Current separation and purification technologies utilized in the nuclear fuel cycle rely primarily on liquid-liquid extraction and ion-exchange processes. Here, we report a laboratory-scale aqueous process that demonstrates nanoscale control for the recovery of uranium from simulated used nuclear fuel (SIMFUEL). The selective, hydrogen peroxide induced oxidative dissolution of SIMFUEL material results in the rapid assembly of persistent uranyl peroxide nanocluster species that can be separated and recovered at moderate to high yield from other process-soluble constituents using sequestration-assisted ultrafiltration. Implementation of size-selective physical processes like filtration could results in an overall simplification of nuclear fuel cycle technology, improving the environmental consequences of nuclear energy and reducing costs of processing.

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The feasibility of decontaminating spent fuel cladding hulls using hydrofluoric acid (HF) was investigated as part of the Global Energy Nuclear Partnership (GNEP) Separations Campaign. The concentrations of the fission product and transuranic (TRU) isotopes in the decontaminated hulls were compared to the limits for determining the low level waste (LLW) classification in the United States (US). The 90 Sr and 137 Cs concentrations met the disposal criteria for a Class C LLW; although, in a number of experiments the criteria for disposal as a Class B LLW were met. The TRU concentration in the hulls generally exceeded the Class C LLW limit by at least an order of magnitude. The concentration decreased sharply as the initial 30-40 μm of the cladding hull surface were removed. At depths beyond this point, the TRU activity remained relatively constant, well above the Class C limit.

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Flowsheet parameters for the dissolution of SRE/DR-3 fuel include the following. The Hg catalyst is added gradually after the dissolver has reached temperature to achieve a maximum catalyst concentration of 0.012-0.015 M. The initial nitric acid concentration is in the range of 6-7 M and dependent on the amount of Al, Th, and U to be dissolved, targeting a final nitric acid concentration of 0.5-1.0 M after completion of the dissolution of the last charge. Boric acid (H 3 BO 3) or gadolinium nitrate (Gd(NO 3) 3) may be used as a nuclear safety poison. Concentrations of up to 2 g/L of B or Gd in surrogate dissolver solutions have been observed to be stable from precipitation. Hydrogen flammability calculations were performed using the experimental data to determine the conditions that H-Canyon can safely dissolve SRE/DR-3 fuel in L-Bundles with respect to the H 2 levels during the projected peak off-gas rates. To stay under the 60% LFL for H 2 , the charges of L-Bundles containing SRE shall be limited to four, where the Hg concentration is added to the recommended 0.012-0.015 M. In order to process the L-Bundles of DR-3 fuel, a minimum of 0.17 M Al must be in solution. This minimum dissolved Al could be reached by first dissolving SRE fuel or by adding Al(NO 3) 3 to the dissolver solution. The number of L-Bundles of DR-3 fuel that could be charged successively to the H-Canyon dissolver is dependent on the concentration of U in the U-Al alloy and the dissolved Al concentration. The number of bundles increases as the Al concentration increases in the dissolving solution. Instructions and limitations regarding the use of this document are provided in the transmittal letter.

