Am-Be Neutron Source Research Articles

The design of novel neutron shielding (Gd+B4C)/6061Al composites and its properties after hot rolling

In this paper, a novel approach to evaluate the neutron shielding performance of multiphase neutron absorber reinforced aluminum matrix composites was proposed through the establishment of a direct relationship between equivalent B areal density (EBAD) of the composite and its thermal neutron shielding coefficient. It was found when the EBAD of the composite was 0.1105 g/cm2, the thermal neutron shielding coefficient will achieve 99%. Based on this proposed approach, (1% Gd+15% B4C)/6061Al composite plates were successfully fabricated using vacuum hot pressing method followed hot rolling. The results showed that when the EBAD of the composite was 0.1871 g/cm2, the thermal neutron shielding coefficient was above 99.9%, which was consistent with the theoretical calculation. The 5 mm-thick composite plate also had a good shielding effect on the Am-Be neutron source. Meanwhile the composite possessed a 380 MPa tensile strength as well as an elongation of 5% after hot rolling. In all, the (Gd + B4C)/6061Al composite designed in this paper showed better comprehensive performance compared with the literature's. The work of this paper has a great guiding significance for the design and preparation of neutron absorbing composites with high thermal neutron shielding performance as well as high strength and plasticity.

Determination the Uranium Concentration in Some Plants and Soils From Baghdad Using CN-85 Track Detector

The aim of this study is to measure the concentrations of uranium for (17) samples of plants and soils that collected from different areas in Baghdad city. The technique is based on the calculation of the probability of registration α–particles by using the CN-85 track detector that is one of solid state nuclear track detectors SSNTDs and fission fragment track technique. The nuclear reaction used as source of uranium fission fragment is U-235 (n-f) obtained by the bombardment of U-235 with thermal neutrons emitted from (Am-Be) neutron source with thermal neutron flux (5000n/cm2.s) for seven days. The concentration of uranium was calculated by comparison with the standard samples. The concentrations of uranium in plants samples was ranging between (0.8-2.37 ppm) with average of (1.59 ppm). The concentrations of uranium in soils samples was ranging between (3.67-3.99ppm) with average of (3.82ppm). All results are agreement with the permissible limit and less than the allowed limit (11.7 ppm) from UNSCEAR.

The performance simulation of the LiH-SiC-based Fast Neutron Detector for harsh environment monitoring using Geant4 and TCAD

In this work, Lithium Hydride (LiH) and Silicon Carbide (SiC) based sensor for fast neutron detection has been reported for the first time. LiH is used as a hydrogenous converter material for converting neutrons to charged particles (recoil protons) and wide band-gap semiconductor material SiC is employed as a charged particle sensing detector in the form of a Schottky Barrier Diode (SBD). Geant4 Monte-Carlo simulations have been performed for the optimization of the LiH converter layer thickness for different mono-energetic neutron sources as well as Am-Be standard neutron source. The optimization of LiH thickness is essential to attain the maximum neutron detection efficiency. The simulation study has revealed that the maximum neutron detection efficiency of ∼ 0.1% can be achieved at ∼800μm thickness of LiH for the Am-Be neutron source. Furthermore, a device simulation software (Silvaco TCAD) has been utilized to investigate the radiation-induced effects on the electrical characteristics of a SiC-based SBD detector. The gamma and neutron irradiation induced damage model has been developed from the experimental deep level defects information reported in the literature for the TCAD simulation study. The effects on different device parameters viz., ideality factor, Schottky barrier height, series resistance, built-in-potential, effective carrier concentration, image force lowering, carrier removal rate, etc., due to 60Co-gamma (dose=100 Mrad) and 1 MeV equivalent neutron irradiation (fluence up to 1016 neutrons/cm2) have also been reported. The 1 MeV equivalent neutron fluence is the fluence of 1 MeV neutrons producing the same damage in a detector material as induced by an arbitrary particle fluence with specific energy distribution. This model will be helpful in predicting the performance and behavior of the SiC-based devices which are subjected to heavy irradiation. The study reveals that the effect on the electrical characteristics of SiC-based SBD detectors is tolerable up to the fluence of 1015 neutrons/cm2 at room temperature, but preventive measures must be adopted to reduce the leakage current at elevated temperatures.

