Separations for the Nuclear Fuel Cycle in the 21st Century - American

Chapter 5

Dissolution of Irradiated Nuclear Fuel from the Big Rock Point Reactor

Downloaded by COLUMBIA UNIV on August 10, 2012 | http://pubs.acs.org Publication Date: June 9, 2006 | doi: 10.1021/bk-2006-0933.ch005

Allen J. Bakel, Delbert L. Bowers, Kevin J. Quigley, Monica C. Regalbuto, John A. Stillman, and George F. Vandegrift Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439

Summary The Advanced Fuel Cycle Initiative (AFCI), funded by the Department of Energy, is developing proliferation-resistant technologies that allow safe and economical disposal of waste from reactors. A critical element is the separation of key radionuclides followed by either waste disposal, or conversion of long-lived isotopes to reactor fuel. A sample of Big Rock Point uranium oxide fuel was dissolved in nitric acid at elevated temperature to provide feedstock for the U R E X + demonstration. Elevated temperature led to the complete dissolution of noble metals at relatively low nitric acid concentrations. The conditions used in this study are not suitable for plant-scale application. Three products were obtained: (1) a dissolved fuel solution, (2) undissolved residue, and (3) leached cladding containing no observable undissolved fuel. Elemental analyses of the dissolved fuel, residue, and leached cladding are presented. The data show that 99% of the fuel, including the noble metals was dissolved. The small amount of residue contained primarily Zr, M o , and Pu. While the total amount of residue is small, approximately 20% of the total Pu was found in the residue. Several proposals are made for the prevention of precipitation of the residue.

© 2006 American Chemical Society

In Separations for the Nuclear Fuel Cycle in the 21st Century; Lumetta, G., et al.; ACS Symposium Series; American Chemical Society: Washington, DC, 2006.

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Introduction The Advanced Fuel Cycle Initiative (AFCI), funded by the Department of Energy's Office of Nuclear Energy, is developing advanced proliferationresistant technologies that allow the safe and economical disposal of waste from nuclear reactors. A critical element of this initiative is the separation of key radionuclides followed by either superior waste disposal forms, or conversion of long-lived isotopes to reactor fuel. To that end, the A F C I is developing advanced fuel reprocessing systems that separate key radionuclides from spent fuel. The U R E X + process is a series of five solvent-extraction process steps that separates irradiated nuclear fuel into seven product and waste streams: (1) U 0 for recycle or disposal as L L W , (2) Np/Pu for mixed oxide fuel for thermal reactors, (3) Tc for disposal, (4) I for disposal, (5) Am/Cm for fast-reactor fuel, (6) Cs/Sr for decay storage, and (7) mixed short-lived fission products for repository disposal. The overall process was designed for > 99% recovery of fission products and > 99.99% of actinides. Decontamination factors are 10 -10 as required by process goals. The complete U R E X + solvent extraction process was demonstrated in the Chemical Engineering Division (CMT) of Argonne National Laboratory (ANL) in F Y 03. The purpose of this work was to completely dissolve a pin of irradiated B i g Rock Point uranium oxide fuel to provide high quality feedstock to the U R E X + demonstration done at A N L . In particular, we will monitor the dissolution of the noble metals, which are known to be problematic in fuel dissolution, and uranium, which makes up the bulk of the fiiel. Any solids collected at the end of the dissolution will be characterized to determine their origin and effect on the dissolution product. 3

3

8

8

Experimental Dissolver Equipment. Dissolution of irradiated fuel has previously been done in open vessels with H N 0 and H F [1]. The presence of fluoride ion is undesirable for the proposed U R E X + process because Pu (IV) fluoride complexes are much stronger than other Pu (IV) complexes [2], and extract poorly with T B P [3]. Therefore, we dissolved irradiated nuclear fuel in pure nitric acid at an elevated temperature. A stainless steel (304L) pressure vessel was designed and fabricated. The following criteria were to be met: 1) about 2.5 L in volume, 2) safe operation above 1000 psi, 3) resistance to nitric acid and 4) convenient operation inside a shielded cell facility. A photograph of the cylindrical vessel is shown in Figure 1. A mesh basket was used to hold the cladding/fuel pieces. The dissolver was equipped with a liquid inlet line, a gas vent valve, a pressure gauge, a pressure 3


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relief valve set at U 0 ( N 0 ) + 2/3NO + 4/3H 0. 2

