Technetium separation from fission products

Chromatographic methods for the separation of technetium from fission products are based on the strong sorption of pertechnetate from weakly acidic, neutral, and... 

Technetium is an artificial element, available as Tc (a P-particle emitter, fi = 2.13 x 10 yr) which is isolated from fission product wastes by oxidation to [Tc04]. Separation... 

Progress in technetium chemistry obviously depends on sensitive analytical methods to detect this radioelemcnt, to efficiently separate it, and to determine technetium accurately. Several reviews on the analytical chemistry of technetium have been published [1-7]. Even the discovery of technetium was conclusive only because separation techniques known at the time for the homologous element rhenium were used 8. Furthermore, the detection of naturally occurring technetium or the determination of the element in the environment presupposed the application of highly sensitive methods. The effective extraction of pcrtechnctate into organic solvents for the isolation of technetium from fission product waste solutions is. in addition, an example of the significance of appropriate techniques in analytical chemistry. 

Separation of Technetium from Uranium Fission Products.114... 

Subsequently, solvent extraction was applied to recover the fission product technetium from the residue remaining after the fluorination of irradiated uranium fuel elements . The residue was leached with concentrated aluminum nitrate solution, which was extracted by 0.3 M trilaurylamine in a hydrocarbon diluent. After separation of uranium, neptunium, and aluminum nitrate, technetium was back extracted into a 4 N sodium hydroxide solution. 

Campbell has studied the separation of technetium by extraction with tributyl phosphate from a mixture of fission products cooled for 200 days. Nearly complete separation of pertechnetate is achieved by extraction from 2 N sulfuric acid using a 45 % solution of tributyl phosphate in kerosene. Ruthenium interferes with the separation and is difficult to remove without loss of technetium other radioisotopes can be removed by a cation-exchange process. However, this separation procedure has not been widely applied because of the adverse influence of nitrate. 

Goishi and Libby have investigated the extraction of pertechnetate from alkali solutions with pyridine. Later work showed that a better extraction is obtained using a mixture of sodium hydroxide and sodium carbonate as the aqueous phase. Since the uranyl carbonate complex is not extracted into pyridine, this system may be used for the separation of technetium from uranium. Distribution coefficients of fission products in pyridine are given in Table 4. Substituted pyridine such as 2,4-dimethylpyridine or 4-(5-nonyl)pyridine ) are useful for separating technetium from solutions containing appreciable amounts of aluminum nitrate. 

The solvent extraction of pertechnetate with cyclohexanone has proved to be an efficient and selective method which can be applied to the separation of Tc from long-lived fission products in the bum-up analysis . The recovery of technetium from the fission products is about 93 %. 

Tucker et al. have separated " Tc from the fission product Mo using chromatographic aluminum oxide washed by dilute nitric acid at pH 1.5. Mo dissolved in the same dilute HNO3 of pH 1.5 is passed through the column which is then eluted with 0.1 M HNO3. Tc is only slightly adsorbed and can easily be eluted while molybdenum is retained on the column. The purity of technetium is 99.99%. 

A method has been developed for the determination of technetium-99 in mixed fission products by neutron activation analysis Tc is separated from most fission products by a cyclohexanone extraction from carbonate solution, the stripping into water by addition of CCI4 to the cylohexanone phase, and the adsorption on an anion exchange column. Induced Tc radioactivity is determined using X-ray spectrometry to measure the 540 and 591 keV lines. The sensitivity of the analysis under these conditions is approximately 5 ng. The method has been successfully applied to reactor fuel solutions. 

The yield of primary fission products in relation to their mass number is shown in Fig. 8. Most of the radioactive fission products have short half-lives, but some decay slowly technetium-99 has 2.2 x 10 yr and promethium-147, 2.6 yr. Two of the most useful fission products obtained in the operation of a pile are caesium-137 (/ -, 33 yr) and strontium-90 ( +, 22 yr). The former is used in medicine and the latter in medicine and industry both are separated from shorter-lived material. 

From the above discussion it follows that tetravalent and hexavalent thorium, uranium, and plutonium can be separated from the trivalent rare-earth fission products by taking advantage of differences in complexing properties. More highly charged cation fission products, such as tetravalent cerium and the fifth-period transition elements zirconium, niobium, molybdenum, technetium, and ruthenium, complex more easily than the trivalent rare-earths and are more difficult to separate from uranium and plutonium by processes involving complex formation. 

For both technetium and iodine, capture of a single neutron followed by P decay results in the production of stable isotopes of ruthenium and xenon, respectively. Further, as the irradiation proceeds, additional neutrons can be captured but these products are also stable. In practice, the transmutation of technetium and iodine in conventional thermal reactors will be difficult for several reasons. First, because these are major fission products, they are produced in high yield and thus result in large inventories that must be separated fi-om the HLW and transmuted. Second, the thermal neutron capture cross sections are not particularly large and therefore much higher neutron fluxes are required to significantly reduce the inventory in a reasonable amount of time. Third, the available neutron flux from conventional LWRs is too small (lO neutrons cm s ) to achieve the required level of transmutation in a reasonable length of time. 

Each of these elements may be used for production of nuclear fuel or other purposes. The recovery efficiency for uranium is reported as 99.87% and for plutonium 99.36%-99.51% (NEA 2012). The extended PUREX includes separation of neptunium and technetium as well as recovery of americium and curium that are also separated from each other by additional extraction stages as given in detail in the flowsheet (NEA 2012). The advanced UREX-i-3 process generates six streams after separation uranium for re-enrichment Pu-U-Np for mixed oxide fuel c for managed disposal Am-Cm to be used as burnable poisons and for transmutation high-heat-generating products (Cs and Sr) and a composite vitrified waste with all other fission products. Some fuel types may require preliminary steps like grinding to enable their dissolution. 

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