Independent verification of reprocessing input and process volumes. Progress report

Abstract

One of the most difficult challenges in the nuclear fuel cycle is to establish a material balance for tanks containing spent fuel solutions. These tanks present an extremely hostile environment to the analyst, making determination of the amount of fissile material in them difficult. Historic methods used to determine the volumes of solutions in these tanks are usually based on use of tank geometry in conjunction with depth and density measurements, both of which vary with temperature and are normally controlled by the operator of the facility. Changes in the interior geometry of tanks due to addition of various types of equipment, minor structural modifications, and accumulation of insoluble materials at the bottom and on the interior surfaces of the tanks affect the available volume and make difficult its accurate measurement. For safeguards purposes, it is thus desirable to have a repeatable method of volume determination independent of the operator and other factors. To determine the quantity of uranium and plutonium in a given tank, aliquots are withdrawn and subjected to analysis; both isotopic abundances and the amount of each element present are determined. Isotope dilution mass spectrometry is the method of choice for quantitative measurements in this application and yields values more precise and accurate than other methods. In the research that is the focus of this project, the application of isotope dilution mass spectrometry has been extended, through use of a double lutetium spike, to the determination of the volume (or weight) of the solution in tanks of any size and shape, regardless of the nature of the solutions within them. The goal of the work described in this report was to develop hot-cell compatible chemical separation procedures for lutetium and to try to devise a better method for quantitatively introducing natural lutetium to the tank