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Abstract  

Lead is determined in environmental samples and in rocks using the206,207,208Pb(p,xn)206Bi reaction. Bismuth is separated by anion exchange or by extraction with antimony diethyldithiocarbamate. Sources of errors such as volatilization of the matrix due to heating during the irradiation, variations of the abundance of the lead isotopes and the standardization were studied in detail. For concentrations between 11 mg/g and 3.7 g/g the relative standard deviation ranges from 2.6 to 5.4%. The detection limit is 10 ng/g.

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Abstract  

Electrophoretic focussing of ions was applied to the separation of fission products present in solutions of nuclear uranium fuel irradiated in various European reactors. By combining two separation methods, all the long-lived fission products could be determined individually and quantitatively by counting with a NaI(T1) and a GM detector of known detection efficiency. Radiography and autoradiography were used for semi-quantitative purposes. The concentrations of235U and238U were determined from a short post-irradiation of the fuel solution and counting of140Ba−140La and239Np, respectively. An iterative calculus method is presented which allows calculation of the irradiation history of the fuel solution from the above analyses. without any a priori knowledge.

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Determination of trace impurities in tin by neutron activation analysis

III. Simultaneous determination of 15 elements

Journal of Radioanalytical and Nuclear Chemistry
Authors:
W. Maenhaut
,
F. Adams
, and
J. Hoste

Abstract  

A method was developed for the determination of 15 trace elements in tin. High-purity tin samples (99.9999% and 99.999%) as well as tin of technical quality were analysed. Reactor neutron activation of the tin samples was followed by distillation of the matrix activities from a HBr−H2SO4 medium and Ge(Li) gamma-ray spectrometry of the distillation residue. The sensitivity of the method is generally high. For the high-purity samples the detection limits vary from 0.02 ppb (scandium) to 200 ppb (iron) for irradiation of 1 g of tin for 1 week at a thermal flux of 5·1012n·cm−2. ·sec−1. To decontaminate the surface of the tin samples, pre- and post-irradiation etching procedures were applied. The efficiency of these etching techniques was studied.

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Abstract  

For the determination of very low concentrations of copper in tin, an analytical method involving reactor neutron activation was developed whereby the copper activity was separated from the tin matrix by extraction of the Cu(I) cuproin complex in n-amyl alcohol. A new decontamination technique was sought in order to remove the copper contamination present on the tin surface. Pre-irradiation removal of the tin surface combined with post-irradiation etching appeared to be the most efficient.

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Abstract  

Reactor neutron activation analysis of antimony, indium and cadmium in high-purity tin is interfered with by nuclear reactions on the tin matrix. For a number of interfering reactions the cross-sections were determined. The following results were obtained:122Sn(n,γ)123mSn:σth=0.145 barn, I=0.79 barn;122Sn(n,γ)113Sn:σth=0.52, I=25.4 barn;112Sn(n, 2n)111Sn:
\documentclass{aastex} \usepackage{amsbsy} \usepackage{amsfonts} \usepackage{amssymb} \usepackage{bm} \usepackage{mathrsfs} \usepackage{pifont} \usepackage{stmaryrd} \usepackage{textcomp} \usepackage{upgreek} \usepackage{portland,xspace} \usepackage{amsmath,amsxtra} \pagestyle{empty} \DeclareMathSizes{10}{9}{7}{6} \begin{document} $$\bar \sigma _F = 290$$ \end{document}
microbarn;118Sn(n, α)115Cd:
\documentclass{aastex} \usepackage{amsbsy} \usepackage{amsfonts} \usepackage{amssymb} \usepackage{bm} \usepackage{mathrsfs} \usepackage{pifont} \usepackage{stmaryrd} \usepackage{textcomp} \usepackage{upgreek} \usepackage{portland,xspace} \usepackage{amsmath,amsxtra} \pagestyle{empty} \DeclareMathSizes{10}{9}{7}{6} \begin{document} $$\bar \sigma _F = 0.252$$ \end{document}
microbarn; and114Sn(n, p)114m1In:
\documentclass{aastex} \usepackage{amsbsy} \usepackage{amsfonts} \usepackage{amssymb} \usepackage{bm} \usepackage{mathrsfs} \usepackage{pifont} \usepackage{stmaryrd} \usepackage{textcomp} \usepackage{upgreek} \usepackage{portland,xspace} \usepackage{amsmath,amsxtra} \pagestyle{empty} \DeclareMathSizes{10}{9}{7}{6} \begin{document} $$\bar \sigma _F = 42.3$$ \end{document}
microbarn.
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Abstract  