Investigation of thermal neutron detection capability of a CdZnTe detector in a mixed gamma-neutron radiation field

Abstract The aim of this study was to investigate the thermal neutron measurement capability of a CdZnTe detector irradiated in a mixed gamma-neutron radiation field. A CdZnTe detector was irradiated in one of the irradiation tubes of a 241Am-Be source unit to determine the sensitivity factors of the detector in terms of peak count rate (counts per second [cps]) per neutron flux (in square centimeters per second) [cps/neutron·cm−2·s−1]. The CdZnTe detector was covered in a 1-mm-thick cadmium (Cd) cylindrical box to completely absorb incoming thermal neutrons via 113Cd(n,γ) capture reactions. To achieve, this Cd-covered CdZnTe detector was placed in a well-thermalized neutron field (f-ratio = 50.9 ± 1.3) in the irradiation tube of the 241Am-Be neutron source. The gamma-ray spectra were acquired, and the most intense gamma-ray peak at 558 keV (0.74 γ/n) was evaluated to estimate the thermal neutron flux. The epithermal component was also estimated from the bare CdZnTe detector irradiation because the epithermal neutron cutoff energy is about 0.55 eV at the 1-mm-thick Cd filter. A high-density polyethylene moderating cylinder box can also be fitted into the Cd filter box to enhance thermal sensitivity because of moderation of the epithermal neutron component. Neutron detection sensitivity was determined from the measured count rates from the 558 keV photopeak, using the measured neutron fluxes at different irradiation positions. The results indicate that the CdZnTe detector can serve as a neutron detector in mixed gamma-neutron radiation fields, such as reactors, neutron generators, linear accelerators, and isotopic neutron sources. New thermal neutron filters, such as Gd and Tb foils, can be tested instead of the Cd filter due to its serious gamma-shielding effect.

Experimental characterization of a long counter for neutron fluence measurement

A De Pangher-type long counter was constructed at Peking University for neutron fluence measurement. The responses and effective centers of the long counter were calculated using the Monte Carlo method. The variations with neutron energy of the position of the effective center, calculated by the Monte Carlo code, was experimentally validated with a 241AmBe neutron source and several mono-energetic neutron sources ranging from 100 keV to 6 MeV. Long counter calibration was performed using a radionuclide source from Peking University, and neutron fluence measured by the calibrated long counter was successfully compared to values determined with reference instruments (two recoil proton counters and a 238U fission chamber). The relative deviations were lower than 6% along the whole energy range. The calibrated long counter was thus successfully applied in the calibration of a Bonner sphere spectrometer.

Comparison of the pulse shape discrimination performance of plastic scintillators coupled to a SiPM

We report on the pulse shape discrimination (PSD) performance of plastic scintillators manufactured by Eljen Corporation and Amcrys. In this study we investigate the fast neutron and gamma performance of the plastic scintillators when coupled to the SensL J-series silicon photomultiplier (SiPM) and read out with fast waveform digitisers with an ADC resolution of 14-bits and a sample rate of 500 MS/s. The investigation observes a significant PSD performance increase for the SensL J-series SiPM in comparison to the previous C-series, and also for the latest variants of plastic scintillator from both suppliers.Analysis was performed using a Synchronous Charge Integration Pulse Shape Discrimination (PSD) algorithm which was applied to data acquired from a mixed fast neutron/gamma radiation field from an AmBe neutron source. The collected pulses were processed offline with the energy and PSD parameters calculated. The quality of the PSD performance was characterised by a common figure of merit (FoM). The best n-γ separation was found by the newer Eljen EJ-276 scintillator with a FoM value of 3.03 ± 0.03 at an energy of 1.5 MeV gamma equivalent. The Amcrys UPS-113NG material achieved a FoM value of 2.60 ± 0.04.

