Design and simulation of neutron radiography system based on 241Am–Be source

Neutron imaging is extended rapidly as a means of non-destructive testing (NDT) of materials. Various effective parameters on the image quality are needed to be studied for neutron radiography system with good resolution. In the present study a portable system of neutron radiography has been designed using ²⁴¹Am–Be neutron source. The effective collimator parameters were calculated to obtain relatively pure, collimated and uniform neutron beam. All simulations were carried out in two stages using MCNPX Monte Carlo code. In the first stage, different collimator configurations were investigated and the appropriate design was selected based on maximum intensity and uniformity of neutron flux at the image plane in the outlet of collimator. Then, the overall system including source, collimator and sample was simulated for achieving radiographic images of standard samples. Normalized thermal neutron fluence of 2.61×10^?5 cm^?2 per source particle with n/γ ratio of 1.92×10⁵ cm^?2 μSv^?1 could be obtained at beam port of the designed collimator. Quality of images was assessed for two standard samples, using radiographic imaging capability in MCNPX. The collimated neutron beam in the designed system could be useful in a transportable exposure module for neutron radiography application.     

A transportable neutron radiography system

A transportable neutron radiography system, incorporating a 50 mg ²⁵²Cf source, has been simulated using the MCNPX code. The materials considered were compatible with the European Union Directive on ‘Restriction of Hazardous Substances’ (RoHS) 2002/95/EC, hence excluding the use of cadmium and lead. The design was optimized with respect to neutron moderation, shielding and collimation. High density polyethylene was chosen as the material for moderator and also shielding, which was further enhanced with layers of bismuth and borated polyethylene. Variable values for the collimator ratio were calculated. With suitable aperture and collimator design it was possible to optimize the neutron radiography parameters. Beam filters also were treated in order to improve the results. The proposed system has been considered with a wide range of radiography parameters, which are comparable with neutron radiography facilities from low power reactors.     

A simple and sensitive particle induced gamma-ray emission method for non-destructive quantification of lithium in lithium doped Nd2Ti2O7 ceramic sample

A particle induced gamma-ray emission method using proton beam in conjunction with in situ current normalization approach was standardized for non-destructive determination of low Z element lithium and was applied for quantification of Li in lithium doped neodymium dititanate (Nd₂Ti₂O₇) ceramic sample. Thick pellets of heat treated samples, their precursors and Li standards were prepared separately by homogeneously mixing with cellulose and fixed amount of F used for in situ current normalization. For validation of the method, four synthetic samples were also analyzed. Samples and standards were irradiated with 4?MeV proton beam (~5?nA current) from folded tandem ion accelerator (FOTIA) BARC, Mumbai. Characteristic ??-rays of 478?keV from ⁷Li to 197?keV from ¹⁹F were measured by high resolution ??-ray spectrometry. The Li concentrations determined in the six samples were in the range of 0.29?C0.85?wt%. The Li contents in heat treated samples gave the idea about loss of Li compared to their precursors.     

Design and evaluation of a TNA explosive-detection system to screen carry-on luggage

Thermal neutron analysis (TNA) technology has been used for the non-destructive detection of explosives. The system uses a relatively weak ²⁵²Cf neutron source (1.03·10⁷ n/s) and two 3"×3" NaI(Tl) detectors. The presence of explosives is confirmed via detection of the 10.83 MeV prompt gamma-ray associated with nitrogen decay. The MCNP4A code was used to simulate the neutron and gamma transport through the system. The thermal neutron flux in the activation position was measured using gold and indium foils. The measured thermal neutron flux was lower, by not more than 9.5%, than that of simulation. In this report the results of the preliminary tests on the system are described.     

Eine 3mg-252Cf-Bestrahlungsanordnung-Aufbau und Anwendungen

In the last years Cf-252 neutron sources will be incresingly used for nuclear analytical purposes. In the Central Institute of Isotope and Radiation Research, Leipzig, an irradiation equipment with a 3mg²⁵²Cf neutron source was constructed. It reaches thermal neutron fluxes of about 10⁷ n·cm^–2·s^–1. The construction of this equipment and the different moderation geometries are described. Possibilities of the application for neutron induced autoradiography, neutron radiography and neutron activation analysis are demonstrated on examples.     