3

2

3

2

2

Scoping tests were done with pressed and sintered depleted U 0 pellets to test the dissolver, to determine the dissolution behavior of the U 0 , and to measure the corrosion behavior of the vessel as a function of temperature. Because irradiated fuel has a different microstructure, different surface morphology and different composition than sintered U 0 , the results of these tests were not quantitatively applied to irradiated nuclear fuel. However, the observed temperature dependence of the dissolution and corrosion rate was used to determine the operational limits of the system. A series of small scale dissolutions of depleted U 0 pellets was carried out in 5.84 M H N 0 at 125, 150, 175°C for 2 hours. Each solution was analyzed for U and H concentrations, and the results are shown in Table 1. The U concentrations were measured by ICP/AES, and were all within 6% of the target. The H concentrations were measured by titrations, and were all within 16% of the target. Data from the scoping tests (Table 1) confirmed that the observed residual pressure of generated gas is equal to or lower than predicted pressure calculated from the suggested chemical reaction and the ideal gas law. A portion of the generated gas might dissolve in the liquid at experimental temperature and 2

2

2

2

3

+

+


74 pressure. These results suggest that the dissolution of U 0 in nitric acid can be accurately represented by the reaction suggested [4]. 2

Table 1. Results from dissolution tests Test temperature


m 125 150 175

U concentration (gU/L) 441 422 413

Total acidity (M) 0.67 0.79 0.78

Residual pressure observed, psi 110 120 160

Residual pressure expected, psi 110 150 220

The extent of corrosion of the vessel during the dissolution tests was also investigated as a function of temperature. The corrosion of stainless steel is known to be temperature-dependant [5] therefore; the corrosion rate represented a limit on the operating temperature. Concentrations of Fe, Cr, and N i were measured in the final solutions, and were used as qualitative measures of corrosion rates. The results in Table 2 show that the corrosion rate is significantly higher at 175°C, than at 150°C. Based on this data, the dissolutions were carried out at temperatures below 175°C.

Table 2. Metal concentrations measured in dissolver solutions from tests conducted at different temperatures Temperature, °C 125 150 175

Cr, ng/mL 13.4 22.5 104

Ni, g/mL 10.0 12.1 59.0

Fe, fig/mL 182 126 446

Irradiated Fuel Dissolution. The first task in the dissolution was to choose the fuel/cladding sections to go into the first dissolution batch (Figure 2). Our initial examination of the sections suggested that about half of the sections were intact, i.e., no fuel had fallen out during the cutting, shipping and unpacking operations. A basket (Figure 1) was placed into the dissolver and 23 weighed fuel/cladding pieces were placed into the basket. The dissolver was sealed, the vent valve was closed, about 800 mL of the acid was pumped into the dissolver, and the heater was switched on. After 8 hours at temperature, the dissolver was cooled overnight. The residual pressure, i.e., the pressure from generated gases, was 180 psi at room temperature. The brown residual gas was vented through a base scrubber. The dissolved fuel solution was then drained from the vessel and weighed. The vessel, basket and cladding pieces were rinsed three times with about 100 mL of dilute H N 0 . Solids were collected during the first two rinses; during the third rinse, no solid material was observed. 3



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Figure 2. Photograph offuel/cladding segments being poured from the shipping container into a pan for examination. The leached cladding removed from the basket contained no visible undissolved fuel (Figure 3). The leached cladding hulls weighed 116.4 g, while the fuel/cladding sections added at the beginning of the first dissolution weighed 492.7 g. Therefore, 376.3 g of material, assumed to be fuel was dissolved. If we assume that 84% of the fuel is U , then 316.1 g of U were dissolved during the first batch dissolution. Two more batches were dissolved. The details of each batch and the calculated totals are shown in Table 3. Table 3. Summary of masses of material dissolved and the products Mass of Fuel/cladding (measured) Mass of Hulls (measured) Mass of Fuel (by difference) Mass of U (fuel *0.84) Mass offuel solution (measured) Volume of dissolvedfuel solution

Dissolution 1

Dissolution 2

Dissolution 3

Total

492.7 g

434.6 g

514.6 g

1442 g

116.4 g

59.7 g

141.2 g

317.3 g

376.3 g

375.3 g

373.4 g

1125 g

316.1 g

315.3 g

313.7 g

945.1 g

1279 g

1330 g

1229 g

3838 g

836 mL

869 mL

803 mL

2510 mL



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Figure 3. Photograph of a representative piece of cladding after fuel dissolution.

