Prompt Gamma Neutron Activation Research Articles

Process analytical technologies for precise, timely and representative elemental and moisture measurement for conveyed flows

Industrial and technological processes are very difficult to manage when the quality of feed and product or discard are not measured with confidence. Effective control can occur when process analytical technologies are chosen that provide representative, precise, and timely measurements. For the measurement technique to be representative it must comply with the Theory of Sampling (TOS) and provide an equal chance of any component in the streaming material to be included in the support for the measurement. This generally precludes technologies that measure only the surface of materials, or biased measurements stemming from a limited portion of the material only, particularly in the minerals processing and recycling sectors, which usually display high compositional variability. The location of the analytical technology should relate to the benefit being targeted and allow for enough reaction time to respond to the quality in some way; diverting short increments based on composition and decision parameters based on process impact, blending with other quality materials, or feeding information backwards or forwards. Feed forward options can include flow rate control, reagent control, operational process variables that impact recoveries, etc. Major benefits have been achieved in measuring coarse conveyed flows with high specification Prompt Gamma Neutron Activation Analysis (PGNAA) over short increments (thirty seconds to two minutes) for most elements, or over five to ten minute increments for trace elements, such as gold. PGNAA applied to conveyed flows allow the full flow to be measured continuously and composition averaged for each increment in real time. The use of penetrative and continuous moisture measurement using transmission microwaves has also proved effective for moisture monitoring and management. Precisions between laboratory samples of the flow and analysis data from analysers can be sufficient to have high confidence in resulting process control decisions. This paper explains the benefits in more detail and includes case studies to highlight actual benefits derived from the application of such systems. It should be noted that sampling of the materials is still required for calibration and adjustments for the process analytical tools.

Thermal neutron beam optimization for PGNAA applications using Q-learning algorithm and neural network

As a powerful, non-destructive analysis tool based on thermal neutron capture reaction, prompt gamma neutron activation analysis (PGNAA) indeed requires the appropriate neutron source. Neutrons produced by electron Linac-based neutron sources should be thermalized to be appropriate for PGNAA. As a result, thermalization devices (TDs) are used for the usual fast neutron beam to simultaneously maximize the thermal neutron flux and minimize the non- thermal neutron flux at the beam port of TD. To achieve the desired thermal neutron flux, the optimized geometry of TD including the proper materials for moderators and collimator, as well as the optimized dimensions are required. In this context, TD optimization using only Monte Carlo approaches such as MCNP is a multi-parameter problem and time-consuming task. In this work, multilayer perceptron (MLP) neural network has been applied in combination with Q-learning algorithm to optimize the geometry of TD containing collimator and two moderators. Using MLP, both thickness and diameter of the collimator at the beam port of TD have first been optimized for different input electron energies of Linac as well as for moderators’ thickness values and the collimator. Then, the MLP has been learned by the thermal and non-thermal neutron flux simultaneously at the beam port of TD calculated by MCNPX2.6 code. After selecting the optimized geometry of the collimator, a combination of Q-learning algorithm and MLP artificial neural network have been used to find the optimal moderators’ thickness for different input electron energies of Linac. Results verify that the final optimum setup can be obtained based on the prepared dataset in a considerably smaller number of simulations compared to conventional calculation methods as implemented in MCNP.

Biomedical NiTi and β-Ti Alloys: From Composition, Microstructure and Thermo-Mechanics to Application

A comprehensive, bottoms-up characterization of two of the most widely used biomedical Ti-containing alloys, NiTi and β-Ti, was carried out applying a novel combination of neutron diffraction, neutron prompt-gamma activation, surface morphology, thermal analysis and mechanical tests, to relate composition, microstructure and physical-chemical-mechanical properties to unknown processing history. The commercial specimens studied are rectangular (0.43 × 0.64 mm~0.017 × 0.025 inch) wires, in both pre-formed U-shape and straight extended form. Practical performance was quantitatively linked to the influence of alloying elements, microstructure and thermo-mechanical processing. Results demonstrated that the microstructure and phase composition of β-Ti strongly depended on the composition, phase-stabilizing elements in particular, in that the 10.2 wt.% Mo content in Azdent resulted in 41.2% α phase, while Ormco with 11.6 wt.% Mo contained only β phase. Although the existence of α phase is probable in the meta-stable alloy, the α phase has never been quantified before. Further, the phase transformation behavior of NiTi directly arose from the microstructure, whilst being highly influenced by thermo-mechanical history. A strong correlation (r = 0.878) was established between phase transformation temperature and the force levels observed in bending test at body temperature, reconfirming that structure determines performance, while also being highly influenced by thermo-mechanical history. The novel methodology described is evidenced as generating a predictive profile of the eventual biomechanical properties and practical performance of the commercial materials. Overall, the work encompasses a reproducible and comprehensive approach expected to aid in future optimization and rational design of devices of metallic origin.

