Determination of cadmium in zinc ores by thermal neutron absorption analysis

A method has been developed for routine determination of cadmium in zinc ores by thermal neutron absorption analysis, based on the attenuation of a thermal neutron flux passing through a neutron absorbing material. The thermal neutron flux is related to the⁵²V-activity induced in a vanadium detector, surrounded by pellets pressed from a mixture of powdered material with graphite. Besides cadmium, also the major constitutents zinc, iron and sulfur contribute significantly to the total attenuation of the thermal neutron flux. Calibration lines for these elements are worked out. All irradiations are carried out for 200 s in the partially thermalized neutron flux of a 5 Ci²²⁷Ac—Be isotope neutron source. After a decay of 30 s, the⁵²V-activity of the vanadium detector is measured for 400 s with a NaI(T1) scintillation detector. The analysis sequence, including the computation of the results from the counting data, is automated by means of a LSI—11 microprocessor with 12K×16 bit memory. Zinc ores, containing 0.02 to 1.45% cadmium, have been analyzed with a precision ranging from 12.6% to 0.54% relative. As a test for the reliability of the method, two NBS standard reference materials were analyzed in the same way as the zinc ore samples.     

Determination of manganese in natural waters by atomic absorption spectrometry with a carbon tube atomizer.

A method is described for the accurate and precise determination of alumina and silica in bauxites with the aid of a ²²⁷Ac-Be isotopic neutron source with a total neutron output of 10⁸ n s^-1. Three ores can be analysed in triplicate within 4 h, including the determination of the natural radioactivity of the ores. Samples are pellets pressed from a mixture of 4.5 g of powdered bauxite and 0.9 g of a wax as pelleting agent. The special flux distributions of the source allow irradiations at very different fast-to-thermal flux ratios without cadmium neutron absorbers. The drying and water re-uptake of bauxite and the natural radioactivity in these ores are discussed. The method was tested with several certified standards. A relative precision of 0.7% for the alumina determination was obtained for triplicate analyses. Vanadium is the only interfering element, the concentration of which should be determined separately or estimated; a correction procedure is given.     

A method for bulk hydrogen analysis based on transmission and back scattering of fast neutrons

A method was proposed for bulk hydrogen analysis. It is based on simultaneous detection of transmitted fast neutrons and back scattered thermal neutrons from the investigated samples by ³He detectors. The fast neutron beams were obtained from ²⁵²Cf and Pu–Be neutron sources. The experimental set-up as well as samples preparation were described. Incident thermal neutrons beams obtained from either ²⁵²Cf or Pu–Be sources, were used to investigate the samples by neutron backscattering. The results obtained from transmission and backscattering of fast neutrons were compared and discussed. The advantage and capabilities of the proposed method were presented. The results obtained using fast neutron beams are more sensitive than those obtained using thermal neutron beams. Validation procedures were proposed to credit the results.     

An241Am−Be source based thermal neutron activation analysis facility at KFUPM

An instrumental thermal neutron activation analysis facility based, on a 16 Ci²⁴¹ Am–Be source, a high resolution -ray spectrometry setup and a PC-based data acquisition system at KFUPM is described. The thermal neutron flux distribution was determined from the induced activities of high purity indium foils. The absolute thermal neutron flux was calculated from the activities of bare and cadmium-covered gold foils at a position of 3 cm from the soource at which the flux reaches a maximum. The facility tests were carried out with the determination of manganese concentrations in six types of industrially important steel samples. The result of 1.33% manganese in SS-304 steel sample was in excellent agreement with the literature value. The method is nondestructive, economical and ideal for bulk analysis.     

Determination of manganese in some pyrolusite ores of Myanmar by neutron activation analysis

Manganese in pyrolusite ores from various regions of Myanmar was determined by thermal neutron activation analysis using an Am(Be) neutron source. The induced activities of⁵⁶Mn were monitored by a -counting technique.     

