The Determination Of Uranium By Neutron Irradiation And Delayed Neutron Counting

The counting of delayed neutrons emitted by fission is presented as a valuable technique for the measurement of uranium concentrations in a variety of matrices. The concentrations successfully analyzed can vary from well under one part per million to the high concentrations found in uranium ores. In a sample analysis, fission is induced in the uranium in the sample by irradiation in a thermal neutron flux, then the sample is rapidly transferred to a counting assembly capable of detecting delayed neutrons. In the system described, irradiation is performed in a TRIGA reactor, and counting is done in a paraffin-moderated assembly of BF3 gas-filled detectors. All equipment needed for the analysis and the necessary procedures are discussed. The delayed fission neutron (DFN) counting technique is compared to other methods of analysis for uranium, and the experiences of other researchers using the DFN technique are summarized. When compared to many other methods, DFN counting is relatively free of interferences. The interferences which may occur, such as high energy gammas, unknown neutron-emitting nuclides or strong neutron absorbers in the sample, are discussed. Any uncertainties associated with a DFN measurement are also analyzed. DFN counting has been used in many applications, such as the measurement of uranium in geological samples, phosphate products and seawater adsorbers. It can also be used for the measurement of thorium in many samples. These applications are presented, and the results of many different analyses are listed. Experience gained at Oregon State University is examined in detail, and several improvements are suggested.

The method of neutron activation analysis of U by delayed-neutron counting was investigated in order to ascertain if the method would be suitable for routine application to such analyses. It was shown that the method can be used extensively and routinely for the determination of U. Emphasis was placed on the determination of U in the types of sample materials encountered in nuclear technology. Determinations of U were made on such materials as ores, granite, sea sediments, biological tissue, graphite, and metal alloys. The method is based upon the fact that delayed neutrons are emitted from fission products from the interaction of neutrons with U/sup 235/. Since the U/sup 235/ component of U undergoes most of the fissions when a sample is in a neutron flux, the method is predominately one for the determination of U/sup 235/. The total U in a sample or the isotopic composition of the U in a sample can be determined provided there is a prior knowledge of one of these quantities. The U/sup 235/ content of a test sample is obtained by comparing its delayed-neutron count to that obtained with a comparator sample containing a known quantity of U/sup 235/. (auth).