Neutron Activation Analysis of Some Building Material

Neutron activation analysis (NAA), based on the comparator method, has the potential to fulfill the requirements of a primary ratio method as defined in 1998 by the Comité Consultatif pour la Quantité de Matière — Métrologie en Chimie (CCQM, Consultative Committee on Amount of Substance — Metrology in Chemistry Studies of Neutron Activation Analysis (NAA) have been carried out on different Egyptian building material samples. The technique of neutron activation analysis is based on the measurement of radiation released by the decay of radioactive nuclei formed by neutron irradiation of the material. The most suitable source of neutrons for such an application is usually a research reactor. The samples that can be analyzed with this method stem from a number of different fields, including medicine, nutrition, biology, chemistry, forensics, the environment and mining. Neutron activation analysis can be performed in a variety of ways. This depends on the element and the corresponding radiation levels to be measured, as well as on the nature and the extent of interference from other elements present in the sample. Most of the methods used are non-destructive, based on the detection of gamma radiation emitted by the irradiated material after or during the irradiation. Next to education and training, neutron activation analysis is the most widely used application of research reactors. Almost any reactor operating at 10-30 kilowatt of thermal power is capable of providing a sufficient neutron flux to irradiate samples for selective applications of this analysis technique. Another method of NAA by using two Am-Be isotopic neutron sources of activity 5 Ci were used in this investigation. The accomplished gamma rays were measured using 70 % HPGe spectrometer. This work demand to estimate the elements contained in cement products and its quality control. X-ray Fluorescence (XRF) measurements were done for confirming our results, and for determining the average neutron flux of 3.7× 103 n/cm2sec. The Natural radioactivities of these samples were measured before the analysis to know the background level of 40K, 238U and232Th nuclei. The results investigated that NAA agree with the results of XRF and the world range of the cement concentration of the essential elements Ca, Al, Na, Fe, Mn, V, Sr and Si.

Tritium release from molten salt FLiNaK under low flux neutron irradiation with an AmBe neutron source

For examination of tritium transport in a self-cooled liquid blanket system using fluoride molten salt breeder/coolants, a series of neutron irradiation experiments have been conducted under the low flux neutron environment with an AmBe neutron source at the OKTAVIAN facility of Osaka University in Japan. Tritium release from molten salt FLiNaK (LiF-NaF-KF) at 773 K is evaluated with the neutron source, energies of which are up to 10.7 MeV and the intensity is ∼2 × 106 n/s. To measure the amount of tritium released from FLiNaK and permeating through an Inconel crucible, TF and/or T2O, and T2 are recovered separately using water bubblers and CuO oxidation beds. The amount of tritium recovered for one week in a steady state, 12.0 Bq, has closed to the calculated value, 10.6 Bq, obtained with MCNP6. More than 98% of the recovered tritium from FLiNaK has been TF and/or T2O. The overall mass transfer coefficients for tritium released from a FLiNaK free surface have been evaluated from the transient changes in recovered tritium and those are comparable to the value for HF released from FLiBe previously reported in the literature.

Investigation on the Neutron Transmission of B4C/CF/PI/AA6061 Hybrid Composite Laminates: Experimental and Numerical Simulation Results

A novel neutron shielding B4C/CF/PI/AA6061 composite laminate (NSCL) with different layups containing 10 to 50 wt% of boron carbide (B4C) particles was successfully fabricated using a hot molding process. The effects of different B4C loadings and various configurations on the neutron transmission of the NSCLs were evaluated correspondingly. The MCNP 5.0 program was used to probe the neutron transmission mechanism of the NSCLs. The results showed that B4C particles are an effective absorbent, and neutron transmission of the NSCLs decreased with the increment of layups, B4C loadings, and the laminate thickness. Fast neutrons emitted from a 241Am-Be neutron source were first moderated by low atomic elements (hydrogen) and then absorbed by 10B nuclide contained in the B4C particles. Numerical simulation corroborated the experimental testing results.

