New approaches to reprocessing of oxide nuclear fuel
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Dissolution of UO2, U3O8, and solid solutions of actinides in UO2 in subacid aqueous solutions (pH 0.9–1.4) of Fe(III) nitrate was studied. Complete dissolution of the oxides is attained at a molar ratio of ferric nitrate to uranium of 1.6. During this process actinides pass into the solution in the form of U(VI), Np(V), Pu(III), and Am(III). In the solutions obtained U(VI) is stable both at room temperature and at elevated temperatures (60 °C), and at high U concentrations (up to 300 mg mL−1). Behavior of fission products corresponding to spent nuclear fuel of a WWER-1000 reactor in the process of dissolution the simulated spent nuclear fuel in ferric nitrate solutions was studied. Cs, Sr, Ba, Y, La, and Ce together with U pass quantitatively from the fuel into the solution, whereas Mo, Tc, and Ru remain in the resulting insoluble precipitate of basic Fe salt and do not pass into the solution. Nd, Zr, and Pd pass into the solution by approximately 50 %. The recovery of U or jointly U + Pu from the dissolution solution of the oxide nuclear fuel is performed by precipitation of their peroxides, which allows efficient separation of actinides from residues of fission products and iron.
KeywordsSpent oxide nuclear fuel Dissolution Ferric nitrate solution Fission products
The progress of atomic power engineering in Russia directed towards on closed nuclear fuel cycle, which makes it necessary to develop novel innovation environmentally safe and economically advantageous low-waste technologies for reprocessing of spent nuclear fuel (SNF) both from the operating reactors and from fast reactors operating on mixed uranium–plutonium oxide fuel. To solve this problem SNF dissolution can be performed in subacid aqueous solutions of ferric nitrate, in which the acidity after the fuel dissolution is close to 0.1 M and allows direct recovery of U and Pu by precipitation of their fluorides, carbonates, oxalates, or peroxides [1, 2]. In this feature, the suggested SNF reprocessing technology differs from the Purex process in which the fuel is dissolved in concentrated HNO3 and then U and Pu are extracted from the strongly acidic solution with tributyl phosphate solutions in organic solvents with the subsequent stripping with subacid aqueous solutions (~0.1 M). As a result, 7–12 tons of acidic aqueous and organic solutions are formed as the waste per a ton of the reprocessed SNF. These solutions require further reprocessing and disposal. Thus, high acidity of the solutions in the SNF reprocessing technology is its major drawback. In this work we show that the use of subacid ferric nitrate solutions allows application of strong nitric acid to be discarded, which results in reduction in the waste solution volume, and mitigate the environmental impact of the waste. After the dissolution of the oxide fuel, the recovery of U and Pu from the solution is performed by precipitation of their peroxides. The behavior of the large number of fission products (FPs) in the course of dissolution of simulated SNF in ferric nitrate solutions and the recovery of uranium from these solutions was studied.
The commercial samples of UO2, U3O8, pellets enriched with 235U and MOX fuel (95.4 % 238UO2 and 4.6 % 239PuO2) in both powdered and granulated forms were used. Solid solutions of NpO2 and AmO2 in UO2 were synthesized from U–Np and U–Am oxalate mixtures, prepared in advance, by calcining these mixtures in an atmosphere of Ar +20 % H2 at 850 °C for 8 h. The data on the contents of FPs in irradiated fuel of WWER-1000 reactor (UO2, initial enrichment 5.5 % 235U, burn-up fraction 80 MW day kg−1), given in  we used for preparation of simulated SNF (SSNF). The calculated amounts of the elements––simulators of FPs in the form of their salts, manly nitrates or chlorides, were introduced into 3 M nitric acid solution of uranium. The solution obtained was evaporated up to solid residue and then calcinated in an oven at 850 °C in Ar + 10 % H2 atmosphere. The relative content of FPs in the SSNF sample obtained (counting on the sum of metals) were as follows (wt%): Cs 0.60, Sr 0.19, Ba 0.35, Y 0.10, La 0.28, Ce 0.59, Nd 0.84, Zr 0.80, Mo 0.79, Tc 0.18, Ru 0.56, Pd 0.40 (ΣFP 5.68 %), and U 94.32 %. In SSNF reprocessing, the concentrations of most mentioned FPs in the resulting solutions were determined by atomic emission spectroscopy. It was shown that the presence of 100- and 200-fold amounts of U and Fe did not noticeably affect the accuracy of FPs determination. Weighed portions of the powdery oxides or SSNF samples were introduced into polypropylene centrifuge test tubes containing aqueous ferric nitrate solutions [Fe(NO3)3·9H2O] with pH 0.9–1.4. It should be noted that such acidity in the solutions of Fe(III) nitrate is caused by hydrolysis of these salts. Naturally, the solutions contain partially hydrolyzed soluble species Fe(OH)(NO3)2. The test tubes were hermetically stoppered and placed into a stirring device. After definite time intervals, the stirring was stopped. The suspensions were centrifuged, and the aqueous phases (mother liquors) were analysed. For that, aliquots of the solutions were deposited onto the targets (polished stainless steel disks), dried and calcined for measuring their α-activity, using α-spectrometer Alpha Analyst (Canberra). The uranium contents after dissolving MOX fuel samples was determined by spectrophotometry, because reliable radiometric determination of 238, 235, 234U at ~5 wt% 239Pu content in the solutions is impossible. The Tc content was found by measuring its β-activity on a UMF-2000 radiometer. The behavior of Cs and Fe was monitored after spiking the solutions with 137Cs and 59Fe, using a γ-ray spectrometer with a semiconductor germanium detector (Canberra). The concentrations were calculated from the γ-activity using Genie-2000 program. The oxidation states of U, Pu, Np, and Am in solutions were determined from the electronic absorption spectra recorded with a Unicam UV-340 spectrophotometer. The pH was measured with a Mettler Toledo MP230 pH meter with a combined glass electrode (Hanna Instrument HI 1131B) calibrated using buffer pH standards (pH 1–13, Merck).
