutron imager works properly with low gamma ray sensitivity and exhibits a good performance as a new simple digital neutron imaging device.

Digital neutron image enhancement based on total variation-based ℓ0 minimization

For the digital neutron radiography applications, low contrast is inevitable due to the neutron and photon statistics limited or high transmission (and low attenuation) in the sample. In this paper, we introduce a total variation-based ℓ0 minimization model and solve it by an alternating minimization approach. The proposed model is applied to base & detail decomposition for the dynamic range modification and detail magnification. Experimental results on several real neutron images are presented to demonstrate the effectiveness of the proposed method.

Precision of Porosity Calculation from “Material Stopping Power” Using Neutron Radiography

The ability of digital neutron radiography to determine several important physical properties (e.g. porosity) of porous media and to identify the location of constituencies inside the sample in a non-destructive manner, are of great important in the fields of nuclear waste encapsulation and –shielding, civil engineering and geological studies. Unfortunately material thickness and radiation scatter hamper the accuracy and precision ofmeasurements. Correction methods are used to minimize the effect of neutron scattering in quantitative analysis neutron radiography experiment. To further minimize the error due to neutron scattering, it is advisable to position the sample as far away as possible from the neutron detector when perform quantitative neutron radiography. That requires the trade between sharpness and neutron scattering. In this work the neutron radiographyquantitative experiments were performed when the samples are placed too close (≈ 0.1cm) to the detector; and accurate results of porosity measurement were obtained without applying any neutron scattering correction tool or method.

Test of digital neutron–gamma discrimination with four different photomultiplier tubes for the NEutron Detector Array (NEDA)

A comparative study of the neutron–γ discrimination performance of a liquid scintillator detector BC501A coupled to four different 5in. photomultiplier tubes (ET9390kb, R11833-100, XP4512 and R4144) was carried out. Both the Charge Comparison method and the Integrated Rise-Time method were implemented digitally to discriminate between neutrons and γ rays emitted by a 252Cf source. In both methods, the neutron–γ discrimination capabilities of the four photomultiplier tubes were quantitatively compared by evaluating their figure-of-merit values at different energy regions between 50keVee and 1000keVee. Additionally, the results were further verified qualitatively using time-of-flight to distinguish γ rays and neutrons. The results consistently show that photomultiplier tubes R11833-100 and ET9390kb generally perform best regarding neutron–γ discrimination with only slight differences in figure-of-merit values. This superiority can be explained by their relatively higher photoelectron yield, which indicates that a scintillator detector coupled to a photomultiplier tube with higher photoelectron yield tends to result in better neutron–γ discrimination performance. The results of this work will provide reference for the choice of photomultiplier tubes for future neutron detector arrays like NEDA.

Drying Kinetics of Particulate Corundum Layers

The results of studies on drying of layers of particulate corundum performed with digital neutron radiography are presented. The statistical analysis of images was employed to reveal main features of the drying process. It is shown that the rst period of drying proceeds within the whole body of the sample. Marked signatures of the end of rst period of drying were found for sample mass, temperature, average image brightness as well as standard deviation of brightness time evolutions for the system composed of ne grains. The chaotic nature of this period is delineated by a distinct maximum in the standard deviation of sample image brightness. The presence of the drying front moving from the open surface to the sample bottom is proven to be an attribute of the advanced drying period.