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Dissolution of Irradiated Commercial UO2 Fuels in Ammonium Carbonate and Hydrogen Peroxide Chuck Z. Soderquist,* Amanda M. Johnsen, Bruce K. McNamara, Brady D. Hanson, Jeffrey W. Chenault, Katharine J. Carson, and Shane M. Peper Pacific Northwest National Laboratory, P.O. Box 999, Mail Stop P7-25, Richland, Washington, 99352, United States ABSTRACT: We propose and test a disposition path for irradiated nuclear fuel using ammonium carbonate and hydrogen peroxide media. We demonstrate on a 13 g scale that >98% of the irradiated fuel dissolves. Subsequent expulsion of carbonate from the dissolver solution precipitates >95% of the plutonium, americium, and curium and substantial amounts of fission products, effectively partitioning the fuel at the dissolution step. Uranium can be easily recovered from solution by any of several means, such as ion exchange, solvent extraction, or direct precipitation. Ammonium carbonate can be evaporated from solution and recovered for reuse, leaving an extremely compact volume of fission products, transactinides, and uranium. Stack emissions are predicted to be less toxic, less radioactive, chemically simpler, and simpler to treat than those from the conventional PUREX process.
1. INTRODUCTION Commercial nuclear power has re-emerged as a low-carbon alternative energy source, but concerns regarding the disposition of irradiated nuclear fuel have become the most significant challenge facing a potential nuclear renaissance. Disposition options include geologic repositories, long-term dry storage, and reprocessing. The focus of this work is on the development of a reprocessing scheme with simpler chemistry and improved waste streams. Most commercial nuclear fuel reprocessing operations worldwide use the PUREX process, where irradiated fuel is dissolved in hot 8-12 M nitric acid. The nitric acid oxidizes UO2 to UO22þ and dissolves all the actinides and nearly all the fission products, leaving only the noble metal fission products (an alloy of molybdenum, technetium, ruthenium, rhodium, and palladium) undissolved as a black sludge. The halogens (including 129I), noble gases (including 85Kr), ruthenium, 3H, and 14C are largely evaporated from solution, accompanied by a large volume of NOx. After the fuel has dissolved, uranium and plutonium are recovered from the solution by extraction into tributyl phosphate dissolved in C10-C13 straight-chain hydrocarbons. Most of the nitric acid is evaporated and recovered from the dissolver solution. The remaining solution is either left acidic (most modern PUREX processes) or neutralized with sodium hydroxide (formerly done for plutonium production in the United States) and stored in tanks as high-level radioactive waste. The organic solvent is scrubbed and reused but must eventually be disposed of as mixed hazardous and radioactive waste. To avoid the acids, the mixed radioactive-organic solvent waste stream, and the stack emissions associated with the PUREX process, we propose and investigate a method for reprocessing irradiated nuclear fuel using ammonium carbonate and hydrogen peroxide, followed by simple chemical processing steps. All of the oxides of uranium will dissolve in a carbonate solution under oxidizing conditions. The cation UO22þ readily forms a series of stable, soluble anionic carbonate complexes1-4 and stays in solution under conditions where most of the lanthanide r 2011 American Chemical Society