Neutron beam preparation with Am–Be source for analysis of biological samples with PGNAA method

Material analysis with prompt gamma neutron activation analysis (PGNAA) requires a proper geometrical arrangement for equipments in laboratory. Application of PGNAA in analysis of biological samples, due to small size of sample, needs attention to the dimension of neutron beam. In our work, neutron source has been made of ²⁴¹Am–Be type. Activity of ²⁴¹Am was 20 Ci which lead to neutron source strength of 4.4 × 10⁷ neutrons per second. Water has been considered as the basic shielding material for the neutron source. The effect of various concentration of boric acid in the reduction of intensity of fast and thermal components of the neutron beam and gamma ray has been investigated. Gamma ray is produced by (α, n) reaction in Am–Be source (4.483 MeV), neutron capture by hydrogen (2.224 MeV), and neutron capture by boron (0.483 MeV). Various types of neutron and gamma ray dosimeters have been employed including BF₃ and NE-213 detectors to detect fast and thermal neutrons. BGO scintillation detector has been used for gamma ray spectroscopy. It is shown that the gamma and neutron radiation dose due to direct beam is of the same magnitude as the dose due to radiation scattered in the laboratory ambient. It is concluded that 14 kg boric acid dissolved in 1,000 kg water is the optimum solution to surround the neutron source. The experimental results have been compared with Monte Carlo simulation.     

Flux calculation in LSNAA using an <Superscript>241</Superscript>Am-Be source

The CITATION code based on neutron diffusion theory is used for flux calculation inside voluminous sample in prompt gamma activation analysis with an isotopic neutron source (²⁴¹Am-Be). The code used the specific parameters related to energy spectrum source, irradiation system materials (shielding, reflector, etc.), geometry and elemental composition of the sample. The flux distribution (thermal and fast) was calculated on three-dimensional geometry for the system: source, air, and polyethylene and water cylindrical sample of 125 liters. The thermal flux was calculated in series of points inside the sample, and agreed with the results obtained by measurements with good statistical uncertainty. The maximum thermal flux was measured at distance of 4.1 cm and calculated at 4.3 cm by the CITATION code. Beyond a depth of 7.2 cm, the ratio of thermal flux to fast flux increases up to twice and allows us the optimization of the detection system in the scope of in-situ PGNAA.     

Measurements of protein in limbs by IVNAA

A method for the determination of protein in human limbs by in vivo neutron activation analysis of their nitrogen content is described. The neutron flux is provided by a collimated 740 GBq (20 Ci) Pu–Be source. The 10.83 MeV thermal neutron capture gamma-rays from¹⁴N are detected by a 12.7 cm by 10.2 cm NaI(T1) detector. The nitrogen and hydrogen in an arm, the lower and the upper legs of two volunteers were measured with a statistical error around 3.6%–18% and 0.26%–0.56%, respectively. Also, the ratio of nitrogen to hydrogen in the limbs was compared. For a 1.000 s irradiation the dose equivalent was 0.3 mSv (30 mrem).     

Determination of227Ac in geological materials through neutron activation and alpha-spectrometric assay of induced228Th

A new route has been developed for the micro-determination of²²⁷Ac in geological materials by neutron activation. The method is based on intense neutron irradiation of the analysed samples followed by separation and α-spectrometric determination of²²⁸Th, the β-decay product of the 6.1 hrs²²⁸Ac isotope formed. Two alternatives are considered for analysis related to the origin of the analysed matrix. The high sensitivity of the method is documented by the determination of 10^?17 g²²⁷Ac/g sample. The method is successfully applied for age determination of five uranium containing materials and old uranium glass from Bohemia, CSSR.     

A borehole sonde using neutron activation technique

A simulated borehole sonde has been assembled, with an aluminium casing of 70 cm in length and 12 cm in inner diameter. It contains a 5 Ci Pu–Be source with a neutron yield of about 5.45·10⁶+10% n·cm^–2·s^–1, a shadow shield, and a Hp Ge detector of 15% efficiency and 2 keV FWHM for the 1.33 MeV line of⁶⁰Co. Evaluations of the assembly through the dependence of thermal neutron flux on water content, matrix composition and borehole configuration have been performed. Accordingly, thermal neutron flux distributions have been measured around the sonde and inside the ore in both the simulated dry and water filled borehole. From these measurements one could estimate that the effective moderating thickness of water is about 4 cm, while the volume matrix of the ore to be investigated by this assembly is a slab of about 8 cm width and a height of about 15 cm. It also follows that the uranium-thorium ore analysis method described in this work may become important as a field neutron activation technique.This work was financially supported by the IAEA under research project No. 3534/R.B.     

Determination of fluorine in zinc ores using an isotope neutron source based automated neutron activation analysis system

A method has been developed for routine determination of fluorine in zinc sulfide ores by activation with fast neutrons from a 6.6 Ci²²⁷Ac−Be isotope source and counting of the 4.5 to 7 MeV gamma-rays of the product nuclide¹⁶N. Samples and standards consist of pellets pressed from a mixture of powdered material with wax or graphite. Samples and standards alternate in a sequence of 20 seconds irradiation, 4 seconds decay and 20 seconds counting. This analysis sequence, including the computation of the analysis results from the counting data automated by means of a LSI-11 Microprocessor with 12K×16 bit memory. The zinc ores, containing 0.3 to 0.7% fluorine have been analysed with a precision ranging from 1.56 to 1.33% relative. As a test for the reliability of the method, three standard reference materials were analysed in the same way as the zinc ore samples.     

