Purex process fission products

Some aluminosilicate minerals appear to meet the above criteria rather well, especially with regard to low leachability and chemical and physical stability 1, 2). A low-temperature process for converting the wastes to aluminosilicates with low leachability has now been found (3). Aqueous waste solutions containing NaOH, NaNOa, NaN02, NaAl02, mixed fission products, and minor amounts of other salts are mixed with powdered clays (kaolin, bentonite, halloysite, or dickite) and allowed to react at 30°-100°C to form small crystals of the mineral cancrinite. The sodium aluminosilicate crystal lattice of cancrinite contains large amounts of trapped salts and radioactive fission products. The process is applicable to caustic radioactive liquids such as neutralized Purex wastes or to salts or oxides produced by evaporation or calcination of these liquid wastes. 

Computer simulation programs for process design optimization have been developed for the PUREX process utilizing these relationships (22). A subroutine has also been developed which describes the behavior of fission products (23). 

These variations permit the separation of other components, if desired. Additional data on uranium, plutonium, and nitric acid distribution coefficients as a function of TBP concentration, solvent saturation, and salting strength are available (24,25). Algorithms have also been developed for the prediction of fission product distributions in the PUREX process (23). 

Then the fuel elements are dissolved in 7m HNO3 to give a solution containing U and Pu which, in the widely used plutonium-uranium-reduction, or Purex process, are extracted into 20% tributyl phosphate (TBP) in kerosene leaving most of the fission products... 

Purex [Plutonium and uranium recovery by extraction] A process for the solvent extraction of plutonium from solutions of uranium and fission products, obtained by dissolving spent nuclear fuel elements in nitric acid. The solvent is tri-n-butyl phosphate (TBP) in... 

The solvent extraction process that uses TBP solutions to recover plutonium and uranium from irradiated nuclear fuels is called Purex (plutonium uranium extraction). The Purex process provides recovery of more than 99% of both uranium and plutonium with excellent decontamination of both elements from fission products. The Purex process is used worldwide to reprocess spent reactor fuel. During the last several decades, many variations of the Purex process have been developed and demonstrated on a plant scale. 

A primary goal of chemical separation processes in the nuclear industry is to recover actinide isotopes contained in mixtures of fission products. To separate the actinide cations, advantage can be taken of their general chemical properties [18]. The different oxidation states of the actinide ions lead to ions of charges from +1 (e.g., NpOj) to +4 (e.g., Pu" " ) (see Fig. 12.1), which allows the design of processes based on oxidation reduction reactions. In the Purex process, for example, uranium is separated from plutonium by reducing extractable Pu(IV) to nonextractable Pu(III). Under these conditions, U(VI) (as U02 ) and also U(IV) (as if present, remain in the... 

Figure 12.5 illustrates the basic components of the Purex process three purification cycles for both uranium and plutonium are shown. High levels of beta and gamma radioactivity are present only in the first cycle, in which 99.9% of the fission products are separated. The other two cycles, based upon the same chemical reactions as the first cycle, obtain additional decontamination and overall purity of the uranium and plutonium products. 

Distribution ratios of the actinides in largest concentration and valency state, and some important fission products, which form the basis for the design of the Purex process flow sheet, are shown in Table 12.9. 

Uranium stripping Dilute HNO3 solutions at 45-50°C are used to remove uranium from the TBP phase. Traces of the fission products ruthenium and zirconium are eliminated in the second and third cycles of the Purex process. Also, in the second and third cycles, neptunium and the last traces of plutonium are removed from the uranium product. 

Np, and fission products. The Thorex solvent extraction process is generally used to reprocess spent Th-based fuels. As in the Purex process, the solvent is TBP diluted in an appropriate mixture of aliphatic hydrocarbons. Figure 12.9 shows the Thorex process flow sheet used by Kuchler et al. [41] for reprocessing high-burn-up thorium fuel. 

