[en] Currently much of the world's supply of ^99mTc for medical purposes is produced from ⁹⁹Mo derived from the fissioning of high enriched uranium (HEU). This paper presents the results of our continuing studies on the effects of substituting low enriched uranium (LEU) for HEU in targets for the production of fission product ⁹⁹Mo. Improvements in the electrodeposition of thin films of uranium metal continue to increase the appeal for the substitution of LEU metal for HEU oxide films in cylindrical targets. The process is effective for targets fabricated from stainless steel or zircaloy. Included is a cost estimate for setting up the necessary equipment to electrodeposit uranium metal on cylindrical targets. Further investigations on the effect of LEU substitution on processing of these targets are also reported. Substitution of uranium silicides for the uranium-aluminium alloy or uranium aluminide dispersed fuel used in current target designs will allow the substitution of LEU for HEU in these targets with equivalent ⁹⁹Mo-yield per target and no change in target geometries. However, this substitution will require modifications in current processing steps due to (1) the insolubility of uranium silicides in alkaline solutions and (2) the presence of significant quantities of silicate in solution. Results to date suggest that substitution of LEU for HEU can be achieved. (author). 30 refs, 4 figs, 3 tabs

[en] A production process for fission Mo-99 is proposed in which natural uranium as uranium oxide is used as target material. This procedure is particularly interesting for those which do not dispose of enriched nuclear fuel material. Approximately 400g of uranium oxide enclosed in irradiation cans are dissolved in nitric acid after irradiation for 100 hrs at a neutron flux of 5.10¹³ cm^-2 s^-1 in a research reactor. The separation of Mo-99 from the fuel-fission product solution is performed by ion exchange with alumina in a chromatographic column. Final purification includes the repeated chromatographic separation and subsequently a sublimation stage. By this process about 100 Ci of Mo-99 is obtained at the end of the production process. The quality (radiochemical and radioactive purity) corresponds to international standards to medically used Mo-99/Tc-99m generators. (author). 6 refs, 5 figs, 3 tabs

[en] The irradiation device is centered in a beryllium plug with a height roughly equal to that of the core. By using a target holder system, 3 targets can be placed in the irradiation site with a water circulation at a rate of approximately 7 m s^-1. The handling is completely done underwater as well as for loading and unloading into the shipping container. This paper is reproduced under the explicit authorization of the Director General of the Institut National des Radioelements. (author). 3 figs

[en] Generally two different techniques are available for molybdenum-99 production for use in medical technetium-99 generation. The first one is based on neutron irradiation of molybdenum targets of natural isotopic composition or enriched in molybdenum-98. In these cases the Mo-99 is generated via the nuclear reaction ⁹⁸Mo (n,γ) ⁹⁹Mo. Although this process can be carried out at low expenditure it gives a product of low specific activity and, hence, restricted applicability. In a second process Mo-99 is obtained as a result of the neutron induced fission of U-235 according to ²³⁵U (n,f) ⁹⁹Mo. This technique provides a product with a specific activity several orders of magnitude higher than that obtained from the ⁹⁸Mo (n,γ) ⁹⁹Mo nuclear reaction and perhaps even more important up to several thousands curies of Mo-99 per production run. In this paper a modern production procedure of Mo-99 via the fission reaction, which was developed at the Institute of Radiochemistry of the Nuclear Research Centre Karlsruhe will be described. The targeting, irradiation of U-235, the separation and purification steps involved as well as the recycling of the non-converted U-235, which should be a major consideration in any production technique, will be discussed. (author). 24 refs, 14 figs, 1 tab

[en] The paper describes the efforts of the Argentine Atomic Energy Commission to establish a programme to produce fission ⁹⁹Mo for the preparation of ^99mTc generators. The production plant has been completed in 1987 and has started limited production runs. The hot cells consist of four hot cells with a stainless steel lining. The chemical separation process of ⁹⁹Mo from the irradiated Al/Alloy (90% enriched) targets is similar to the process developed by A. Sameh in the Federal Republic of Germany. The product specification conforms very well with the requirements for a safe use in the preparation of ^99mTc generator for medical use. (author). 3 figs

[en] A fission ⁹⁹Mo plant was constructed in 1977 at the facilities of the Japan Atomic Energy Research Institute (JAERI). UO₂ pellets (2.6% enrichment) were utilized on target material containing 120g of UO₂. The pellets were irradiated in the JRR-2 and JRR-3 reactors up to 7 days at a neutron flux of 2 or 3 x 10¹³n/cm² x sec. About 20 Ci of ⁹⁹Mo per week was routinely produced for 39 weeks. However, due to severe constraints of a research establishment, commercial production runs had been stopped. This paper provides information on the target configuration, post-irradiation separation process and waste disposal problems. (author). 14 refs, 4 figs, 3 tabs

[en] The paper describes in general the production programmes of ⁹⁹Mo, ¹³¹I and ¹³³Xe at IRE by irradiating and reprocessing uranium targets. Given are technical data regarding the target technology, reactor and neutron flux requirements as well as the transport of irradiated targets. Different reactors are used for the irradiation of IRE's targets so that the different irradiation devices have been adapted to encounter the various conditions of the reactor characteristics and to obtain the desired activity by loading and unloading the targets independently of the reactor operation. This paper is reproduced under the explicit authorization of the Director General of the Institut National des Radioelements. (author). 6 refs, 6 figs

[en] The irradiation device occupies a position located in the peripheral area of the reactor core behind the beryllium reflector. This facility can contain 12 targets and consists of a coolant guide and four target holders. The targets are unloaded into a hot cell with a hermetic connection with the special shipping container. This paper is reproduced under the explicit authorization of the Director General of the Institut National des Radioelements. (author). 8 figs, 2 tabs

[en] This paper presents the results of preliminary studies on the effects of substituting low enriched uranium (LEU) for highly enriched uranium (HEU) in targets for the production of fission product ⁹⁹Mo. Issues that are addressed include (1) purity and yield of the ⁹⁹Mo/^99mTc product, (2) fabrication of LEU targets and related concerns, and (3) disposal of radioactive waste. Laboratory experimentation was part of the efforts for issues (1) and (2); thus far, radioactive waste disposal has only been addressed in a paper study. Although the reported results are still preliminary, there is reason to be optimistic about the feasibility of utilizing LEU targets for ⁹⁹Mo production. (author). 37 refs, 1 fig., 5 tabs

[en] The irradiation can take place in the PSF (Pool Side Facility) or in a special in-core facility called TRIO-FIT. In the PSF, two positions with high neutron beams allow to irradiate 6 targets whereas the TRIO-FIT accepts 9 targets in three parallel channels, simultaneously. This paper is reproduced under the explicit authorization of the Director General of the Institut National des Radioelements. (author). 4 refs, 2 figs