The H-Canyon facility will be used to dissolve Pu metal for subsequent purification and conversion to plutonium dioxide (PuO 2) using Phase II of HB-Line. To support the new mission, the development of a Pu metal dissolution flowsheet which utilizes concentrated (8-10 M) nitric acid (HNO 3) solutions containing potassium fluoride (KF) is required. Dissolution of Pu metal in concentrated HNO 3 is desired to eliminate the need to adjust the solution acidity prior to purification by anion exchange. The preferred flowsheet would use 8-10 M HNO 3 , 0.015-0.07 M KF, and 0.5-1.0 g/L Gd to dissolve the Pu up to 6.75 g/L. An alternate flowsheet would use 8-10 M HNO 3 , 0.1-0.2 M KF, and 1-2 g/L B to dissolve the Pu. The targeted average Pu metal dissolution rate is 20 mg/min-cm 2 , which is sufficient to dissolve a "standard" 2250-g Pu metal button in 24 h. Plutonium metal dissolution rate measurements showed that if Gd is used as the nuclear poison, the optimum dissolution conditions occur in 10 M HNO 3 , 0.04-0.05 M KF, and 0.5-1.0 g/L Gd at 112 to 116 C (boiling). These conditions will result in an estimated Pu metal dissolution rate of ~11-15 mg/min-cm 2 and will result in dissolution times of 36-48 h for standard buttons. The recommended minimum and maximum KF concentrations are 0.03 M and 0.07 M, respectively. The maximum KF concentration is dictated by a potential room-temperature Pu-Gd-F precipitation issue at low Pu concentrations. Testing at 8-10 M HNO 3 , 0.1-0.2 M KF, and 1-2 g/L B demonstrated that ~20-35 mg/min-cm 2 Pu metal dissolution rates can be achieved at 112 to 116 C (boiling). The concentration of B in solution did not have a significant effect on dissolution rate. The data also indicate that lower KF concentrations would yield dissolution rates for B comparable to those observed with Gd at the same HNO 3 concentration and dissolution temperature. The low-temperature Pu precipitation issue associated with the use of Gd does not occur for dissolution with B; however, the B concentration must be maintained below the H 3 BO 3 solubility limit and the KF concentration must not exceed the value where B precipitates as KBF 4. To confirm that the optimal conditions identified by the dissolution rate measurements can be used to dissolve Pu metal up to 6.75 g/L in the presence of representative concentrations of Fe and Gd or B, a series of experiments was performed to demonstrate the flowsheets. In three of the five experiments, the offgas generation rate during the dissolution was measured and samples were analyzed for hydrogen gas (H 2). The use of 10 M HNO 3 containing 0.03-0.05 M KF, 0.5-1.0 g/L Gd, and 1.9 g/L Fe resulted in complete dissolution of the metal in 2.0-3.5 h. When B was used as the neutron poison, 10 M HNO 3 solutions containing 0.05-0.1 M KF, 1.9 g/L Fe, and 1 g/L B resulted in complete dissolution of the metal in 0.75-2.0 h. All experiments were performed using a dissolution temperature of 100 C. No residues were observed following the dissolutions in either the Gd or B system. Dissolut...

SummaryPlutonium (Pu)-containing solutions currently stored in H-Canyon Tanks 12.1 and 16.3 do not meet acceptance criteria for conversion to a mixed oxide fuel. Therefore, the solutions will be neutralized and discarded to the Savannah River Site (SRS) high level waste (HLW) system. Prior to disposal, the addition of gadolinium nitrate (Gd(NO 3 ) 3 ) as a neutron poison is proposed to allow neutralization of quantities of Pu greater than a minimum critical mass per neutralization batch. This disposition strategy was previously studied and used to discard approximately 100 kg of Pu to the HLW system. However, the current solutions have a distinct difference in composition from that material. These current solutions contain slightly enriched uranium (U), 0.8% 235 U, at concentrations equivalent to a 3:1 ratio with Pu. The caustic precipitation behavior of Pu-U-gadolinium (Gd) mixtures had not been previously investigated.Before implementation, the effect of U on the precipitation would have to be evaluated to ensure that a sufficient quantity of Gd is always present in the precipitate slurry to ensure nuclear safety.A sample of the Tank 12.1 solution was obtained to evaluate the precipitation behavior of Pu-U-Gd mixtures during caustic neutralization. Experiments were also performed using surrogate solutions containing 3 g/L U or 3 g/L U with 1 g/L Pu. In each experiment, Gd was added to the acidic solutions as Gd(NO 3 ) 3 prior to neutralization with 50 wt% sodium hydroxide (NaOH). Samples from the Tank 12.1 solution were neutralized in a step-wise manner to a pH of 4.5 and 7 to measure the Pu/Gd ratio in the solids which formed prior to complete neutralization above pH 14. Subsequent experiments were performed in which samples from Tank 12.1 and surrogate solutions were neutralized to 1.2 and 3.6M excess hydroxide (OH -). Samples of the precipitate slurry and supernate were then analyzed to evaluate the effect of U on the precipitation.During the neutralization experiments, the initial solids formed at pH 4.5 in contrast to the previous studies on solutions in which solids were first observed at pH 3. The formation of solids at the higher pH is consistent with the behavior of U solutions. At pH 4.5, 6% of the Gd was found in the solids. This value is essentially the same as the 5% measured at pH 3 in the previous studies. At pH 7, at least 95% of the Gd, U, and Pu were removed from the solutions. Upon complete neutralization, greater than 99% of these elements were found in the precipitated solids. One week after neutralization, analysis of additional liquid and solid samples revealed no significant changes in composition. X-ray diffraction analysis confirmed the formation of sodium diuranate and gadolinium hydroxide in the solids after standing for one week. Scanning