Partial Cross-Section Calculations for PGNAA Based on a Deterministic Neutron Transport Solver

The measurement facility QUANTOM is being developed for the material analysis of radioactive waste packed up in 200-L drums. QUANTOM enables a spatially resolved elemental analysis based on prompt gamma neutron activation analysis. The evaluation of the spatially resolved gamma spectra relies on the calculation of partial cross sections. Hereby, the neutron flux spectrum enters as a parameter, which needs to be simulated in the full three-dimensional geometry of the measurement facility. To ensure that the simulations can be carried out within an acceptable time frame, we use a deterministic neutron transport code specially developed for this purpose based on the SPN approximation of the linear Boltzmann equation. The following question arises: Does the approximation in the neutron transport model still allow a calculation of the partial cross sections at a sufficient level of accuracy. Therefore, in this paper, we study the calculation of partial cross sections in light of the approximation in the neutron transport model in the geometrical setting of the measurement facility. In a simulation study we consider four typical matrix materials and compare cross sections for all elements of the periodic table to reference results obtained by Monte Carlo simulations.

Digital Stabilization Algorithm for the Gamma Spectra of Scintillator Detectors in PGNAA

The drift of the acquired gamma spectra used in prompt gamma neutron activation analysis (PGNAA) has a large influence on the analysis accuracy. To address this problem, a set of digital algorithms for wide-region spectral stabilization was investigated. The spectra were first calibrated by using the nonlinear method, and the counts in each channel were corrected with cubic spline interpolation. Then, the corrected spectra were used for stabilization according to the established stabilization method. The experimental results showed a significant reduction in the maximum relative deviation of the peak position for both sodium iodide [NaI(Tl)] (7.26%–1.25%) and bismuth germanate oxide (BGO) (20.15%–2.91%) detectors.

Conceptual study of salt-meter with 252Cf neutron source for on-site inspection of bridge structure

We studied the feasibility of a portable salt-meter incorporating a prompt gamma neutron activation analysis with a californium-252 neutron source to meet urgent demands for non-destructive methods without pre-processing in large bulk materials such as concrete structures. This technique is aiming at the chloride ion concentration in concrete structures with a depth profile from the surface to steel bar. From the portability point of view, a californium-252 neutron source is adopted and we have performed a preliminary experiment and simulations with a simple geometry. In this paper, we describe the performances of salt detection sensitivities by applying the prompt gamma neutron activation analysis.

Neutronic design of pulsed neutron facility (PNF) for PGNAA studies of biological samples

This paper introduces a novel concept of the pulsed neutron facility (PNF) for maximizing the production of the thermal neutrons and its application to medical use based on prompt gamma neutron activation analysis (PGNAA) using Monte Carlo simulations.The PNF consists of a compact D-T neutron generator, a graphite pile, and a detection system using Cadmium telluride (CdTe) detector arrays. The configuration of fuel pins in the graphite monolith and the design and materials for the moderating layer were studied to optimize the thermal neutron yields. Biological samples – normal and cancerous breast tissues – including chlorine, a trace element, were used to investigate the sensitivity of the characteristic γ-rays by neutron-trace material interactions and the detector responses of multiple particles.Around 90 % of neutrons emitted from a deuterium-tritium (D-T) neutron generator thermalized as they passed through the graphite stockpile. The thermal neutrons captured the chlorines in the samples, then the characteristic γ-rays with specific energy levels of 6.12, 7.80 and 8.58 MeV were emitted. Since the concentration of chlorine in the cancerous tissue is twice that in the normal tissue, the count ratio of the characteristic γ-rays of the cancerous tissue over the normal tissue is approximately 2.

