At I.R.I. a new, fully automated facility for short half life INAA is being developed and installed at the Institutes 2 MW reactor. The fast rabbit transfer system is constructed only of plastic and carbonfiber parts, so that rabbit contamination is minimized. This system is automated in such a way that it can operate safely without direct supervisionn; the sequence of irradiations and measurements is optimized by a computer program for a given set of samples and analysis procedures. The rabbit system is controlled by an Apple IIe-computer connected to the central PDP 11/44 system of the Radiochemistry department. For a given set of samples and required analysis procedures (irradiation-, decay-, and measurement times) the central computer calculates an optimal sequence of individual actions (transfer from and to the reactor, sample storage or detector) to be carried out by the system. This sequence is loaded into the Apple-computer as a series of commands together with timing information. Actual control of the procedure occurs through the peripheral computer, which makes the system independent of delays or break-downs of the central multi-user computer system. Hardware, software and operating characteristics of the fast rabbit system will be discussed.
Practical application of oow energy gamma rays and X-rays in I.N.A.A. was restricted because of the complexity of the X-ray
spectrum and sample self-absorption. This paper describes a method for the calculation of sample self-absorption on the basis
of the actual sample spectra only, as measured with a high resolution semiconductor X-ray detector. In the 20–400 keV energy
range, the attenuation coefficient can be represented by a three parameter function of photon energy. This was verified by
measuring the transmission of photons of different energies through a range of materials. Experiments with neutron irradiated
U.S.G.S. standard reference materials with known major oxide composition showed that self-absorption thus calculated from
the observed spectra is in good agreement with the results of theoretical calculations based on known attenuation coefficients.
The system for routine instrumental neutron activation analysis, in use for several years at the IRI at Delft, has been evaluated.
Basis of this evaluation are: quality of the results, costs per analysis, capacity and ease of operation. A comprehensive
description of the analysis system and associated hardware and software is included.
In order to obtain reliable data about short-lived isotopes for use in thermal neutron activation analysis, experiments have
been carried out using a fast rabbit transfer system. Half-lives of 28 short-living isotopes have been measured by using a
counting system with a fixed dead-time. A Ge(Li) spectrometry system was used to determine the most important γ-ray energies
and intensities of these isotopes. For the half-lives an accuracy of better than 1% was attained, while for the γ-ray energies
the accuracy was 0.1 keV.
An extractive scintillating (ES) resin was evaluated for its performance as an on-line monitor of uranium in water. The TRU-ES resin is comprized of an inert macroporous polystyrene core impregnated with the organic fluors [diphenyloxazole (PPO) and 1,4-bis-(4-methyl-5-phenyl-2-oxazolyl)benzene (DM-POPOP)) and the selective extract (octyl(phenyl)-N, N-diisobutyl-carbamoylmethylphosphine oxide (CMPO) in tri-butyl phosphate (TBP)]. The TRU-ES resin, packed into translucent FEP Teflon tubing, was placed into a flow-cell scintillation detection system. Acidified aqueous solutions, 233U spiked synthetic ground water and EPA natural uranium QA samples, were pumped through the flow-cell while gross count rate and pulse-height spectra were collected. The increase in count rate is attributed to the uranium being extracted from the aqueous medium and retained by the TRU-ES resin with simultaneous detection of the resultant scintillation photons. The TRU-ES loading efficiency was nearly quantitative out of 2M HNO3 with a flow rate of 0.5 ml . min-1. The detection efficiency was measured to be 51& for 233U. The detection limit was determined to be ~2 Bq . l-1 for 233U based on a resin free column volume of 0.2 ml, and 50 ml of 2M HNO3 acidified groundwater.
Neutron activation analysis is attractive for trace-element determinations in large samples. Facilities for reactor irradiation and -ray spectrometry of kilogram-size cylindrical samples are described. The thermal neutron flux is ca. 5·1012m–2·s–1 with ath/epi>104, so neutron self-thermalization can be neglected. The correction for the neutron attenuation within the sample is derived from measurement of the neutron flux depression just outside the sample. Correction for -attenuation in the sample is performed via linear attenuation coefficients derived via transmission measurements. Also the natural radioactivity in the sample is taken into account. Examples are given of materials to which large sample INAA has been applied successfully, and further lines of development and exploration are indicated.
Two examples are given to illustrate how modern experimental techniques may extend the scope of possibilities of radiotracer applications. The first example refers to the use of a Ge-detector -ray spectrometer to measure the transport in plants of 15 elements simultaneously. The second example presented is an in-vivo study of the binding of Cd-ions in plants using meansurements of perturbed --directional correlations.
Uranium mining activities at Poços de Caldas plateau, Brazil, generated huge amounts of sulfidic waste rocks which were dumped into piles around the mine pit, requiring control measures to minimize the environmental impact. In this work, the sampling and the homogenization of these waste rocks are studied and their elemental analysis by instrumental neutron activation analysis (INAA) using normal (about 200 mg) and large (about 2 kg) samples. The results obtained allowed us to conclude that the waste rocks are extremely heterogeneous requiring even larger quantities of sample for its characterization.
Principles of the expression of uncertainty in measurements are briefly reviewed and special aspects of the uncertainty quantification in NAA are discussed in detail regarding the relative and k0-standardization in both modes of the technique, i.e., INAA and RNAA. A survey of uncertainty sources is presented and calculation of the combined uncertainty is demonstrated by an example of manganese determination in biological material by RNAA.
An alternative convention for use in the k0-method describing the (n, )-reaction rate upon reactor neutron irradiation has been derived by dividing the cross-section in a (v)=
0v0/v part and a pure resonance integral, instead of splitting up the neutron spectrum. It describes the (n, )-reactions with the Westcott factor g(T)1 but without resonances below 0.35 eV, and should yield better results for those with resonances below this limit. The resulting formulas are simpler than the ones currently used. An important practical aspect of this new convention is that no irradiations under Cd-cover are needed to determine the parameters to be used in the k0-method. The parameters determined previously for (n, )-reactions with g(T)=1 can still be used.