Measurement of the neutron capture cross-section of 238U using the neutron activation technique

The ²³⁸U(n, ??)²³⁹U reaction cross-section at average neutron energy of 3.7?±?0.3?MeV from the ⁷Li(p, n)⁷Be reaction has been determined using activation and off-line ??-ray spectrometric technique. The ²³⁸U(n, ??)²³⁹U and ²³⁸U(n, 2n)²³⁷U reaction cross-sections at average neutron energy of 9.85?±?0.38?MeV from the same ⁷Li(p, n)⁷Be reaction have been also determined using the above technique. The experimentally determined ²³⁸U(n, ??)²³⁹U and ²³⁸U(n, 2n)²³⁷U reaction cross-sections were compared with the evaluated data of ENDF/B-VII, JENDL-4.0, JEFF-3.1 and CENDL-3.1. The experimental values were found to be in general agreement with the evaluated value based on ENDF/B-VII, and JENDL-4.0 but not with the JEFF-3.1 and CENDL-3.1. The present data along with literature data in a wide range of neutron energies were interpreted in terms of competition between different reaction channels including fission. The ²³⁸U(n, ??)²³⁹U and ²³⁸U(n, 2n)²³⁷U reaction cross-sections were also calculated theoretically using the TALYS 1.2 computer code and were also found to be in agreement experimental data.     

Studies on the production feasibility of 64Cu by (n, p) reactions on Zn targets in Dhruva research reactor

⁶⁴Cu is an useful radionuclide for both PET imaging and targeted therapy, as it decays by three different modes, namely, electron capture (41%), ??^? (40%) and positron emission (19%). ⁶⁴Cu is generally produced by ⁶⁴Ni (p, n) reaction in a cyclotron for medical use. High specific activity ??no carrier added?? grade ⁶⁴Cu by ⁶⁴Zn (n, p) route is an alternative for research studies and was hence explored. 10?mg zinc foil target (48.63% in ⁶⁴Zn) was irradiated in the medium flux reactor Dhruva at a thermal neutron flux of ~5.6?×?10¹³ n?cm^?2?s^?1 for 3?days. The irradiated Zn foil was dissolved in 5?mL 10?M HCl and ⁶⁴Cu was separated by anion exchange chromatography (Dowex 1?×?8; 100?C200 mesh) at 3?M HCl conditions. ⁶⁴Cu radioactivity content and its radionuclide purity were ascertained by ??-ray spectrometry using HPGe detector coupled to a 4?K multichannel analyser system. Radiochemical separation yielded a radionuclidic purity of 99.9% ⁶⁴Cu.     

Characterization of a new external neutron beam facility at the Ohio State University

A thermal neutron beam facility has been designed and implemented at the Ohio State University Research Reactor. A project is underway to construct a large vacuum chamber such that the facility could have neutron depth profiling and neutron radiography capabilities as intended. The neutron beam is extracted from the reactor through a neutron collimator emplaced in Beam Port #2. The neutron spectrum entering the neutron collimator was unfolded from foil activation analysis results and also simulated with a full reactor core model in the MCNP Monte Carlo code. The neutron collimator uses polycrystalline bismuth as a gamma ray filter and single-crystal sapphire as a fast neutron filter. The beam is defined by multiple 3.0 cm diameter apertures made of borated aluminum. Characterization of the beam was performed using foil activation to find the flux and a low-budget neutron imaging apparatus to see the beam profile. The modulation transfer function was calculated to offer insight into the resolution of the imaging system and the collimation of the beam. The neutron collimator delivers the filtered thermal neutron beam with a ~4 cm diameter and a thermal equivalent flux of (1.27 ± 0.03) × 10⁷ n/(cm²s) at 450 kW power at the end of the collimator.     

Measurements of the neutron spectra from the 7Li(p,n) accelerator based neutron source: Position and angular dependences

Thick target ⁷Li(p,n) neutron spectra were measured using a ³He ion chamber in the proton energy range of 1.95 to 2.30 MeV. The fast neutron spectra were collected for various distances from the lithium target as well as for various neutron emission angles. By unfolding the ³He raw data with the iterative van Cittert algorithm, the neutron fluence spectra were obtained. The ³He measured neutron spectra were compared with both analytically computed and Monte Carlo simulated spectra to account for neutron scatterings in the lithium target assembly and in the experimental area. To verify the accuracy of the neutron dose computation, the fast neutron kerma was obtained for each neutron spectrum using the fluence to kerma conversion coefficients and was compared with the measured neutron dose using tissue-equivalent proportional counters. In the position dependence investigation at the 0° emission angle, the analytically computed neutron kerma overestimates the experimental kerma by a factor of two mainly due to neutron moderation. The corresponding neutron kerma from the ³He measured spectra were in agreement with the neutron doses measured using tissue-equivalent proportional counters within 20% for lower proton energies, but the discrepancy increased to ～50% for higher proton energies. In the angular distribution investigation, a notable discrepancy between measured and computed neutron spectra were observed due to the neutron scattering effects in the target assembly and experimental room.     

