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New approaches to reprocessing of oxide nuclear fuel.

Myasoedov BF, Kulyako YM - J Radioanal Nucl Chem (2012)

Bottom Line: 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)).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 %.

View Article: PubMed Central - PubMed

Affiliation: Vernadsky Institute of Geochemistry and Analytical Chemistry, Russian Academy of Sciences, ul. Kosygina 19, Moscow, 119991 Russia.

ABSTRACT

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.

No MeSH data available.


Related in: MedlinePlus

Spectrum of the solution after dissolution of MOX fuel in the ferric nitrate solution. (pH ~ 1), [Fe(NO3)3·9H2O] = 0.67 M, [U] = 0.212 M, [Pu(III)] = 5.1·10−3 M
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Fig4: Spectrum of the solution after dissolution of MOX fuel in the ferric nitrate solution. (pH ~ 1), [Fe(NO3)3·9H2O] = 0.67 M, [U] = 0.212 M, [Pu(III)] = 5.1·10−3 M

Mentions: Data on the dissolution of MOX fuel taken as a powder in a ferric nitrate solution (pH ~ 1) at ~22 °C are given in Table 2. Our results show that MOX fuel, like UO2, also dissolves in this medium. Figure 4 shows the spectrum of the solution obtained after dissolution of MOX fuel in the ferric nitrate solution.Table 2


New approaches to reprocessing of oxide nuclear fuel.

Myasoedov BF, Kulyako YM - J Radioanal Nucl Chem (2012)

Spectrum of the solution after dissolution of MOX fuel in the ferric nitrate solution. (pH ~ 1), [Fe(NO3)3·9H2O] = 0.67 M, [U] = 0.212 M, [Pu(III)] = 5.1·10−3 M
© Copyright Policy
Related In: Results  -  Collection

Show All Figures
getmorefigures.php?uid=PMC4514683&req=5

Fig4: Spectrum of the solution after dissolution of MOX fuel in the ferric nitrate solution. (pH ~ 1), [Fe(NO3)3·9H2O] = 0.67 M, [U] = 0.212 M, [Pu(III)] = 5.1·10−3 M
Mentions: Data on the dissolution of MOX fuel taken as a powder in a ferric nitrate solution (pH ~ 1) at ~22 °C are given in Table 2. Our results show that MOX fuel, like UO2, also dissolves in this medium. Figure 4 shows the spectrum of the solution obtained after dissolution of MOX fuel in the ferric nitrate solution.Table 2

Bottom Line: 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)).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 %.

View Article: PubMed Central - PubMed

Affiliation: Vernadsky Institute of Geochemistry and Analytical Chemistry, Russian Academy of Sciences, ul. Kosygina 19, Moscow, 119991 Russia.

ABSTRACT

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.

No MeSH data available.


Related in: MedlinePlus