DISSOLUTION OF FB-LINE METAL RESIDUES CONTAINING BERYLLIUM IN H-CANYON

Scrap materials containing plutonium (Pu) metal from FB-Line vaults are currently being dissolved in HB-Line for subsequent disposition through the H-Canyon facility. However, milestone and schedule commitments may require the dissolution of material containing Pu and beryllium (Be) metals in H-Canyon. To support this option, a flowsheet for dissolving Pu and Be metals in H-Canyon was demonstrated using a 4 M nitric acid (HNO{sub 3}) solution containing 0.3 M fluoride (F{sup -}). The F{sup -} was added as calcium fluoride (CaF{sub 2}). The dissolving solution also contained 2.5 g/L boron (B), a nuclear safety contingency for the H-Canyon dissolver, and 3.9 g/L iron (Fe) to represent the dissolution of carbon steel cans. The solution was heated to 90-95 C during the 8 h dissolution cycle. Dissolution of the Be metal appeared to begin as soon as the samples were added to the dissolver. Clear, colorless bubbles generated on the surface were observed and were attributed primarily to the generation of hydrogen (H{sub 2}) gas. The generation of nitrogen dioxide (NO{sub 2}) gas was also evident from the color of the solution. Essentially all of the Pu and Be dissolved during the first hour of the dissolution as the solution was heated to 90-95 C. The amount of residual solids collected following the dissolution was < 2% of the total metal charged to the dissolver. Examination of residual solids by scanning electron microscopy (SEM) showed that the largest dimension of the particles was less than 50 {micro}m with particles of smaller dimensions being more abundant. Energy dispersive spectra from spots on some of the particles showed the solids consisted of a small amount of undissolved material, corrosion products from the glassware, and dried salts from the dissolving solution

DISSOLUTION OF FISSILE MATERIALS CONTAINING TANTALUM METAL

The dissolution of composite materials containing plutonium (Pu) and tantalum (Ta) metals is currently performed in Phase I of the HB-Line facility. The conditions for the present flowsheet are the dissolution of 500 g of Pu metal in the 15 L dissolver using a 4 M nitric acid (HNO{sub 3}) solution containing 0.2 M potassium fluoride (KF) at 95 C for 4-6 h.[1] The Ta metal, which is essentially insoluble in HNO{sub 3}/fluoride solutions, is rinsed with process water to remove residual acid, and then burned to destroy classified information. During the initial dissolution campaign, the total mass of Pu and Ta in the dissolver charge was limited to nominally 300 g. The reduced amount of Pu in the dissolver charge coupled with significant evaporation of solution during processing of several dissolver charges resulted in the precipitation of a fluoride salt contain Pu. Dissolution of the salt required the addition of aluminum nitrate (Al(NO{sub 3}){sub 3}) and a subsequent undesired 4 h heating cycle. As a result of this issue, HB-Line Engineering requested the Savannah River National Laboratory (SRNL) to optimize the dissolution flowsheet to reduce the cycle time, reduce the risk of precipitating solids, and obtain hydrogen (H{sub 2}) generation data at lower fluoride concentrations.[2] Using samples of the Pu/Ta composite material, we performed three experiments to demonstrate the dissolution of the Pu metal using HNO{sub 3} solutions containing 0.15 and 0.175 M KF. When 0.15 M KF was used in the dissolving solution, 95.5% of the Pu in the sample dissolved in approximately 6 h. The undissolved material included a small amount of Pu metal and plutonium oxide (PuO{sub 2}) solids. Complete dissolution of the metal would have likely occurred if the dissolution time had been extended. This assumption is based on the steady increase in the Pu concentration observed during the last several hours of the experiment. We attribute the formation of PuO{sub 2} to the complexation of fluoride by the Pu. The fluoride became unavailable to catalyze the dissolution of PuO{sub 2} as it formed on the surface of the metal. The mass of Pu dissolved is equivalent to the dissolution of 343 g of Pu in the HB-Line dissolvers. In the initial experiment with 0.175 M KF in the solution, we achieved complete dissolution of the Pu in 6 h. The mass of Pu dissolved scales to the dissolution of 358 g of Pu in the HB-Line dissolvers. The second experiment using 0.175 M KF was terminated after approximately 6 h following the dissolution of 92.7% of the Pu in the sample; however, dissolution of additional Pu was severely limited due to the slow dissolution rate observed beyond approximately 4 h. A small amount of PuO{sub 2} was also produced in the solution. The slow rate of dissolution was attributed to the diminishing surface area of the Pu and a reduction in the fluoride activity due to complexation with Pu. Given time (>4 h), the Pu metal may have dissolved using the original solution or a significant portion may have oxidized to PuO{sub 2}. If the metal oxidized to PuO{sub 2}, we expect little of the material would have dissolved due to the fluoride complexation and the low HNO{sub 3} concentration. The mass of Pu dissolved in the second experiment scales to the dissolution of 309 g of Pu in the HB-Line dissolvers. Based on the data from the Pu/Ta dissolution experiments we recommend the use of 4 M HNO{sub 3} containing 0.175 M KF for the dissolution of 300 g of Pu metal in the 15 L HB-Line dissolver. A dissolution temperature of nominally 95 C should allow for essentially complete dissolution of the metal in 6 h. Although the H{sub 2} concentration in the offgas from the experiments was at or below the detection limit of the gas chromatograph (GC) used in these experiments, small concentrations (<3 vol %) of H{sub 2} are typically produced in the offgas during Pu metal dissolutions. Therefore, appropriate controls must be established to address the small H{sub 3} generation rates in accordance with this work and the earlier flowsheet demonstrated for Pu metal.[3

