Determination of U235 burn-up in irradiated natural uranium by chemical and counting methods

Abstract

A method was desired for determining, with a precision of 1 per cent, the U 235 content of U samples taken lengthwise of a highly irradiated NRX natural U fuel rod. A mass spectrometric method capable of achieving this accuracy was not available at the time this work was begun. The method chosen was based on measurements of the fission and α-counting rates of natural U samples and U samples from the fuel rod after chemical purification, in order to compare their U 235 : U 238 ratios. Due to the thin sources necsssary for the counting, the time required for the direct determination of U 238 by α-counting was too long to be practical. An indirect measure of the U 238 content was, therefore, obtained by counting the Pu 239 produced by neutron irradiation of the sources in a position of high flux in NRX. To ensure a valid basis for comparison of samples, all irradiation and counting was done with a source of natural U and one of depleted U placed back to back. The sources were not disturbed on changing from fission- to α-counting. This type of counting was accomplished by use of an ion chamber with a double grid, located at the opening of a thermal-neutron beam hole. Fission counting was done with the beam hole shutter open; α-counting was done with the shutter closed, a higher gas pressure in the chamber, and a higher electronic gain. Because of difficulties caused by radiation damage to the sources, it did not appear possible to approach the precision desired for U 235 burn-up values, and a further complexity had to be introduced into the method. The new procedure consisted of the following steps: (1) A set of sources and a set of targets were prepared from natural U and each depleted U sample. The sources were designed for the determination of U 235 by fission counting; the targets, after processing, were used for the determination of U 238 via Pu 239 . Each set of sources and corresponding targets contained Pu 238 which was added as a tracer. For any given U sample, sources and targets were prepared at the same time and in a manner designed to maintain the Pu 238 : U 238 ratio constant for both sets. (2) Sources were fission- and α-counted in the gridded ion chamber to obtain U 235 : Pu 238 ratios. (3) Targets were irradiated in the back-to-back arrangement used previously. (4) After irradiation, the U and Pu were dissolved off the Al target backing and the Pu was carefully purified. (5) This Pu was used to prepare sources on Pt for α pulse height analysis which gave Pu 239 : Pu 238 ratios directly and thus, indirectly, U 238 : Pu 238 ratios. Prom these and the U 236 : Pu 238 ratios obtained in (2), U 235 : U 238 ratios were calculated. The use of an α-active tracer was necessary to link the two sets of counting results. The advantage of using Pu 238 as a tracer was that the method is independent of chemical yields during the processing of irradiated samples. The precision of the counting methods used for bum-up measurements varied from 1.3 to 2.6 per cent. Before the chemical and counting methods had been fully worked out, a satisfactory mass spectrometric method was developed and put into routine use. The precision of the mass spectrometer measurements, for samples which were analyzed by the counting techniques, varied from 0.5 to 1.4 per cent. Results obtained by analysis with the mass spectrometer and the counting methods agreed within the experimental errors listed above.