Evaluation of the response of a Bonner Sphere Spectrometer with a 6LiI detector using 3D meshed MCNP6.1.1 models

In order to undertake further studies on neutron spectra deconvolution in radiotherapy LinAc bunkers after using high megavolts treatment beams, it has been calculated the theoretical Response Function for a widespread neutron Bonner Sphere Spectrometer (BSS) exposed to arbitrary neutron sources. The neutron response function of the Bonner spectrometer is of essential importance for its neutron spectrum unfolding procedure and is directly related to the quality of the unfolded spectrum. Response detector curves from 10 keV to 20 MeV have been obtained by Monte Carlo (MC) simulation with MCNP6.1.1, where the use of unstructured mesh geometries is introduced as a novelty. In order to validate the accuracy of the MCNP6 simulation, we have used the detector model to measure an 241Am-Be neutron source, and the obtained neutron counts of the spectrometer and simulated counts are found to be very consistent, with a relative error below 10%. This comparison shows that the estimation of the Bonner sphere neutron response by MCNP6 is highly precise.

Towards radiation hard converter material for SiC-based fast neutron detectors

In the present work, Geant4 Monte-Carlo simulations have been carried out to study the neutron detection efficiency of the various neutron to other charge particle (recoil proton) converter materials. The converter material is placed over Silicon Carbide (SiC) in Fast Neutron detectors (FNDs) to achieve higher neutron detection efficiency as compared to bare SiC FNDs. Hydrogenous converter material such as High-Density Polyethylene (HDPE) is preferred over other converter materials due to the virtue of its high elastic scattering reaction cross-section for fast neutron detection at room temperature. Upon interaction with fast neutrons, hydrogenous converter material generates recoil protons which liberate e-hole pairs in the active region of SiC detector to provide a detector signal. The neutron detection efficiency offered by HDPE converter is compared with several other hydrogenous materials viz., 1) Lithium Hydride (LiH), 2) Perylene, 3) PTCDA . It is found that, HDPE, though providing highest efficiency among various studied materials, cannot withstand high temperature and harsh radiation environment. On the other hand, perylene and PTCDA can sustain harsh environments, but yields low efficiency. The analysis carried out reveals that LiH is a better material for neutron to other charge particle conversion with competent efficiency and desired radiation hardness. Further, the thickness of LiH has also been optimized for various mono-energetic neutron beams and Am-Be neutron source generating a neutron fluence of 109 neutrons/cm2. The optimized thickness of LiH converter for fast neutron detection is found to be ∼ 500 μm. However, the estimated efficiency for fast neutron detection is only 0.1%, which is deemed to be inadequate for reliable detection of neutrons. A sensitivity study has also been done investigating the gamma background effect on the neutron detection efficiency for various energy threshold of Low-Level Discriminator (LLD). The detection efficiency of a stacked structure concept has been explored by juxtaposing several converter-detector layers to improve the efficiency of LiH-SiC-based FNDs . It is observed that approximately tenfold efficiency improvement has been achieved—0.93% for ten layers stacked configuration vis-à-vis 0.1% of single converter-detector layer configuration. Finally, stacked detectors have also been simulated for different converter thicknesses to attain the efficiency as high as ∼ 3.25% with the help of 50 stacked layers.

Preparation, shielding properties and mechanism of a novel neutron shielding material made from natural Szaibelyite resource

In this paper, different amounts of Szaibelyite (100, 300, 500, 700 and 900) were used to prepare Szaibelyite/epoxy resin composites for neutron shielding. Microstructures of fracture surfaces of the prepared composites were tested by scanning electron microscope while the neutron shielding properties were determined by Am-Be neutron source. It was found that the shielding percentage of all composites increases with the addition of Szaibelyite as well as the thickness of composites increment. M5 was found to be the best neutron shielding composite, the shielding percentage was nearly 100% when the thickness of M5 chosen to be 1.5 cm. Furthermore, the shielding mechanism of M5 for neutron shielding could be concluded as follows: Firstly, H, O, C, Fe and B-11 isotope contained in M5 slowed down the neutron to thermal neutron, which then absorbed by the B-10 isotope. According to the achieved results, the performance of the new neutron shielding material owns the value of practical application.

