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Actinides from irradiated nuclear fuels, separation

Actinide Separations from Irradiated Nuclear Fuels and/or Waste Solutions... [Pg.240]

Separation of Actinides from the Samples of Irradiated Nuclear Fuels. For the purpose of chemical measurements of burnup and other parameters such as accumulation of transuranium nuclides in irradiated nuclear fuels, an ion-exchange method has been developed to separate systematically the transuranium elements and some fission products selected for burnup monitors (16) Anion exchange was used in hydrochloric acid media to separate the groups of uranium, of neptunium and plutonium, and of the transplutonium elements. Then, cation and anion exchange are combined and applied to each of those groups for further separation and purification. Uranium, neptunium, plutonium, americium and curium can be separated quantitatively and systematically from a spent fuel specimen, as well as cesium and neodymium fission products. [Pg.325]

Uranyl nitrate has an unusual property, shared only by nitrates of a few other actinides, of being very soluble in a number of organic solvents. When such an organic solvent is immiscible with water, it can be used in a solvent extraction process to extract uranium from aqueous solutions and separate it from associated impurities. Such applications of solvent extraction are very important in extracting and purifying uranium from leach solution of uranium ores or from nitric acid solution of irradiated nuclear fuel. Examples of extractants that have been used for such separation processes are listed in Table 5.14. [Pg.230]

The reprocessing involves separating the fission products from the actinides, and then separating the plntoninm from the uranium. The best known procedure of this type is the PUREX (Plutonium, URanium Extraction) process that is used for recovery of uranium and plutonium from irradiated fuel (see details in Chapter 2). The separated plutonium can be used for the production of nuclear weapons or converted into the oxide form, mixed with nraninm oxide and can be used as MOX nuclear fuel. [Pg.37]

The operations and facilities include ore exploration (not included in NFCIS list), mining, ore processing, uranium recovery, chemical conversion to UO2, UO3, UF4, UFg, and uranium metal, isotope enrichment, reconversion of UF to UO2 (after enrichment), and fuel fabrication and assembly that are all part of the front end of the NFC. The central part of the NFC is the production of electric power in the nuclear reactor (fuel irradiation). The back end of the NFC includes facilities to deal with the spent nuclear fuel (SNF) after irradiation in a reactor and the disposal of the spent fuel (SF). The spent fuel first has to be stored for some period to allow decay of the short-lived fission products and activation products and then disposed at waste management facilities without, or after, reprocessing to separate the fission products from the useful actinides (uranium and plutonium). Note the relatively large number of facilities in Table 2.1 dedicated to dealing with the spent fuel. Also listed in Table 2.1 are related industrial activities that do not involve uranium, like heavy water (D2O) production, zirconium alloy manufacturing, and fabrication of fuel assembly components. [Pg.63]

In principle, two fundamentally different methods can be applied to solve this task. The first one is determination of the residual concentrations of the fissile nuclides after irradiation and calculation of the burnup from the difference between final and initial values. For this purpose, the uranium and plutonium fraction has to be separated from the fission and activation products and from each other (e. g. by extraction chromatography) subsequently, the concentrations of the individual isotopes, in particular of the fissile isotopes, are analyzed by mass spectrometry. Well-established analytical techniques for performing such analyses are available, so that only small error margins are to be expected in the determination of the concentrations of the isotopes under consideration. However, there are two problems that can potentially cause systematic errors. The first one is the well-known question of the accuracy of results which have been obtained as a difference between two numbers, which limits the accuracy at lower burnup values in particular. The second problem is that the fissile nuclides are not only consumed by nuclear fission but by neutron capture as well in order to avoid systematic errors here, the capture-to-fission ratio valid for the particular irradiation conditions has to be taken into account in the calculation of depletion during irradiation. If one recalls the complicated buildup and decay mechanisms of actinide nuclides during reactor irradiation (see Fig. 3.5.), it is obvious that such correction requires complex calculations. On the other hand, the direct determination of the residual concentration of fissile nuclides is not influenced by errors due to inaccuracies in the fission yields of fission products to be measured nor by migration-induced inho-mogenities in the fuel. [Pg.91]


See other pages where Actinides from irradiated nuclear fuels, separation is mentioned: [Pg.315]    [Pg.885]    [Pg.546]    [Pg.885]    [Pg.9]    [Pg.7030]    [Pg.240]    [Pg.19]    [Pg.954]    [Pg.460]    [Pg.204]    [Pg.883]    [Pg.11]    [Pg.102]    [Pg.212]   
See also in sourсe #XX -- [ Pg.337 , Pg.338 ]




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