Based on the data compiled in Table 3, we dissolved 1124.6 g of fuel (assuming that all of the mass lost during dissolution was fuel) and 945.1 g of U . The total mass of dissolved fuel solution was measured as 3838.3 g, and the density was determined to be 1.53 g/mL. This gives a total volume of 2.51 L of solution. This yields an estimated U concentration of about 377 g U/L. The dissolved fuel was vacuum filtered using a Btichner funnel with a 90 mm Whatman 42 filter paper. A significant amount of solid material was removed from the solution during filtration. The filtered solution was used as the feedstock for the U R E X + demonstration. The solid removed was analyzed and results will be described later in this paper.

Results And Discussion Estimation of fuel composition. As part of the U R E X + demonstration, process flowsheets were designed and modeled using the Argonne Model for Universal Solvent Extraction ( A M U S E ) code [6]. Accurate modeling results can be obtained only if the chemical composition of the feed, dissolved fuel in this case, is well known. The composition of the irradiated fuel from B i g Rock Point reactor had not been determined analytically, so we estimated the dissolved fuel composition using the ORIGEN2 code. Input for the code includes burnup (30,100 MWd/MT), enrichment (4.6% U ) , burnable poison content (1.2 % 235


77 G d 0 ) , and cooling time (21 years) [7]. Additional input was derived from the known, or assumed operating parameters of the Big Rock Point boiling water reactor [8, 9]. The ORIGEN2 code [10] utilizes highly detailed nuclide depletion chains, tracking over 200 actinides and over 800 individual fission product isotopes. The code uses a point depletion model and pre-evaluated libraries of one-group neutron cross section data and decay parameters. Cross section libraries which have been evaluated for several reactor systems with their characteristic neutron spectra are available; specifically, a library for a U0 -fueled B W R is available. Results generated by ORIGEN2 with these "generic" cross section libraries must be carefully considered. Recent comparisons of pressurized water reactor (PWR) spent fuel actinide concentrations generated by ORIGEN2 and the WIMS8 lattice code [11] exhibited differences of several percent for some of the higher mass actinides (isotopes of Pu, Am, and Cm). The higher mass actinides such as Pu-241, Am-241, and Cm-244 are important in spent fuel processing due to their contribution to decay heat and dose. WIMS8 utilizes a multi-group cross section library (172 groups) and two-dimensional spatial modeling to predict the nuclide reaction rates for the particular system under analysis. The code has built-in depletion chains that track the concentrations of 20 actinides, from U 233 to Cm-245. The code also tracks around 100 individual fission products that are important from a neutronics point of view. WIMS8 has been used extensively for P W R spent fuel characterization in the AFCI program. A WIMS8 model of a B W R fuel pin lattice was developed. The model was first benchmarked against spent fuel results reported by O R N L and other external references. The model was expanded to include heterogeneous fuel pin loadings, which are typical in B W R assemblies. The elemental composition of the fuel was calculated using the procedure described above, and is shown in Table 4. The estimated composition was used to design a preliminary flowsheet for the U R E X + process for the demonstration. 2

3


2

Composition of the Dissolver Solution. Aliquots of the filtered dissolved fuel solution were analyzed using inductively-coupled plasma mass spectrometry (ICP-MS) and thermal ionization mass spectrometry (TIMS). The ICP-MS method separates complex mixtures of metal ion on the basis of their atomic mass. The TIMS method allows for the precise and accurate determination of isotope ratios and concentrations of U and Pu. The concentration of metals measured by ICP-MS at any given atomic mass includes the concentrations of all of the isotopes of that mass. For example, the concentration of m/z = 241 ( P u + Am) in the solution was measured as 0.336 parts per million (mg/L). Plutonium has several isotopes, including one with molecular weight of 240; no other isotope expected in the fuel occurs at this mass. Therefore, the m/z=240 data from ICP-MS represents only plutonium with no interference from other elements. The ratio of P u to P u was known from TIMS data. Therefore, the concentration of P u in the dissolved fuel can be 24l

2 4 1

240

241

241


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Table 4. Estimated elemental composition of an irradiated fuel rod from the Big Rock Point reactor calculated using the O R I G E N 2 code.


