The separation of cesium by use of the inorganic ion exchanger ammonium molybdatophosphate from nitric acid solutions of intermediate level waste (ILW) from reprocessing of spent fuel elements according to the PUREX PROCESS has been demonstrated.Other inorganic exchange materials have shown high sorption values only for certain pH ranges: -ammonium hexacyano cobaltous ferrate (pH 12, 35 g Cs/kg) -potassium hexacyano nickel ferrate (pH 10, 30 g Cs/kg) -zirconium phosphate (pH 7, 100 g Cs/kg) -titanium phosphate (pH 7, 15 g Cs/kg) -antimony pentoxide (pH 2, 30 g Cs/kg) -titanium oxide (pH 7, 1 g Cs/kg) Except for high salt loading of 3.6 AT NaN0 3 , a significant loss of capacities usually occurs; this does not allow the use of these exchangers. However, ammonium molybdatophosphate shows excellent performance with high salt loadings and in a broad pHrange from pH 9 to conc. HN0 3 with a breakthrough-capacity of 60 g Cs/kg.

Production of fission Mo-99 ¡Reprocessing of nuclear fuel /Target fabrication Abstract 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 isotopie composition or enriched in molybdenum-98. In these cases the Mo-99 is generated via the nuclear reaction "Mo (n,y) "Mo.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 235 U (n, 0" Mo.This technique provides a product with a specific activity several orders of magnitude higher than that obtained from the " Mo (η, γ)" Mo nuclear reaction and perhaps even more important up to several thousands curies of Mo-99 per production run.

No abstract

AbstiactTreatment of medium activity waste (MAW) solutions generally includes a concentration step followed by a chemical treatment to remove radioactive materials. The latter can be returned in an appropriate manner to the process or stabilized in a suitable matrix. Presently, mostly precipitation methods are considered for the decontamination of MAW concentrates. However, more recent developments suggest that specific decontamination steps can be successfully carried out in various process streams of well defined composition by using Chromatographie techniques, especially ion-exchange systems. Such a process stream originates, e. g., from the basic sodium carbonate Solution which is used for the removal of organic degradation products from the organic extraction medium before the latter is returned to the process. Düring this procedure actinides, such as uranium, neptunium, and plutonium, which were previously bound in the organic phase, are reextracted in form of their water-soluble carbonato-complexes. In order to develop a technique to separate these actinides from the aqueous Solution in a chemical form in which they can be returned to the process, several ion-exchange systems were tested. It was found that the uranium retention reached a maximum of about 300 g U/kg ion-exchange resin in the case of Bio-Rex-5 in a stoichiometric Solution of uranyl and carbonate ions. The presence of organic degradation products did not affect the retention; however, a large excess of carbonate ions tends to reduce the amount of retained uranium. A further advantage of this ion-exchange system is that the uranium can be -quantitatively eluted with approximately three column volumes of elution Solution (NH4NO3). Similar results were obtained in experiments with neptunium and plutonium ions. Other MAW streams are the nitric acid solutions from the first and following solvent extraction cycles, from which the actinides were separated by using a strong base anionic exchanger, and the Pu-oxalatecontaining Solution from the nuclear fuel fabrication, from which Pu was removed via fixation preferentially on ion exchanger with phosphonic acid groups.

No abstract