Fabrication of a micro-direct methanol fuel cell using microfluidics

A direct-methanol fuel cell containing three parts: microchannels, electrodes, and a proton exchange membrane (PEM), was investigated. Nafion resin (NR) and polystyrene-block-poly(ethylene-ran-butylene)-block-polystyrene (PS) were used as PEMs. Preparation of PEMs, including compositing with other polymers and their solubility, was performed and their proton conductivity was measured by a four point probe. The results showed that the 5 % Nafion resin has lower conductivity than the 5 % PS solution. The micro-fuel cell contained two acrylic channels, PEM, and two platinum catalyst electrodes on a silicon wafer. The assembled micro-fuel cells used 2 M methanol at the flow rate of 1.5 mL min^?1 in the anode channel and 5 × 10^?3 M KMnO₄ at the flow rate of 1.5 mL min^?1 in the cathode channel. The micro-fuel cell with the electrode distance of 300 ??m provided the power density of 59.16 ??W cm^?2 and the current density of 125.60 ??A cm^?2 at 0.47 V.     

Amperometric determination of nitric oxide at a carbon nanotube modified electrode with redox polymer coating

A novel chemically modified electrode based on an osmium complex-containing redox polymer film coated on single-walled carbon nanotube (SWNT) modified glassy carbon electrode (GCE) has been described for the determination of nitric oxide. The results showed that the oxidation current increased significantly at the SWNT/redox polymer coated GCE, as compared to that observed on a bare GCE- and SWNT-modified GCE. Amperometric measurement was carried out at the potential of +0.80?V (vs. Ag|AgCl) and the current response to NO was found to be directly proportional to its concentration in the range from 2.0?×??0^?? to 4.0?×??0^?? M, and the detection limit was estimated to be 5.0?×??0^?? M.     

The comparison of four neutron sources for Prompt Gamma Neutron Activation Analysis (PGNAA) in vivo detections of boron

A Prompt Gamma Ray Neutron Activation Analysis (PGNAA) system, incorporating an isotopic neutron source has been simulated using the MCNPX Monte Carlo code. In order to improve the signal to noise ratio different collimators and a filter were placed between the neutron source and the object. The effect of the positioning of the neutron beam and the detector relative to the object has been studied. In this work the optimisation procedure is demonstrated for boron. Monte Carlo calculations were carried out to compare the performance of the proposed PGNAA system using four different neutron sources (²⁴¹Am/Be, ²⁵²Cf, ²⁴¹Am/B, and DT neutron generator). Among the different systems the ²⁵²Cf neutron based PGNAA system has the best performance.     

Design, testing and optimization of a neutron radiography system based on a Deuterium–Deuterium (D–D) neutron generator

Simulations show that significant improvement in imaging performance can be achieved through collimator design for thermal and fast neutron radiography with a laboratory neutron generator. The radiography facility used in the measurements and simulations employs a fully high-voltage-shielded, axial D–D neutron generator with a radio frequency driven ion source. The maximum yield of such generators is about 10¹⁰ fast neutrons per seconds (E = 2.45 MeV). Both fast and thermal neutron images were acquired with the generator and a Charge Coupled Devices camera. To shorten the imaging time and decrease the noise from gamma radiation, various collimator designs were proposed and simulated using Monte Carlo N-Particle Transport Code (MCNPX 2.7.0). Design considerations included the choice of material, thickness, position and aperture for the collimator. The simulation results and optimal configurations are presented.     

Comparison between 2-hydroxypropyl-��-cyclodextrin and 2-hydroxypropyl-��-cyclodextrin for inclusion complex formation with danazol

