Of special interest from a practical point of view is the modification of the extraction mixture, i. e. the use of dipicolinic acid diamides (DPA) instead of carbamoylphosphineoxide. The mixture of CCD with DPDA shows the synergistic effect for Am extraction [37]. The proposed mixture of CCD + DPDA + polyethylene glycols manifests the extraction of Cs, Sr, REE and TPE, similar to the classical UNEX-extractant [38]. The formulae of some investigated DPDAs are presented in Figs 9.11 and 9.12.

The data on extraction of radionuclides are given in Table 9.18 and Fig. 9.13, from where it can be seen that diamides effectively extract radionu­clides even from 3 M HNO3.

The advantages of DPAs compared to carbamoylphosphineoxides are their rather simple synthesis and the possibility for a wide variation in their

9.12 N, N,-diphenyl-N, N,-dimethyldiamide of dipicolinic acid (PhMDPA).

1e+4 e		Я		л, и
1e+3i	О
					□
1e+2n	■A.
						в
1e+11		о				□
1e+0n	0.0 M TBDPA		о о
1e— 1	□ 0.03 M TBDPA 0.06 M TBDPA		о
1e-2n	О 0.09 M TBDPA			о	о
1e-3-	——— 1—- 1— 1…………… і——— 1—	—1—1—1-І	III,-	—- 1—	—,—	………..

0.01

0.1

1

[HNO3], M

9.13 Extraction of europium by mixture of 0.13 M CCD + 0.027 M PEG-400 + TBDPA in FS-13 as function of HNO3 concentration and content of diamide in organic phase.

Table 9.18 Extraction of Cs, Sr and Eu by 0.02 M CCD + 0.01 M diamide + 0.002M PEG-400 in F-3 as function of HNO3 concentration HNO3 Cs137 Sr85 Eu152 TBDPA 1 M 13 64 >100 3 M 2.8 6 60 6 M 0.6 1.3 1.1 PhMDPA 1 M 14 84 54 3 M 1.7 2.5 1.3 6 M 0.5 1.1 0.2

structure. Further, metal solvates with DPDA are more easily soluble when compared to those with CMPO, and thus one can use higher concentrations of DPDA in the extraction mixture which leads to an increase in the extrac­tion mixture capacity for trivalent metals. This is of importance for the treatment of solutions with a high content of REE. Furthermore, DPAs extract TPE to a greater extent than REE.

9.14 Formulas of nitrobenzotrifluoride (F-3), phenyltrifluoromethylsulfone (FS-13) and bis-tetrafluoropropyl ether of diethylene glycol (F-8).

As in other systems based on CCD, the extraction ability of CCD — DPDA and CCD — DPDA — PEG systems varies depending on the diluent in the row of F-3 > FS-13 > F-8 (see Fig. 9.14).

The above data indicate that diamides of dipicolinic acid are of interest as synergistic additions to CCD. The CCD — DPDA mixtures effectively extract REE and An from acidic solutions of HNO3. Diamides are easier to synthesize and thus they are significantly cheaper than carbamoylphos — phineoxides. At the same time, the extraction properties of CCD — DPDA and CCD — CMPO mixtures are practically the same. It should be also emphasized that, by applying DPDA instead of CMPO one can use higher concentrations of DPDA and thus increase the limiting concentration of REE and An in the extractant. This property is of prime importance for treatment of waste with a high REE content. The modified UNEX-extractant CCD-TBDPA — Slovafol-909 in F-3 was tested on simulated waste with high REE content. Cesium, strontium and minor actinides were recovered rather efficiently. At extraction from the simulated solution bearing more than 4 g/L REE, the high separation factors of americium remained even after three successive contacts with fresh portions of aqueous solutions. Dynamic testing of this extractant was performed at Mayak PA in collaboration with RI. The extraction process was stable and americium was extracted more effectively than europium. Hydrodynamic and extraction properties were unaffected during the tests.

Selective stripping is one possible direction for UNEX process develop­ment. The data on selective stripping of metals from both organic solvents are presented in Tables 9.19 and 9.20. It can be seen that, in the case of a saturated UNEX solvent (0.08 M CCD + 0.015 M PEG-400 + 0.013 M CMPO in FS-13), actinides and lanthanides can first be stripped with a solution (A) of ammonium carbonate and acetohydroxamic acid (AHA), and then Sr can be stripped separately from Cs with a solution of (NH4)2CO3 with AHA and DTPA (B). Cesium is stripped in the last stage using 2 M methylamine carbonate solution [39].

