Patent Number: 059109718
Section: summary

BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to methods and systems for separating isotopes from nuclear reactors, and in particular to a method of producing molybdenum-99 (Mo-99) used for medical purposes from the uranyl sulfate nuclear fuel of an aqueous homogeneous solution nuclear reactor. 2. Description of the Prior Art At the present time more than 50% of the world's annual production of radionuclides are used for medical purposes, such as for the early diagnoses of diseases and for therapy. A basic condition of the use of radionuclides in medicine is the requirement that the radiation exposure of a patient be minimal. This necessitates the use of short-lived radionuclides. A nuclide with a short half-life, however, creates difficulties in transportation and storage. The most used radionuclide for medical purposes is Mo-99 with a half-life of 66 hours. Mo-99 decay results in Tc-99m with a half-life of 6 hours and about 140 keV of gamma (.gamma.) energy convenient for detection. Currently, more than 70% of diagnostic examinations are performed using this radionuclide. One method of Mo-99 production involves using a target of natural molybdenum or molybdenum enriched in Mo-98 irradiated by a neutron flux in a nuclear reactor. Mo-99 results from a neutron radiation capture .sup.98 Mo(n,.gamma.).sup.99. The irradiated target with Mo-99 then undergoes radiochemical reprocessing. This method, however, has a low productivity and the Mo-99 produced is characterized by a low specific activity due to the presence of Mo-98 in the final product. Another method of Mo-99 production is based on uranium fission under neutron irradiation of a U-Al alloy or electroplated target in a nuclear reactor. The target contains 93% enriched uranium (U-235). After irradiation, the target is reprocessed by one of the traditional radiochemical methods to extract Mo-99 from the fission products. The specific activity achieved by this method is several tens of kilocuries per gram of molybdenum. A serious disadvantage of this method is the necessity of recovering large amounts of radioactive wastes that are byproducts of the fission process. These wastes exceed the Mo-99 material produced by two orders of magnitude. A 24-hour delay in processing the irradiated uranium targets results in a decrease of total activity by about an order of magnitude, during which time the Mo-99 activity decreases by only 22%. After two days, the activity of the waste byproducts exceeds that of the Mo-99 by a factor of six or seven. The problem of long-lived fission product management is the major disadvantage in the production of Mo-99 by this method. U.S. Pat. No. 5,596,611 discloses a small, dedicated uranyl nitrate homogeneous reactor for the production of Mo-99 in which the radioactive waste products are recirculated back into the reactor. A portion of the uranyl nitrate solution from the reactor is directly siphoned off and passed through columns of alumina to fix some of the fission products, including Mo-99, to the alumina. The Mo-99 and some fission products on the alumina column are then removed through elution with a hydroxide and the Mo-99 is either precipitated out of the resultant elutriant with alpha-benzoinoxime or passed through other columns. This uranyl nitrate reactor has the advantage of recycling the fission byproducts, but the conventional extraction method to obtain Mo-99 is relatively inefficient. It is an object of the present invention to produce Mo-99 directly from the uranyl sulfate solution of an aqueous-homogeneous solution nuclear reactor in a manner that minimizes the radioactive byproducts and most efficiently uses the reactor's uranium fuel. The process is relative simple, economical, and waste free. SUMMARY OF THE INVENTION In the present invention, Mo-99 is generated, along with other fission products, in a uranyl sulfate nuclear-fueled homogeneous-solution nuclear reactor. This reactor operates at powers of from 20 kW up to 100 kW for a period from of several hours to a week producing various fission products, including molybdenum-99. After shutdown and following a cool-down period, the resultant solution is pumped through a solid sorbent material that selectively absorbs the Mo-99. The uranyl sulfate and all fission products not adhering to the sorbent are returned to the reactor vessel, thus containing the fission byproducts and conserving the uranium.