Synthesis of two complexes of cadmium(II) iodide with benzenecarbothioamide (BCTA): Crystal and molecular structure of [Cd(BCTA)2I2]

The reaction between CdI₂ and benzenecarbothioamide (BCTA) in ethanol produces crystalline [Cd(BCTA)₂I₂] but in 1,2-dichloroethane it yields [Cd(BCTA)I₂] in powder form. [Cd(BCTA)₂I₂] has been characterized by X-ray crystallography, and both complexes have also been identified by infrared spectra in the solid phase, ¹H NMR, and electrical conductivity studies in solution. The crystals of [Cd(BCTA)₂I₂] are monoclinic, space group P2₁/c, with a = 747.6(1) pm, b = 1958.7(13) pm, c = 1363.6(6) pm, β = 110.307(3)° and Z = 4. Least-squares refinement of the structure based on 4260 observations led to final discrepancy indices of R = 0.043 and R_w = 0.054. The geometry around the cadmium atom is slightly distorted from the tetrahedral. The BCTA is coordinated through the sulphur atom: CdI(1) = 277.2(1) pm, CdI(2) = 272.6(1) pm, CdS(1) = 256.7(3) pm, CdS(2) = 257.5(2) pm. The electrical conductivity studies and NMR and IR spectra are consistent with the structure found for [Cd(BCTA)₂I₂] and suggest a dimeric trans symmetric tetrahedral structure with halogen bridging (C_2h) for [Cd(BCTA)I₂].     

A new reflector structure for facility thermalizing D–T neutrons

A facility for thermalization of fast neutrons (14.2 MeV) emitted by compact deuterium–tritium (D–T) neutron generators (NGs) for thermal neutron activation analysis is proposed. Its final design is based on Monte Carlo calculations (MCNP5). To maximize the ratio between the thermal neutron flux and the total neutron flux and simultaneously to ensure the highest possible value of the thermal neutron flux at the output surface, the facility should consist of a two-layer reflector [tungsten (W)—the inner part, molybdenum—the outer part], a two-layer multiplier (W followed by lead), a moderator (polyethylene followed by magnesium fluoride) and a collimator (molybdenum and nickel near the output surface). For the D–T NG producing the maximum available neutron yield 10¹⁵ n s^?1, the facility provides the thermal neutron flux 2.0 × 10¹¹ n cm^?2 s ^?1 and a slightly higher fast neutron flux 2.3 × 10¹¹ n cm^?2 s^?1. To improve the ratio of the thermal neutron flux to the fast neutron flux (above 2.7) an addition of a silicon layer to the moderator and especially a proper adjustment and a threefold increase of the multiplier thickness is necessary.     

Thermal neutron activation analysis of Cu in its ores by using an241Am−Be neutron source

A thermal neutron activation method has been developed for the determination of Cu in Cu–Pb–Zn ores and chalcopyrite ore concentrates using the reaction⁶³Cu(N, )⁶⁴Cu. The samples were irradiated with thermal neutrons from an²⁴¹Am–Be neutron source and the annihilation -radiations of 0.511 MeV were counted on 3×3 NaI(T1) detector coupled with single channel pulse height analyzer. The method is nondestructive, economical and ideal for bulk analysis of ores with 1–16% Cu.     

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.     

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.     

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.     

Non-destructive determination of silicon in aluminium-silicon alloys by neutron activation analysis with a 227Ac-Be isotope neutron source

A method has been developed for the accurate determination of silicon in binary Al-Si alloys by instrumental neutron activation analysis with the aid of an ²²⁷Ac-Be isotope neutron source. Discs, weighing ca. 4 g, are irradiated at the maximum of the fast flux gradient in a rabbit lined internally with two layers of cadmium foil. Each irradiation is followed by an integral mode counting with discriminator baseline setting at 1400 keV. Standards consist of discs from high purity semiconductor silicon and aluminium (99.99%) metal. An iterative calculation procedure yields results with a relative precision from 0.6 to 8.1% for silicon contents between 13.5 and 0.5%. Triplicate analysis requires 21 min.