Cyclotron-produced neutrons measurements using chlorine activation

Certain proton reactions emit fast neutrons, and this study examines the feasibility of measuring the neutrons emitted from a typical medical 18 MeV proton cyclotron using prompt gamma neutron activation on chlorine. This neutron detection method, employed by previous studies, has been deemed successful and highly promising. The measurement system includes a gamma radiation scintillator, for example NaI(Tl), with a spectrometry setup facing a convertor that contains a chlorine solution or compounds thereof.In its natural state, elemental chlorine is a chemically unstable gas, making its use as a convertor challenging. Therefore, chlorine is usually dissolved in solutions. Such solutions are diluted with water and contain hydrogen atoms that interact intensively with the neutrons and may interrupt their detection.Accordingly, in this study, a 25.5 L container of KCl powder was placed in front of a NaI(Tl) (7.62 cm diameter by 7.62 cm length) spectrometer near a cyclotron facility. The gamma radiation spectroscopy allowed the identification of the chlorine INS gamma line at 788 keV, alongside the K-40 gamma peak. In addition, it was possible to detect the fast neutrons using a total-counting mode, when the INS counts appear above the K-40 counting level.The chlorine INS peak in the KCl convertor was detected simultaneously during the cyclotron operations and provided a direct indication of neutron exposure around the cyclotron. We conclude that using KCl as a converter, with a NaI(Tl) gamma rays counter, is successful in detecting cyclotron-produced neutron emissions.

Characterization of (252Cf-ZrH2) Monte Carlo model for detection of nitrogen and chlorine by thermal neutron-capture PGNAA

Monte Carlo calculations were conducted to explore the potentials of zirconium hydride (ZrH2) as possible alternative moderator for high density polyethylene (HDPE) and water (H2O) in a simple (100MBq) 252Cf-based Prompt Gamma Neutron Activation Analysis (PGNAA) assembly. The thickness of moderators was optimized against thermal neutron flux (ϕth) as well as the ratio of (thermal/fast) neutron flux (ϕth/ϕfast), calculated at the sample cavity. The capabilities of moderating 252Cf source neutrons and minimizing the γ-background due to 2525Cf fission γ-rays, 1H(nth,γ)2H γ-rays originates in the moderator as well as other (nth,γ) capture reactions from surroundings were investigated. The optimal selected moderator with sufficient (ϕth) at the sample cavity and minimum γ-contamination was selected as ZrH2 for use in a252Cf-PGNAA for detection of nitrogen and chlorine in bulk sample simulants of melamine (C3H6N6) and sodium chloride (NaCl). The prompt γ-rays spectra were analyzed to detect the 10.83 MeV and 6.11 MeV prompt γ-peaks released by 14N(nth,γ)15N and 35Cl(nth,γ)36Cl, respectively. The results showed that the (252Cf-ZrH2)-PGNAA model has positively identified nitrogen with net counts (1.25 × 105 n.cm−2.s−1) and Chlorine with net counts of (4.19 × 106 n.cm−2.s−1), achieving minimum detectable concentrations of about (350 ppm at 10.83 MeV) and (640 ppm at 6.11 MeV), respectively.

Experimental determination and simulation of neutron and gamma flux parameters in horizontal channel for Prompt Gamma Neutron activation analysis implementation at Es-Salam research reactor

Since 2004, the k0-standardization method has been applied and evaluated in three experimental channels at Es-Salam research reactor of the Algerian Centre de Recherche Nucléaire de Birine (CRNB). The NAA laboratory at CRNB is well involved in the implementation of the Prompt Gamma Neutron Activation Analysis to extend the field of analyses to some additional elements to the NAA method. In the frame work of the PGNAA research project, several works have been carried out in the present study including the experimental and simulation characterization of neutron and gamma flux parameters of the horizontal channel. The absolute and relative neutron fluxes of the studied channel have been determined by using a set of monitors including Al, Au, Cu, Fe and Zr. Furthermore, measurements of gamma and thermal neutron doses have been achieved using LiF: Mg, Ti (TLD 600 and TLD 700) detectors. The gamma dose rate at the experimental channel center position was found to be 0.6 Gy/min·MW. The experimental results obtained in this study have been used as input information for simulations and design of neutron facilities and radiation shielding, which are successfully established by SCALE code. The results were compared in terms of spectral detail and precision and a good agreement has been revealed. The value of the thermal neutron flux was found to be (6.29 ± 0.25) E+08 n/cm2.s which allows the implementation of PGNAA at the Es-Salam research reactor.