Measurement of thermal neutrons reflection coefficients for two-layer reflectors

In this research, thermal neutrons albedo coefficients and relative number of excess counts have been measured experimentally for different thicknesses of two-layer reflectors by using 241Am-Be neutron source (5.2Ci) and BF3 detector. Our used reflectors consist of two-layer which are combinations of water, graphite, polyethylene, and lead materials. Experimental results reveal that thermal neutron reflection coefficients slightly increased by addition of the second layer. The maximum value of growth for thermal neutrons albedo is obtained for lead-polyethylene compound (0.72 ± 0.01). Also, there is suitable agreement between the experimental values and simulation results by using MCNPX code.

Multi-source irradiation facility with improved space configuration for neutron activation analysis: Design optimization

A neutron irradiation facility consisting of six 241Am-Be neutron sources of 30 Ci total activity and 6.6 × 107 n/s total neutron yield is designed. The sources are embedded in a cubic paraffin wax, which plays a dual role as both moderator and reflector. The sample passage and irradiation channel are represented by a cylindrical path of 5 cm diameter passing through the facility core. The proposed design yields a high degree of space symmetry and thermal neutron homogeneity within 98% of flux distribution throughout the irradiated spherical sample of 5 cm diameter. The obtained thermal neutron flux is 8.0 × 104 n/cm2.s over the sample volume, with thermal-to-fast and thermal-to-epithermal ratios of 1.20 and 3.35, respectively. The design is optimized for maximizing the thermal neutron flux at sample position using the MCNP-5 code. The irradiation facility is supposed to be employed principally for neutron activation analysis.

Low Activation-Modified High Manganese-Nitrogen Austenitic Stainless Steel for Fast Reactor Pressure Vessel Cladding

Low and free nickel austenitic stainless steel alloys were developed successfully and proposed to be used as a liquid sodium coolant fast reactor pressure vessel cladding. A standard austenitic stainless steel SS316L (AISI 316L) was produced as a reference sample. The nickel content was partially or totally replaced by manganese and nitrogen. The microstructure of the produced stainless steel alloys was investigated using Schaeffler diagram, optical microscopy and X-ray diffraction patterns (XRD). Mechanical properties of the developed stainless steel grads were investigated using Vickers hardness, impact and tensile tests at room temperature. Sodium chloride was used to study the corrosion rate of the investigated alloys by open circuit potential technique. Slow and total slow neutrons removal cross sections were measured using 241 Am-Be neutron source and highly calibrated He-3 detector. Eight gamma ray lines which emitted from 60 Co and 232 Th radioactive sources and HPGe detector were used to study the attenuation parameters of the produced alloys. Metallography, Schaeffler diagram and XRD results showed that all the produced stainless steels are mainly of austenite phase with a small ferrite phase. The developed manganese-nitrogen stainless steels showed higher hardness, yield and ultimate tensile strength than SS316L. The elongation of developed stainless steels is relatively lower than the standard SS316L. The impact toughness was reduced with replacement of Ni by Mn. The developed manganese stainless steels have a higher total slow removal cross section than SS316L. On the other hand, the slow neutron and gamma rays have nearly the same behavior for all studied stainless steels.

Simulation of the neutron response matrix of an EJ309 liquid scintillator

The neutron response matrix is the basis for measuring the neutron energy spectrum through unfolding the pulse height spectrum detected with a liquid scintillator. Based on the light output of the EJ309 liquid scintillator and the related reaction cross sections, a Monte Carlo code is developed to obtain the neutron response matrix. The effects of the related reactions, the contributions of different number of neutron interactions and the wall effect of the recoil proton are discussed. With the obtained neutron response matrix and the GRAVEL iterative unfolding method, the neutron energy spectra of the 252Cf and the 241AmBe neutron sources are measured, and the results are respectively compared with the theoretical prediction of the 252Cf neutron energy spectrum and the previous results of the 241AmBe neutron energy spectra.

