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It is proposed that a database on the characteristics of research reactor fuel be constructed. This database will permit determining the unknown fuel and the unknown parts of the core. Arduous and precision methods are not needed to identify the fuel parts of pulsed reactors. Inspection and measurements of the geometric dimensions are sufficient in most cases. If a fuel part of a reactor is destroyed and has been converted to scrap, then the enrichment of the fuel, the isotopic composition of the uranium, and the impurity composition must be measured, although even this does not always solve the problem. Consequently, the selection and validation of the identifying indications must be the subject of special investigations.In the last few years the law-enforcement organs in the states formed after the collapse of the USSR and in the European Union have encountered more frequent cases of illegal circulation of nuclear materials. About 500 cases of smuggling of nuclear and radioactive materials have been recorded during the last decade in IAEA's database [1]. Approximately half of the materials confiscated consists of fissile materials -plutonium and enriched uranium, low-enrichment uranium (one-third of the cases), and initial materials (thorium and natural uranium). To fight the illegal circulation of nuclear materials and to obtain the required proofs for detaining the violator, the unknown fuel must be identified in order to determine where it was produced and stolen and what its intended use is.In the identification process, many parameters must be determined and compared with accessible information on the characteristics of the fuel parts of reactors. For this reason, an information database containing design, technological, and materials science characteristics of domestic and foreign nuclear reactors could be a useful tool for research and design purposes and for law-enforcement applications. Rosatom and the Joint Research Center of the European Commission (JRC), specifically, the A. A. Bochvar

The experimental results for the validation of a technology for producing 89 Sr in a reactor with solution fuel are presented. The aggregate state of the fuel -a water solution of uranyl sulfate UO 2 SO 4 -opens up a unique possibility of acting not only on the target radionuclide but also on its genetic precursors, produced as a result of nuclear transformations of the fission products as a whole in the decay chain 89 Se→ 89 Br→ 89 Kr→ 89 Rb→ 89 Sr. The half-life of 89 Kr (T 1/2 = 190 sec) -a gaseous precursor of 89 Sr -is sufficient for it to leave the fuel solution and migrate into the gaseous medium. A series of experiments is performed on obtaining gas samples from the free volume of the solution of the 20 kW Argus minireactor. It is shown that the mechanism of delivery of 89 Sr into the sorption volume of the experimental loop of the Argus reactor system is based on transport along the 89 Kr loop. Measurements of the content of the radioactive impurities in a solution of 89 SrCl 2 showed that the purification of the gas flow by metal-ceramic filters and purification of the solution on DOWEX-50×8 or crown ehter Sr-Resin yields medical quality 89 Sr. The productivity of the new technology, which is many-fold greater than that of the modern industrial methods for producing 89 Sr, is assessed on the basis of experimental data. 89 Sr (T 1/2 = 50.5 days, E β = 1463 keV) is used in oncology for treating a painful syndrome accompanying bone metastases. As a rule, metastization in bones is of a multiplistic nature, which makes it impossible to perform surgery or remote-controlled radiation therapy. In such cases, the most effective method is using compounds which are tropic to metastic tissue and contain radionuclides with high-energy β radiation, which can act on the tumor tissue and nerve endings. The travel distance of 89 Sr β particles in bone tissue does not exceed 7 mm, which localizes the radiation effect in a small region of the skeleton, lowering the dose load on the bone marrow and neighboring sections of the soft tissue [1]. 89 Sr in a pharmaceutical form was first used in 1942 [2]. A radiopharmaceutical preparation Metastron, manufactured by the firm Nikomed Amersham (Great Britain), appeared on the market in the 1980s [3]. In 1999, the production of the preparation "solution of strontium-89 chloride, isotonic for injections" was organized in our country. The recommended single dose of 89 Sr is 150 MBq (4 mCi).Methods for Obtaining 89 Sr in a Nuclear Reactor. At the present time, 89 Sr is obtained for medical purposes using the reactions 88 Sr(n, γ) 89 Sr and 89 Y(n, p) 89 Sr, irradiating a target consisting of highly-enriched strontium ( 88 Sr > 99.9%) in

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