Neutron beam characterization for the Moroccan TRIGA Mark II reactor

The TRIGA Mark II research reactor is equipped with four horizontal beam tubes and one thermal column. The tangential (NB1) and Percy (NB2) beam tube, which are the subject of this study, are dedicated to Prompt Gamma Neutron Activation Analysis and Neutron Imaging, and (Neutron Diffraction) facilities, respectively. A preliminary neutron characterization is needed for each beam tube. Basic beam parameters, including neutron flux, were measured with threshold activation detectors for the two beam tubes (NB1, NB2). Activity measurement through gamma spectrometry was performed with a high-purity germanium detector to evaluate the neutron flux through multiple analyses of activation detectors. The results were compared with those obtained using an approximated Monte Carlo simulation model for the reactor. A comparison between the experimental and simulated results revealed a good level of agreement.

Characterization of Neutron and Gamma Beams at the Tangential Beam Port of TRIGA 2000 Reactor Using Monte Carlo Methods

Characterization neutron and gamma beams at the tangential beam port of TRIGA 2000 reactor was done in order to prepare the development of Prompt Gamma Neutron Activation Analysis (PGNAA) facility. Characterization was performed by simulation using Monte Carlo method with MCNPX and PHITS computer codes. The MCNPX code was used to simulate the neutron-gamma fluence and energy spectra in the reactor core from fission reactions. The PHITS code was used to simulate the distribution of the neutron-gamma fluence and energy spectra in the tangential beam port. The simulation was done by modeling the geometry and elemental composition of reactor material and radiation source model in the form of fission reaction in the reactor core. The results indicate that the tangential beam port provides better neutron spectrum and neutron-gamma fluence ratio for PGNAA. These results will be used to simulate the colimator design and neutron interaction with samples in PGNAA facility.

Modeling and simulating the induced effect on the steel collimator plug used in the PGAA facility of the Moroccan TRIGA Mark-II reactor under different neutron irradiation levels

The main objective of this study is to assess the behavior of the steel collimator plug (steel plug) dedicated to the Prompt Gamma Neutron Activation Analysis (PGAA) facility in the neutron beam tube (NB1) of the 2 MW Moroccan TRIGA Mark-II research reactor. The main function of this steel plug is to reduce the neutron and gamma beam cross section from 15 cm to 5 cm in diameter. This steel plug plays a crucial role in reactor safety because it replaces the original neutron beam plug while also stopping the entire incoming neutron beam. Three aspects were therefore involved in this study, including: i) the released heat induced by both neutron and gamma radiation, ii) the swelling effect caused by both radiation and temperature increases in the steel and iii) the radioactivity induced by neutron radiation in the steel plug. An MCNP6.2 model for the TRIGA Mark-II reactor was used to calculate the neutron spectrum inside the NB1 beam tube at the inlet part of the steel plug. A gamma spectrum of a 900 MW PWR reactor was used as input to account for the gamma radiation's effects on the steel plug. In order to select a convenient steel, two investigations were carried out for two grades of steel, namely mild steel (E235) and 304L stainless steel (SS304L). The results were calculated using an in-house FORTRAN-based program and validated using COMSOL, SPECTER codes, and “Neutron Activation and Scattering Calculator” tools.The results revealed that both the E235 and 304L steels are convenient from a safety point of view, although the E235 steel is preferred at the decommissioning stage.

Optimization of PGNAA device and algorithm for testing basicity index of sinter mixture

Sinter is the main raw material in the blast furnace iron-making process, and basicity (CaO/SiO2) is an important quality index of sinter. Prompt gamma neutron activation analysis is a multi-elemental online detection technology that has been successfully applied in cement, coal, etc. Compared with cement as a raw material, sinter exhibits poor moderation ability and a large neutron absorption cross section. Therefore, cement detection devices are not suitable for sinter mixture detection. In this study, a prompt gamma neutron activation analysis equipment used for testing cement was re-optimized to render it suitable for measuring a sinter mixture. Using Monte Carlo simulation, the comprehensive detection efficiency of the detection device improved by 71.52%. Because of the gamma-ray self-shielding effect of the sinter mixture, the detection errors of CaO and SiO2 are significant. By applying the gamma-ray self-shielding correction algorithm, the detection accuracies improved, and their linear correlation coefficients R2 exceeded 0.99. Furthermore, by applying an improved analyzer to a sintering plant, the first-grade product rate of the factory increased by 4.64%.