and transition metals form insoluble hydroxides and carbonates. This chemistry has been known since the 19th century and was the basis for some early analytical uranium separations.5-7 It has been used to leach uranium from ores since the 1950s,8 and carbonate leaching has been suggested for recovering uranium from contaminated soil.9 Dissolution of solid UO2 in carbonate solution has been studied extensively, mostly to predict the behavior of spent fuel in a repository.10,11 Very recently, our group12 and others13-15 have investigated using carbonate dissolution to recover uranium and possibly other constituents from irradiated nuclear fuel; in a previously published work,12 our group demonstrated on a 50 mg scale that crushed fuel dissolves readily in an ammonium carbonate-hydrogen peroxide solution but leaves the noble metal phase largely unattacked. We chose ammonium carbonate for this work because ammonium carbonate can be completely evaporated from solution, leaving no residue. This allows the final, processed fuel mass and volume to be not much more than the starting mass and volume of UO2 fuel. Other carbonates such as sodium or potassium carbonate will work well chemically, but they cannot be easily removed from solution and increase the bulk of the product. A number of oxidizing agents will oxidize UO2 to UO22þ in carbonate solution.16 We chose hydrogen peroxide because it is inexpensive and easy to handle, works well chemically, and decomposes easily to water and oxygen gas, again leaving no residue. Hydrogen peroxide also serves as a strong ligand for UO22þ and certain other elements found in irradiated fuel (particularly Mo,17 Np, and Pu18). Dissolution of the fuel in an ammonium carbonate-hydrogen peroxide solution has several attractive advantages over the nitric acid used in the PUREX process. The dissolution takes place at room temperature and atmospheric pressure and is essentially Received: June 29, 2010 Accepted: December 19, 2010 Revised: December 13, 2010 Published: January 18, 2011 1813
dx.doi.org/10.1021/ie101386n | Ind. Eng. Chem. Res. 2011, 50, 1813–1818
Industrial & Engineering Chemistry Research
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noncorrosive. The only gases generated are O2 (from decomposition of hydrogen peroxide), NH3, CO2, and H2O (from evaporation of ammonium carbonate solution), and krypton and xenon (fission products). Nitric acid is not used, and no NOx is generated. The pH of a 1 M ammonium carbonate solution is about 10, and iodine and ruthenium are not volatile at this pH. The stack gases are therefore much less radioactive than those produced by the PUREX process, simplifying stack gas scrubbing and filtration. The dissolved uranium is left in the well-known chemical form UO2(CO3)34- and can be readily recovered from the solution by any of several means such as solvent extraction, solid phase ion exchange, or precipitation. Because the reagents used in this process are completely volatile, this process produces a particularly compact volume of separated fission products and transuranic elements. The ultimate intent of this work is the virtual elimination of the irradiated nuclear fuel, by separating it into more manageable components. After irradiated fuel has been out of the reactor about 30 years, most of the remaining radioactivity and most of the heat generated is caused by only two radionuclides, 90Sr and 137 Cs. If strontium and cesium are removed from the fuel, then what remains will have low dose and can be handled with very light shielding. 90Sr has a 29-year half-life and does not require storage on a geologic time scale. 137Cs likewise has a 30-year halflife and also does not require geologic storage (although it is accompanied by a small amount of very long-lived 135Cs which will remain after the 137Cs has decayed). The remaining fuel is still quite radiotoxic, but that radiotoxicity can be substantially reduced by removing the transuranic elements. The transuranic elements can be burned in a reactor, some in an ordinary light water power reactor and the others in a fast reactor. The rest of the fuel, with most of the dose, heat, and radiotoxicity removed, can be stored in a geologic repository. In this paper we demonstrate dissolution of actual commercial spent fuel followed by partitioning into several useful fractions, using only simple chemistry and few reagents.