The average cross-section in a fission-type reactor spectrum was determined experimentally for the reactions:46Ti(n,p)46Sc,47Ti(n,p)47Sc,48Ti(n,p)48Ti(n,α)45Ca and50Ti(n,α)47Ca. In order to obtain the (n,p) cross-sections, reactor irradiation of titanium was followed by measurement of the induced scandium activities with a Ge(Li) detector of calibrated detection efficiency. For this no chemical separations had to be carried out. For the (n,α) reactions, however, the induced calcium activities were separeted and purified by oxalate precipitation, after the bulk of the radioactivity had been removed by precipitation of titanium hydroxide. The47Ca disintegration rate was determined in the same way as for the scandium isotopes, whereas for45Ca liquid scintillation counting was carried out. The shape of the reactor spectrum was investigated by irradiating reference threshold detectors with different effective threshold energies. To correct for (n,γ) interferences, irradiations were carried out with and without cadmium shielding. On the basis of
\documentclass{aastex} \usepackage{amsbsy} \usepackage{amsfonts} \usepackage{amssymb} \usepackage{bm} \usepackage{mathrsfs} \usepackage{pifont} \usepackage{stmaryrd} \usepackage{textcomp} \usepackage{upgreek} \usepackage{portland,xspace} \usepackage{amsmath,amsxtra} \pagestyle{empty} \DeclareMathSizes{10}{9}{7}{6} \begin{document} $$\bar \sigma _F = 0.64$$ \end{document}
mb for the reaction27Al(n,α)24Na, the average cross-sections were as follows:46Ti(n,p)46Sc:10.5±0.4 mb;47Ti(n,p)47Sc: 16.3±0.6 mb;48Ti(n,p)48Sc:0.272±0.005 mb;48Ti(n,α)45Ca: 34μb;50Ti(n,α)47Ca: 8.1±0.3 μb.
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Abstract  

A comparison has been made between flux density distributions from massive and ring-shaped cylindrical isotopic neutron sources. A considerable gain in direct fast neutron flux is obtained for the latter geometry as well as a neat separation of fast and thermal flux density maxima along the axis of the source. Applications of these favourable properties are discussed.

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Abstract  

A Compton suppression gamma-ray spectrometry system has been evaluated for its use as a low level radioactivity counting facility. The system consists of a premium quality Ge(Li) detector surrounded by a shield of NaI(T1) detector material. Compton suppression is obtained by operating the two detector systems in anti-coincidence. Spectrum collection hardware consists of a NP11-A (DEC) interface for two ADC's and a PDP 11/T 10 minicomputer with 64 K byte core memory. Software development and system operation modes are described. System performance as a function of physical characteristics of the sample, scattering angle and gamma-photon energy, is discussed. Continum reduction for cascaded gamma-transitions and natural backgorund are considered separately.

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Abstract  

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.

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Abstract  

The eight segments of five normal human livers are analysed for 25 trace elements by radiochemical NAA. This consits of an automated wet destruction of the samples and two distillations, followed by ion exchange procedures. Ru is used as triple-comparator for the standardisation. Short-lived and matrix-isotopes are standardised by the Bowen's kale powder. The results reveal that the coefficient of variation within the liver is smaller than 10% for the elements Cd, Cl, Cs, Cu, Fe, K, Mg, Mn, Rb, Se and Zn. The highest range observed for the elements As, Br, Co, Cr, Hg, La, Mo, Na and Sb within a liver is smaller than the range observed between the five livers.

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