Complexation in solution between danazol and two different cyclodextrins [2-hydroxypropyl-??-cyclodextrin (HP-??-CD) and 2-hydroxypropyl-??-cyclodextrin (HP-??-CD)] was studied using phase solubility analysis, and one- and two-dimensional ¹H-NMR. The increase of danazol solubility in the aqueous cyclodextrin solutions showed a linear relationship (A_L profile). The apparent stability constant, K _1:1, of each complex was calculated and found to be 51.7 × 10³ and 7.3 × 10³ M^?1 for danazol?CHP-??-CD and danazol?CHP-??-CD, respectively. ¹H-NMR spectroscopic analysis of varying ratios of danazol and the different cyclodextrins in a mixture of EtOD?CD₂O confirmed the 1:1 stoichiometry. Cross-peaks, from 2D ROESY ¹H-NMR spectra, between protons of danazol and H3?? and H5??of cyclodextrins, which stay inside the cyclodextrin cavity, proved the formation of an inclusion complex between danazol and the cyclodextrins. For HP-??-CD, the inclusion complex is formed by entrance of the isooxazole and the A rings of danazol in the cyclodextrin cavity. For HP-??-CD, two different inclusion structures may exist simultaneously in solution: one with the isooxazole and A ring in the cavity and the other with the C and D ring inside the cavity. DLS showed that self-aggregation of the CD??s was absent in the danazol HP-??-CD system up to a CD concentration of 10% and in the danazol HP-??-CD system up to a CD concentration of 5%.     

Design calculation of a horizontal thermal neutronic beam for neutron radiography at the Syrian MNSR

The computer code MCNP4C and the ENDF/B-V cross-section library were used to design calculation of a horizontal thermal beam for neutron radiography (NR) at Syrian MNSR and to evaluate the safety of the reactor after installation of the NR facility (NRF). Thermal, epithermal and fast neutron energy ranges were selected as <0.30 eV, 0.30 eV–10.0 keV and >10.0 keV, respectively. To produce a good neutron beam in terms of intensity and quality, bismuth (Bi) and silicon (Si) were used as photon and neutron filters, respectively. The ratio of L/D of the NRF ranges between 90 and 125. The thermal neutron flux at the beam exit plane can be varied from 1.836 × 10⁵ to 3.057 × 10⁵ n/cm² s. If such thermal neutron beam would be built into the Syrian MNSR, many scientific applications of the NR would be available.     

The application of NAA with irradiation at reduced temperature for the study of element losses during pretreatment of biological samples

The neutron capture gamma-ray spectroscopy facility assembled at the Institute of Radiochemistry, KfK (for analytical purposes) using a²⁵²Cf neutron source with a strength of 6·10⁷ n/s, has been used to check its applicability and sensitivity for quantitative analyses of ores. The analysis of Sm, Cd and Mn in phosphate and monazite rock samples has been carried out. The results from this study show a variation of about 25% from the values determined by RNAA method. This discrepancy could be mainly due to the low signal-to-background ratio observed which is caused by (i) scattering of the source gammarays by the target, and (ii) interference from the 2223.1 keV neutron capture hydrogen gamma-rays produced by the moderated materials and from their compton scattering in the detector. To overcome these difficulties we suggest to introduce a 2.5 cm thick polyethylene sheet between the detector⁶Li-cap shielding and the target as well as to increase the detection solid angle. Also the strength of the²⁵²Cf neutron source should be increased by an order of magnitude and the neutron beam should be collimated to obtain the optimal thermal neutron flux with a low level of²⁵²Cf gamma-rays. This can be achieved by setting up between the neutron source and the target a conical polyethylene collimator with a thickness of 10 cm containing a 1 cm thick lead sheet.     

Inelastic neutron scattering spectra of the hydrogen tungsten bronze H0.4WO3

Inelastic neutron scattering spectra of metallic H_0.4WO₃ show only the vibrations that would be expected of a metal hydroxide. The diffusion coefficient of the proton could not be detected from quasi-elastic scattering with the best available neutron spectrometer setting an upper limit to its value of 10^?7 cm²/sec. Both these results confirm that this material is correctly described as a hydroxide.     

Intramolecular hydrogen bonds C-H?N in bisheterocyclic compounds according to 1H and 13C NMR data

Data of ¹H and ¹³C NMR spectra show that in 2,2??-bipyridyl, 1-vinyl-2(2??-pyridyl)benzimidazole, 1-vinyl-3-vinylsulfanyl-5-(2-furyl)-1,2,4-triazole, and 1-vinyl-5-vinylsulfanyl-3-(2-furyl)-5-vinylthio-1,2,4-triazole exists a weak intramolecular hydrogen bond between the heterocyclic fragments. It causes a downfield shift of the bridging proton signal in the ¹H NMR spectrum by 0.6?C0.7 ppm and an increase in the corresponding direct coupling constant ¹³C-¹H by 1.5?C2.0 Hz. These variations in the spectral parameters can be efficiently used in the conformational analysis for establishing with the use of NMR method which conformers are predominantly populated in the heterocyclic compounds.