DETERMINATION OF NEUTRON EFFECTIVE DOSES IN WHOLE BODY POINT SOURCE EXPOSURES.

The objective of this work is to obtain fluence to effective dose conversion coefficients for neutron point sources, using the GEANT4 toolkit. These calculations aim to investigate the aspects of neutron transport in the human body through Monte Carlo simulation using the International Commission on Radiological Protection (ICRP) voxel phantoms, described in its publication 110. A benchmarking of the code was made for the case of monoenergetic plane parallel neutron beam in the antero-posterior (AP) irradiation geometry and organ absorbed dose conversion coefficients were compared with those found in the ICRP publication 116. The results showed good agreement with ICRP results in the studied energy range. Conversion coefficients were presented for specific conditions with 241Am-Be and 252Cf point neutron sources 1 m away from the phantom in the AP geometry.

Investigation into a suitable scintillator and coded-aperture material for a mixed-field radiation imaging system

Monte-Carlo modelling (MCNPX) methods have been employed to conduct an investigation into a suitable scintillator and coded-aperture material for a scintillator based mixed-field radiation imaging system. Single stilbene crystal, pure and 6Li-loaded plastic scintillators were simulated and their neutron/gamma detection performance compared when exposed to the spontaneous fission spectrum produced by 252Cf. The most suitable candidate was then incorporated into a scintillator based mixed-field coded-aperture imaging system. Coded-aperture models made of three W and 113Cd compositions were tested in different neutron/gamma environments with a square W collimator modelled around the aperture. Each simulation involved recording the interactions of neutron events in organic solid scintillator, whose neutron/gamma detection performance was assessed prior to the coded-aperture material investigation. Three coded-aperture material compositions have been tested with the simulated 252Cf spontaneous fission as well as 241AmBe neutron sources. Results generally claim very good detection sensitivity and spatial resolution for the radioactive sources located in the centre of the aperture.

Gamma and neutron irradiation effects on the structural and optical properties of potash alum crystals

Abstract The effects of gamma and neutron irradiation on the structural and optical properties have been presented in this article. The potash alum crystals were grown by the slow evaporation technique. The crystals were then exposed with 60Co gamma rays and Am-Be neutron source. The radiation induced effects were analysed using XRD, UV–visible techniques. The refractive index was determined by the Brewster’s angle method. There has been an increase in the diffraction peak intensity with the increase in irradiation doses, which indicates an increase in crystalline perfection of crystals. There has been a marginal change in dislocation density, micro strain and distortion parameter, which is attributed to deformation of the structure due to irradiation. The UV visible study reveals a shift in transparency cut off wavelength towards lower wavelength up to 70 kGy of gamma dose and this has resulted in an increase in optical band gap energy with increase in gamma dose. There is a considerable increase in the optical band gap after five hour neutron irradiation.

Investigation of the Shielding Capability of Concrete Matrixed Colemanite Reinforced Shielding Material

In this study, neutron shielding capabilities of colemanite reinforced materials with concrete matrix were investigated. Containing 70% by weight of colemanite a boron compound 20, 40 and 60 mm samples were prepared and their neutron shielding properties were studied. Composite material prepared by hand mixing by the weight of different ratios of colemanite, so that the obtaining process of the samples are extremely smooth. As an excellent thermal neutron absorber, colemanite is the crucial part of the composite. Beside this, concrete act as a role of thermalizing the energetic neutrons, by the way the colemanite can absorb the particles. It was observed that shielding capability of the composite material increased with increasing its thickness. The performance of neutron radiation shielding of specimens prepared by adding 70% colemanite by weight into concrete material was experimentally analyzed with Am-Be neutron source measurement system.