Determination of the neutron yield of Be, V and Ta targets irradiated with protons (22-42 MeV) by means of prompt gamma neutron activation analysis

The neutron yield for beryllium, vanadium and tantalum irradiated with 22, 27, 33 and 42MeV protons is indirectly determined by Prompt Gamma Neutron Activation Analysis (PGNAA). The neutron-to-gamma conversion rate is measured with an AmBe calibration neutron source. Corrections by escaped neutrons are applied via MCNP simulations of the experiment using the ENDF/B-VII.1 database. The experimental results are in good agreement with the neutron yield obtained from simulations deviating by 0.4% to 13%.

Dose rate evaluation in a laboratory for prompt gamma neutron activation analysis by Monte Carlo simulation

As an excellent tool, neutron generators have been used widely in prompt gamma-ray neutron activation analysis (PGNAA) technology. In this work, a new PGNAA laboratory was proposed to equip a D–T neutron generator. For safety reasons, radiation shielding of the laboratory was designed, and its protection efficiency was validated by using MCNP simulations and experiments. The results show that the proposed shielding can meet the radiation protection set by the International Commission on Radiological Protection and the Chinese government.

Development of a Compact Deuterium–Deuterium Neutron Generator for Prompt Gamma Neutron Activation Analysis

A compact deuterium–deuterium neutron generator for prompt gamma neutron activation analysis was developed at Institute of Nuclear Physics and Chemistry. A neutron yield of 3.6 × 108 s–1 was achieved during of the bombardment of a titanium drive-in target by a 6.8 mA deuteron beam at 115 keV. The deuteron beam was generated by a permanent magnet microwave ion source. An 85 h long run with 1.2 × 108 s–1 average neutron yield was performed at 85 keV and 3.7 mA. The operating mode of the neutron generator reached 99.95%.

Detecting shielded explosives by coupling prompt gamma neutron activation analysis and deep neural networks

Prompt Gamma Neutron Activation Analysis is a nuclear-based technique that can be used in explosives detection. It relies on bombarding unknown samples with neutrons emitted from a neutron source. These neutrons interact with the sample nuclei emitting the gamma spectrum with peaks at specific energies, which are considered a fingerprint for the sample composition. Analyzing these peaks heights will give information about the unknown sample material composition. Shielding the sample from gamma rays or neutrons will affect the gamma spectrum obtained to be analyzed, providing a false indication about the sample constituents, especially when the shield is unknown. Here we show how using deep neural networks can solve the shielding drawback associated with the prompt gamma neutron activation analysis technique in explosives detection. We found that the introduced end-to-end framework was capable of differentiating between explosive and non-explosive hydrocarbons with accuracy of 95% for the previously included explosives in the model development data set. It was also, capable of generalizing with accuracy 80% over the explosives which were not included in the model development data set. Our results show that coupling prompt gamma neutron activation analysis with deep neural networks has a good potential for high accuracy explosives detection regardless of the shield presence.

Quantification of boron contents in BN/BCN composites by prompt gamma-ray neutron activation analysis utilizing thermal neutron beam at Dhruva reactor

Samples of boron (B) doped g-C3N4 (BCN) and boron nitride (BN) composite (BN/BCN) were synthesized by a simple in situ technique by varying BN content in the range of 1–10% mass fraction and boron contents are crucial for improving the activity of the composites. The composites were characterized by PXRD and FTIR in order to confirm their successful formation. Prompt Gamma Neutron Activation Analysis (PGNAA) using thermal neutron beam from Dhruva reactor was used for non-destructive determination of total boron concentration by measuring 478 keV prompt gamma-ray from 10B (n, αγ)7*Li reaction. The obtained boron concentrations are in the range of 2.3–6.2% mass fraction.

GEOSCAN-M Use For Process Improvement in Phosphate Plants

Real-time on belt analysis of minerals is becoming increasingly important in mining operations to ensure optimal economic outcomes. Prompt gamma neutron activation analysis has long been utilised for real-time elemental analysis in bulk materials, with Scantech’s GEOSCAN being the instrument of choice in the minerals sector. The GEOSCAN has been successfully adapted to a wide range of commodities, with its use in the phosphate sector one of the fastest growing areas for use of the technology. Using data for both feedback and feedforward control has seen improvements in efficiency at the mine, at the beneficiation plant, and in the production of phosphoric acid and fertiliser products. This paper introduces the GEOSCAN for the phosphate industry, and provides a case study for its use at an example Middle Eastern plant. A series of results are provided for various applications, demonstrating the instrument’s performance, and some of the key benefits of the equipment are discussed.