2. CHEMISTRY OF THE CARBONATE-PEROXIDE DISSOLUTION PROCESS In the presence of ammonium carbonate and hydrogen peroxide, uranium oxidizes and forms a series of carbonateperoxide complexes, which ultimately convert to the soluble anion UO2(CO3)34-. The overall dissolution stoichiometry for UO2 is as follows UO2 þ H2 O2 þ 3ðNH4 Þ2 CO3 f ðNH4 Þ4 UO2 ðCO3 Þ3 þ 2NH4 OH
ð1Þ
Actual consumption of H2O2 is greater than predicted by eq 1 because of the formation and decomposition of various uranyl peroxide-carbonate complexes, the first of which is shown in eq 2 UO2 ðCO3 Þ3 4 - þ H2 O2 f UO2 ðCO3 Þ2 ðO2 Þ4 þ 2Hþ þ CO3 2 -
ð2Þ
Formation of these uranyl peroxide complexes will release hydrogen ions (eq 2). The mixed uranyl peroxide-carbonate complexes vary in their stability, with lifetimes from hours to weeks,
Table 1. Irradiated UO2 Fuels Used for Dissolution type of
burnup,
years out
fuel
reactor
GWd/MTU
of reactor
ref
mass of dissolved fuel, g
ATM-105
BWR
30
27
19
13.0868
ATM-106
PWR
45
28
20
13.1677
ATM-109
BWR
60
16
21
12.0536
and decompose to O2, water, and carbonate complexes. The decomposition of O22- into O2 and water consumes Hþ 2O2 2 - þ 4Hþ f 2H2 O2 f 2H2 O þ O2
ð3Þ
The fuel dissolution reaction generates hydroxide (eq 1), but the changes in pH are small, since the solution is heavily buffered with ammonium ion and carbonate ion. The formation and decomposition of various mixed carbonate-peroxide complexes may cause the pH to wander slightly during the fuel dissolution process, depending on the uranium concentration and the quantity of hydrogen peroxide added. The compound sold commercially as ammonium carbonate is actually a mixture of ammonium carbonate and ammonium carbamate. A saturated solution of ammonium carbonate at room temperature is about 2.2 M in carbonate (measured gravimetrically by calcium carbonate precipitation) but much higher than that in total ammonium. The high ammonium ion concentration used to dissolve the UO2 fuel will affect the solubility of several compounds in the reacting mixture, partly by the common ion effect and partly by raising the total ionic strength. As the dissolution proceeds on actual spent fuel, two phenomena may slow the reaction: (1) higher actinides and fission products may not dissolve as readily as UO2, leaving behind an insoluble rind, coating the UO2 grains; (2) as uranium and carbonate concentrations change during dissolution, any actinides and fission products that dissolved initially may reprecipitate, isolating undissolved fuel from the dissolver solution. The first scenario may be avoided by tumbling the pieces of raw fuel during dissolution to continuously knock off the rind or the fuel can be crushed into small pieces, too small to form much rind. The second scenario may be prevented by periodic replacement of the saturated dissolver solution with fresh solution ammonium carbonate and hydrogen peroxide.
3. EXPERIMENTAL SECTION Approximately 2-cm segments of three commercial UO2 fuels were used, ranging from moderate to very high burnup, as shown in Table 1. Each fuel had previously been extensively characterized to serve as a testing material. Each fuel was removed from its cladding, crushed with a commercial sample crusher (piston-in-cylinder), and passed through a 212-μm soil sieve. The fuel was crushed to avoid the formation of a rind that could slow dissolution rates. The crushed, sieved fuel was placed in a tantalum crucible containing alumina ball-mill balls. The dissolution and processing of the three irradiated fuels followed the scheme shown in Figure 1. One hundred fifty milliliters of saturated ammonium carbonate solution and 30 mL of 30% hydrogen peroxide were added, and the crucible was placed on an orbital shaker. Gentle shaking allowed the alumina balls to roll around and stir the mixture. The solution turned orange with dissolved uranium almost immediately. (Color is 1814
dx.doi.org/10.1021/ie101386n |Ind. Eng. Chem. Res. 2011, 50, 1813–1818
Industrial & Engineering Chemistry Research
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Figure 1. Dissolution and processing of irradiated UO2 fuel with ammonium carbonate and hydrogen peroxide.