Revision 0 iv EXECUTIVE SUMMARYProcesses for the removal of residual sludge from SRS waste tanks have historically used solutions containing up to 0.9 M oxalic acid to dissolve the remaining material following sludge removal. The selection of this process was based on a comparison of a number of studies performed to evaluate the dissolution of residual sludge. [1] In contrast, the dissolution of the actinide mass, which represents a very small fraction of the waste, has not been extensively studied. The Pu, Np, and Am in the sludge is reported to be present as hydrated and crystalline oxides. [2][3][4] To identify aqueous solutions which have the potential to increase the solubility of the actinides, the alkaline and mildly acidic test solutions shown below were selected as candidates for use in a series of solubility experiments. 8 wt % (0.9 M) oxalic acid 2 wt % (0.22 M) oxalic acid 1.8 wt % (0.2 M) oxalic acid/0.1 M citric acid 0.18 M HNO 3 /0.5 wt % (0.056 M) oxalic acid 0.18 M HNO 3 /0.2 M NaMnO 4 0.18 M HNO 3 10 M NaOH 10 M NaOH/0.2 M NaMnO 4 1 M NaHCO 3 /Na 2 CO 3 at pH 9.5 1 M NaHCO 3 /Na 2 CO 3 /0.2 M NaMnO 4 at pH 9.5 0.05 M DTPA at pH 2-4 0.18 M HNO 3 blank 10.0 M NaOH blank 1 M NaHCO 3 /Na 2 CO 3 at pH 9.5 blank The efficiency of the solutions in solubilizing the actinides was evaluated using a simulated sludge prepared by neutralizing a HNO 3 solution containing Pu, Np, and Am. The hydroxide concentration was adjusted to a 1.2 M excess and the solids were allowed to age for several weeks prior to starting the experiments. The sludge was washed with 0.01 M NaOH to prepare the solids for use. Following the addition of an equal portion of the solids to each test solution, the concentrations of Pu, Np, and Am were measured as a function of time over a 792 h (33 day) period to provide a direct comparison of the efficiency of each solution in solubilizing the actinide elements. Although the composition of the sludge was limited to the hydrated actinide oxides (and did not contain other components of demonstrated importance), the results of the study provides guidance for the selection of solutions which should be evaluated in subsequent tests with a more realistic surrogate sludge and actual tank waste.The results from the solubility experiments showed that the test solutions containing 0.2 M NaMnO 4 were highly effective in solubilizing the Pu. The order of the solutions from the highest to the lowest concentrations achieved was as follows: 0.18 M HNO 3 /0.2 M NaMnO 4 , > 10 M NaOH/0.2 M NaMnO 4 , > 1 M NaHCO 3 /Na 2 CO 3 /0.2 M NaMnO 4 . SRNL-STI-2011-00521Revision 0 v The concentrations after 792 h (33 days) fell in the range of 0.4-2.0E-02 mol/L (1.0-4.8 g/L); although, the Pu in each solution was at or near its maximum concentration after 168 h (1 week). The concentrations of Pu in these solutions were higher than the concentration measured in the baseline 0.9 M oxalic acid solution at the conclusion of the experiments (5.2E-03 mol/L or 1.2 g/L) except for the concentration in NaHCO 3 /Na 2 CO 3 /NaMn...