Table 9.19 Selective stripping of metals from UNEX solvent (0.08 M CCD + 0.015 M PEG-400 + 0.013 M CMPO in FS-13) Stripping solution D Cs Sr Eu Pu Np U A 1 M (NHACOa + 0.1 M AHA 4.4 2.3 0.005 0.02 0.005 0.02 B 1 M (NHACO3 + 0.1 + 0.05 M DTPA M AHA 4.1 0.002 0.002 0.01 0.001 0.06

Table 9.20 Stripping of metals from modified UNEX solvent (0.1 0.025 M PEG-400 + 0.05 M TBDPA in FS-13)	M CCD +
Aqueous phase	D
		Cs	Sr	Am	Eu
A	1 M (NHACOa + 0.13 M AHA	1.9	2.7	21	561
C	1 M (NH4)2CO3 + 0.078 M Citric acid	1.7	0.45	11	—
D	1 M (NH4hCO3 + 0.052 M NTA	2.0	0.46	1.2	—
E	1 M (NH4)2CO3 + 0.024 M HEDPA	2.1	1.1	2.9	—
F	1 M (NH4)2CO3 + 0.013 M DTPA	1.8	0.009	0.04	0.02
G	1 M (NH4)2CO3 + 0.039 M DTPA	2.0	0.008	0.04	0.0001
H	1 M (NH4hCO3 + 0.2 M Glycine + 0.013 M DTPA	1.8	0.031	1.6	—
I	1 M (NH4hCO3 + 0.2 M Glycine + 0.026 M DTPA	2.1	0.030	1.4	—

The same stripping solutions were examined for the recovery of metals of interest from the modified UNEX-solvent (0.1 M CCD + 0.05 M TBDPA + 0.025 M PEG in FS-13). Unlike in the previous case, the use of 1 M (NH4)2CO3 + 0.013 M AHA does not provide stripping of Ln and An ele­ments; therefore, different complexants were examined, such as citric acid, nitrilotriacetic acid (NTA), hydroxyethylene-diphosphonic acid (HEDPA), diethylene triamine pentaacetic acid (DTPA), mixed with ammonium car­bonate. The most promising data were obtained with aminoacetic acid (glycine) used as a buffer compound (solutions H and I). The principal scheme for selective stripping is indicated in Fig. 9.15. As a first stage, the selective stripping of Sr with a solution H (1 M (NH4)2CO3 + 0.02 M glycine + 0.013 M DTPA) is applied. At the second stage, a solution G (1 M (NH4)2CO3 + 0.039 M DTPA) provides the selective separation of An and Ln. Cesium can be stripped in the last stage using 2 M methylamine carbon­ate solution.

The data presented confirm that it is possible to achieve the separation of nuclide groups by selective stripping.

9.4

Conclusions

As a result of collaboration between RI and INL, the universal extraction system based on CMPO or DPDA, CCD and PEG in polar diluents have been developed, which permit the recovery of Cs, Sr, An and Ln.

The optimal composition of the UNEX-solvent was determined and the conditions for extraction (combined and fraction) of long-lived radionu­clides were established. The high chemical and radiation resistance of the UNEX system, as well as its corrosion, explosion and fire safety, were dem­onstrated under operating conditions.

Variants of work flows based on the UNEX process were developed and tested at pilot plants belonging to KRI, NIKIMT and MCC in Russia, and at INL in the USA, with the use of real and simulated HLW of different compositions. Optimization of the UNEX process resulted in creation of a simple-to-realize work flow involving the three following operations: com­bined extraction of Cs, Sr, An and REE, scrubbing of the extract and com­bined stripping of all radionuclides under study; the recovery rates of radionuclides attained in the course of tests allowed the main HLW bulk to be transferred into the category of low-level waste.

The possibility for realizing the UNEX process on a commercial scale was verified by its testing with the use of the commercial EZR125 centrifugal contactors and with simulated HLW. As to the secondary (end) products of the UNEX process (raffinate and strip product), techniques for their solidi­fication were proven; the traditional technique of cementing was checked for low-level raffinate; the vitrification process for high-level strip products dem­onstrated the possibility of producing glass blocks with a volume of 5 L for every 1 m3 of HLW being treated. The method for regeneration of spent extractant of the UNEX process was elaborated and tested, which made it possible to return more than 90% FS-13 diluent and 40% CCD for re-use.

Thus, the technologies developed and tested for HLW treatment have shown the possibility of deep recovery of radionuclides which allows to transfer the main bulk of wastes into the category of low-level wastes (LLW) suitable for inexpensive near-surface storage. The results of feasibil­ity study, conducted by Idaho National Laboratory have confirmed that the use of UNEX process should reduce the amount of solidified HLW by a factor of 23.