The Purex process, ie, plutonium uranium reduction extraction, employs an organic phase consisting of 30 wt % TBP dissolved in a kerosene-type diluent. Purification and separation of U and Pu is achieved because of the extractability of U02+2 and Pu(IV) nitrates by TBP and the relative inextractability of Pu(III) and most fission product nitrates. Plutonium nitrate and U02(N03)2 are extracted into the organic phase by the formation of compounds, eg, Pu(N03)4 -2TBP. The plutonium is reduced to Pu(III) by treatment with ferrous sulfamate, hydrazine, or hydroxylamine and is transferred to the aqueous phase U remains in the organic phase. Further purification is achieved by oxidation of Pu(III) to Pu(IV) and re-extraction with TBP. The plutonium is transferred to an aqueous product. Plutonium recovery from the Purex process is ca 99.9 wt % (128). Decontamination factors are 106 — 10s (97,126,129). A flow sheet of the Purex process is shown in Figure 7. 

In tlie PUREX process, the spent fuel and blanket materials are dissolved in nitric acid to form nitrates of plutonium and uranium. These are separated chemically from the other fission products, including the highly radioactive actinides, and then the two nitrates are separated into tv/o streams of partially purified plutonium and uranium. Additional processing will yield whatever purity of the two elements is desired. The process yields purified plutonium, purified uranium, and high-level wastes. See also Radioactive Wastes in the entry1 on Nuclear Power Technology. Because of the yield of purified plutonium, the PUREX process is most undesirable from a nuclear weapons proliferation standpoint,... 

The Purex process is used for almost all fuel reprocessing today. Irradiated UO2 fuel is dissolved in HNO3 with the uranium being oxidized to U02(N03)2 and the plutonium oxidized to Pu(NC>3)4. A solution of TBP in a high-boiling hydrocarbon, such as n-dodecane, is used to selectively extract the hexavalent U02(N03)2 and the tetravalent Pu(NC>3)4 from the other actinides and fission products in the aqueous phase. The overall reactions are... 

Table 2.5 summarizes the developments at the four major stages of the PUREX process. The success of the process is measured by the quantitative recovery (>99.9%) of U and Pu with high DFs (DF > 106) from the fission products and structural materials. There is also growing concern about the volumes of radioactive waste generated during fuel reprocessing. There have been continuous R D efforts in radiochemical laboratories toward these ends. 

In order to make use of thorium as a nuclear resource for power generation, development of efficient separation processes are necessary to recover 233U from irradiated thorium and fission products. The THORium uranium Extraction (THOREX) process has not been commercially used as much as the PUREX process due to lack of exploitation of thorium as an energy resource (157,180). Extensive work carried out at ORNL during the fifties and sixties led to the development of various versions of the THOREX process given in Table 2.6. The stable nature of thorium dioxide poses difficulties in its dissolution in nitric acid. A small amount of fluoride addition to nitric acid is required for the dissolution of more inert thorium (181). 

All partitioning processes described in today s literature that claim to separate the minor An(III) from the fission products by selective extraction of the An(III) actually perform the An(III)/Ln(III) partition from a feed arising from a front-end partitioning step, which has already separated the An(III) + Ln(III) fraction from a PUREX raffinate. There are, however, other processes that perform the An(III)/ Ln(III) partition by using a selective hydrophilic complexant, introduced either in the feed to selectively complex the An(III) and prevent their extraction, or in the stripping solution to selectively back-extract the An(III) from the loaded solvent. The development and achievements of these two families of processes ( selective extraction of An(III) and selective complexation of An(III) ) will be described hereafter. 

The ZEALEX Process Researchers from KRI have shown that the zirconium salt of dibutyl phosphoric acid (ZS-HDBP) was soluble in Isopar-L in the presence of 30% TBP. This super PUREX solvent, known as ZEALEX, extracts actinides (Np-Am) together with lanthanides and other fission products, such as Ba, Cs, Fe, Mo, and Sr from nitric acid solutions. The extraction yields depend on both the molar ratio between Zr and HDBP in the 30% TBP/Isopar-L mixture and the concentration of HN03 (232). Trivalent transplutonium and lanthanide elements can be stripped together from the loaded ZEALEX solvent by a complexing solution, mixing ammonium carbonate, (NH4)2C03, and ethylenediamine-N.N.N. N -tetraacetic acid (EDTA). An optimized version of the process should allow the separation of... 

FIGURE 3.24 SETFICS process flowsheet tested by JNC on an inactive surrogate PUREX raffinate. (Courtesy of Hirano, H., Koma, K., Koyama, T., 7th Information Exchange Meeting on Actinide and Fission Product Partitioning and Transmutation, October 2002, Jeju, Republic of Korea.)... 

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