hard to accurately see through the hot cell window. Yellow solutions appear colorless. We use an in-cell video camera to see color more accurately and to examine fine detail.) The hydrogen peroxide decomposed much faster than expected. Within only an hour or two, the effervescence stopped, the orange color of the uranyl peroxide-carbonate complexes disappeared, the solution turned yellow, and a coarse yellow precipitate formed, which we assume is (NH4)4UO2(CO3)3. Periodic additions of hydrogen peroxide were made to keep the reaction going (four additions of 30 mL apiece). Every day, the orbital shaker was turned off and the solutions were allowed to settle. The clear solution saturated with uranium was removed, fresh ammonium carbonate solution and hydrogen peroxide were added (150 mL of ammonium carbonate and 30 mL of hydrogen peroxide), and stirring was resumed. We judged the dissolution to be complete when the solution no longer had any yellow color after stirring with the undissolved fuel for several hours. The undissolved material at that point was a fine black sludge which settled readily. This dissolution ran for five days altogether, although full dissolution would be reached much faster now that the rapid hydrogen peroxide decomposition and the need for solution replacement are known. After the dissolution was apparently complete (no more orange or yellow color appeared in fresh ammonium carbonate-hydrogen peroxide solution), the solution was filtered through a 0.45-μm cellulose nitrate membrane filter. The material on the filter cake was returned to the reaction crucible, rinsed with fresh ammonium carbonate solution, and filtered through a fresh membrane filter. The filter cake was a thick, fine black sludge. The clear, yellow filtrates from each fuel were warmed to expel much of the ammonium carbonate, and the solution was filtered again. Simple heating does not accurately adjust the ammonium carbonate solution to a particular concentration. We did not have an accurate, simple way to monitor the ammonium carbonate concentration in the hot cell. The solids on the filter cake were combined with previously obtained solids. The total volume of solution for each of the three fuels was 1000-1200 mL. Each of the filtered fuel solutions was passed through a cation exchange column (Bio-Rad AG 50-X12, NH4þ form, 50-100 mesh, 100-mL column volume) to remove cesium and reduce the dose, so that the solution could be removed from the hot cell and handled in a fume hood. After 200-300 mL of solution had
passed through the column, the column was rinsed with 0.01 M ammonium carbonate and then stripped with saturated ammonium carbonate. The strip solution, which had high 137Cs activity, was collected separately. After the cesium was stripped, the column was rinsed with 100 mL of 0.01 M ammonium carbonate to condition the column for further additions of fuel solutions. The cesium fractions for each fuel were evaporated dry and redissolved in 10.0 mL of 0.5 M nitric acid, labeled Cesium Fraction. The uranium solutions, labeled Uranium Fraction, were removed from the hot cell and taken to a fume hood. The undissolved solids were suspended in 0.5 M nitric acid to dissolve the hydroxides and carbonates and centrifuged. The remaining insoluble material was dried, weighed, and labeled Noble Metal fraction. Any unreacted UO2 fuel particles would have reported to the Noble Metal fraction. Each 0.5 M nitric acid solution was boiled to expel any remaining CO2 and made basic with concentrated ammonium hydroxide to precipitate most of the fission products. The only major fission products that are soluble under these conditions are barium and strontium. Each solution was centrifuged. The solutions were evaporated dry, redissolved in 10.0 mL of 0.5 M nitric acid, and labeled Strontium Fraction. The precipitate from the bottom of the centrifuge tube was dissolved in 0.5 M nitric acid, adjusted to 10.0 mL, and labeled Fission Product Fraction. All of the fractions were analyzed for gamma emitters, 90Sr, 239þ240Pu, 241 Am, 242Cm, and 243þ244Cm. The uranium fractions, as removed from the hot cell, were determined to have significant amounts of dissolved europium, plutonium, americium, and curium, indicating that they still contained free ammonium carbonate in solution. The uranium fractions were then warmed gently to expel most of the ammonium carbonate; warming was stopped when the solution began to cloud slightly from precipitation of uranium. The solution was allowed to cool and filtered through a 0.45 μm membrane filter. The filter cake was dissolved in 2 M nitric acid, labeled Second Precipitate, and analyzed like the previous fractions. Each filtrate (Final Uranium Solution) was reanalyzed as before.
4. RESULTS AND DISCUSSION The final results (Table 2) show that the ammonium carbonatehydrogen peroxide process easily dissolves irradiated commercial 1815
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Industrial & Engineering Chemistry Research
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Table 2. Results of Ammonium Carbonate-Hydrogen Peroxide Dissolution of Irradiated UO2 Fuels percentage of entire inventory fuel fraction V
UTotal
90
Sr
125
Sb
137
154
Cs
Eu
237
Np
239þ240
Pu
241
Am
243þ244
Cm
ATM-105 uranium
99.1
0.35
52.0
1.35
22.5
65.5
49.9
15.1
11.5
cesium
0.69
59.8
10.8
98.3