Results and discussion
Dissolution of uranium oxides in subacid ferric nitrate solutions
Content of U(VI) in the precipitate of basic ferric nitrate at 60 °C in relation to the U(VI) content in the solution
U(VI) in solution
U(VI) in precipitatea
Concentration (mg mL−1)
1,498 ± 25
1,510 ± 30
1,478 ± 19
1,490 ± 20
The broad absorption band at wavelengths from 700 to 1,050 nm with a maximum at 950 nm belongs to the Fe2+ cations arising in the course of the UO2 dissolution.
Dissolution of MOX fuel and of solid solutions of NpO2 and AmO2 in UO2 in subacid ferric nitrate solutions
Dissolution of MOX fuel in a ferric nitrate solution (solution volume 4 mL, pH ~ 1, molar ratio Fe(NO3)3·9H2O:MOX fuel ~ 2)
MOX fuel taken (mg)
Found in solution (mg)
120.0 ± 0.1
114.5 ± 0.1
5.5 ± 0.1
111 ± 5
5.0 ± 0.5
As can be seen, U occurs in the solution in the form of U(VI), and Pu, in the form of Pu(III), because Pu(IV) initially present in the MOX fuel is reduced in the course of dissolution to Pu(III) with Fe(II) ions formed by oxidation of U(IV) to U(VI) with Fe(III). Solid solutions of mixed oxides NpO2–UO2 and AmO2–UO2 readily dissolve in the ferric nitrate solutions as well. The spectrophotometry analysis shows, that the dissolution of the mixed oxide UO2–NpO2 is accompanied by oxidation of Np(IV) to Np(V) with Fe(III) cations, similar to the oxidation of U(IV) to U(VI). On dissolution of UO2–AmO2, americium which is present in the crystal lattice of the mixed oxide as Am(IV) is reduced to Am(III) in the course of dissolution, as it is commonly observed on dissolution of AmO2 in mineral acids .
Recovery of U and Pu by precipitation of peroxides from the ferric nitrate solution
Peroxide precipitation of U and Pu from nitrate solution (pH ~ 1) and their separation from the Fe(III)
Element concentration (M)
6.8 × 10−2
4.4 × 10−3
6.6 × 10−2
4.2 × 10−3
6.0 × 10−5
0.2 × 10−2
0.2 × 10−3
Behavior of fission products in dissolution of SSNF in the ferric nitrate solution
Content of FPs, U, and Fe in the ferric nitrate solution (pH ~ 0.5) and in the precipitate formed on dissolution of SSNF
Calculated concentration at complete SSNF dissolution (mg mL−1)
Found in solution
After dissolution of SSNF
After dissolution of precipitate of basic Fe salt (wt%)
99.4 ± 0.6
Behavior of fission products in the course of uranium recovery from subacid aqueous ferric nitrate solutions by precipitation of its peroxide
Content of FPs, U, and Fea in the ferric nitrate solution with pH ~ 1 and in the precipitate formed on the SSNF dissolution
Calculated concentration at complete dissolution of SSNF (mg mL−1)
Relative content of FPs (%)
After separation of U peroxide
After dissolution of U peroxide
There was suggested an alternative method for SNF reprocessing using subacid solutions of ferric nitrate, which allow dissolving SNF and recovering actinides from these solution in the form of their peroxides and efficient separation of actinides from FPs and Fe. As compared to the currently used PUREX process, the number of steps of SNF reprocessing decreases, the liquid waste volume is considerably reduced, and the fuel reprocessing is made safer and more reliable.
The study was performed within the framework of implementation of the Federal Target Program “Scientific and Scientific-Pedagogical Personnel of Innovative Russia” for the years 2009–2013 and was financially supported by the Russian Federation for Basic Research (Project No. 11-03-12033 OFI-m; No. 12-03-00661a; No. 12-08-00559a).
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- 1.Shevchenko VB, Sudarikov BN (1961) Tekhnologiya urana (uranium technology). Gosatomizdat, Moscow, p 236Google Scholar
- 2.Milyukova MS, Gusev NI, Sentyurin IG, Sklyarenko IS (1965) Analiticheskaya khimiya plutoniya (analytical chemistry of plutonium). Nauka, Moscow, pp 288–290Google Scholar
- 3.Kevrolev VV (2000) Rekol—continuons energy Monte Carlo code for neutron transport, Preprint of the Kurchatov Institute of Atomic Energy, 1993, no. IAE- 562115; Kevrolev, V.V. et al., Report of the Russian Research Centre Kurchatov InstituteGoogle Scholar