Element Rb Sr Y Zr Mo Tc Ru Rh Pd Ag Cd Sn Sb Te Cs Ba La Ce Nd Sm Eu Gd U Np Pu Am 1

2

3

g/MTHM in fuel' 3.45E+02 6.49E+02 4.43E+02 3.49E+03 3.00E+03 6.97E+02 1.87E+03 3.85E+02 9.83E+02 5.50E+01 7.78E+01 7.90E+01 1.68E+01 4.19E+02 1.99E+03 2.07E+03 1.10E+03 2.13E+03 3.67E+03 7.91E+02 9.81E+01 9.18E+03 9.60E+05 3.65E+02 8.74E+03 8.17E+02

g/906-g if 3.26E-01 6.12E-01 4.17E-01 3.29E+00 2.83E+00 6.57E-01 1.76E+00 3.63E-01 9.27E-01 5.19E-02 7.34E-02 7.45E-02 1.58E-02 3.95E-01 1.88E+00 1.95E+00 1.04E+00 2.01E+00 3.46E+00 7.46E-01 9.25E-02 8.65E+00 9.05E+02 3.44E-01 8.24E+00 7.70E-01

3

g/L 1.30E-01 2.44E-01 1.66E-01 1.31E+00 1.13E+00 2.62E-01 7.01E-01 1.45E-01 3.69E-01 2.07E-02 2.92E-02 2.97E-02 6.29E-03 1.57E-01 7.49E-01 7.77E-01 4.14E-01 8.01E-01 1.38E+00 2.97E-01 3.69E-02 3.45E+00 3.61E+02 1.37E-01 3.28E+00 3.07E-01

-grams of the indicated element per metric ton of fuel (on a heavy metal basis) -grams of the indicated element per the total amount of uranium dissolved (906 g). - These values are calculated assuming a final solution volume of 2.51 L


79 calculated using the total m/z=241 concentration from ICP-MS data and the P u to P u ratio from TIMS data. Such calculations were carried out at each m/z value. The total concentration of any element was calculated by the sum of the concentrations of all of its isotopes. For example, the total Pu concentration was calculated by summing the concentrations of P u , P u , P u , P u , and P u . The totals for the concentrations of each element present in the dissolved fuel are shown in Table 5. 240

241

238

239

240

241

242


Table 5. Elemental concentrations of dissolved fuel using I C P / M S and T I M S data, and the O R I G E N calculations. Element Rb Sr Y Zr Mo Tc Ru Rh Pd Ag Cd Sn Sb Te Cs Ba' La Ce Nd Sm Eu Gd U Np Pu Am 23

2

2

2

3

3

3

2

2

1

2

3

Measured concentrations, g/L 1.10E-01 2.59E-01 1.47E-01 2.14E-01 1.63E-01 2.55E-01 7.75E-01 1.70E-01 6.60E-02 7.11E-04 5.01E-02 6.19E-02 5.14E-01 5.64E-02 6.26E-01 1.36E+00 3.88E-01 6.70E-01 1.39E+00 2.69E-01 3.59E-02 2.64E+00 3.61E+02 1.68E-01 2.68E+00 3.08E-01

ORIGEN prediction, g/L 1.30E-01 2.44E-01 1.66E-01 1.31E+00 1.13E+00 2.62E-01 7.01E-01 1.45E-01 3.69E-01 2.07E-02 2.92E-02 2.97E-02 6.29E-03 1.57E-01 7.49E-01 7.77E-01 4.14E-01 8.01E-01 1.38E+00 2.97E-01 3.69E-02 3.45E+00 3.61E+02 1.37E-01 3.28E+00 3.07E-01

- This value might include Xe that interferes at masses 134 and 135. - These elements are present in the residue. - These elements are present in the cladding



80 Such calculations can only be made i f accurate isotopic data are available. Results from TIMS analyses are ideal for this purpose. If the TIMS data were not available, the needed isotope ratios could be obtained from O R I G E N code results. Only U and Pu were analyzed by TIMS, isotope ratios for other elements were obtained from O R I G E N calculations. Table 5 compares the results of TIMS and the predictions from O R I G E N for individual isotopes of U and Pu. These results suggest that i f TIMS data were not available, then O R I G E N results could also be used. Data in Table 6 allows comparison of the measured elemental concentrations of the dissolved fuel and the concentrations predicted by the O R I G E N code calculations (Table 4). The agreement is good in most cases. This good agreement shows that the dissolution was successful. In particular, the results for the noble metals Ru and Rh show that the fuel dissolution was effective.

Table 6. Concentrations of individual U and Pu isotopes from TIMS analyses and ORIGEN code predictions. Isotope 233

u

235

[J

236

u u

238

238

Pu Pu Pu Pu Pu Pu

m

240

24,

242

2U

Results of TIMS analysis

Separations for the Nuclear Fuel Cycle in the 21st Century - American

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