Correction to: Cluster identification, selection, and description in Cluster randomized crossover trials: the PREP-IT trials

An amendment to this paper has been published and can be accessed via the original article

Dissolution of FB-Line Residues Containing Beryllium Metal

Scrap materials containing plutonium (Pu) metal were dissolved at the Savannah River Site (SRS) as part of a program to disposition nuclear materials during the deactivation of the FB-Line facility. Some of these items contained both Pu and beryllium (Be) metal as a composite material. The Pu and Be metals were physically separated to minimize the amount of Be associated with the Pu; however, a dissolution flowsheet was required to dissolve small amounts of Be combined with the Pu metal using a dissolving solution containing nitric acid (HNO{sub 3}) and potassium fluoride (KF). Since the dissolution of Pu metal in HNO{sub 3}/fluoride (F{sup -}) solutions was well understood, the primary focus of the flowsheet development was the dissolution of Be metal. Initially, small-scale experiments were used to measure the dissolution rate of Be metal foils using conditions effective for the dissolution of Pu metal. The experiments demonstrated that the dissolution rate was nearly independent of the HNO{sub 3} concentration over the limited range of investigation and only a moderate to weak function of the F{sup -} concentration. The effect of temperature was more pronounced, significantly increasing the dissolution rate between 40 and 105 C. The offgas analysis from three Be metal foil dissolutions demonstrated that the production of hydrogen (H{sub 2}) was sensitive to the HNO{sub 3} concentration, decreasing by a factor of approximately two when the concentration was increased from 4 to 8 M. In subsequent experiments, complete dissolution of Be samples from a Pu/Be composite material was achieved in a 4 M HNO{sub 3} solution containing 0.1-0.2 M KF. Gas samples collected during each experiment showed that the maximum H{sub 2} generation rate occurred at temperatures below 70-80 C. A Pu metal dissolution experiment was performed using a 4 M HNO{sub 3}/0.1 M KF solution at 80 C to demonstrate flowsheet conditions developed for the dissolution of Be metal. As the reaction progressed, the rate of dissolution slowed. The decrease in rate was attributed to the complexation of F{sup -} by the dissolved Pu. The F{sup -} became unavailable to catalyze the dissolution of plutonium oxide (PuO{sub 2}) formed on the surface of the metal which inhibited the dissolution rate. To compensate for the complexation of F{sup -}, an increase in the concentration to 0.15-0.2 M was recommended. Dissolution of the PuO{sub 2} was addressed by recommending an 8-10 h dissolution time with an increase in the dissolving temperature (to near boiling) during the final 4-6 h to facilitate the digestion of the solids. Dilution of the H{sub 2} concentration below 25% of the lower flammability limit by purging the dissolver with air was also necessary to eliminate the flammability concern

DISTRIBUTION OF LANTHANIDE AND ACTINIDE ELEMENTS BETWEEN BIS-(2-ETHYLHEXYL)PHOSPHORIC ACID AND BUFFERED LACTATE SOLUTIONS CONTAINING SELECTED COMPLEXANTS

With the renewed interest in the closure of the nuclear fuel cycle, the TALSPEAK process is being considered for the separation of Am and Cm from the lanthanide fission products in a next generation reprocessing plant. However, an efficient separation requires tight control of the pH which likely will be difficult to achieve on a large scale. To address this issue, we measured the distribution of lanthanide and actinide elements between aqueous and organic phases in the presence of complexants which were potentially less sensitive to pH control than the diethylenetriaminepentaacetic (DTPA) used in the process. To perform the extractions, a rapid and accurate method was developed for measuring distribution coefficients based on the preparation of lanthanide tracers in the Savannah River National Laboratory neutron activation analysis facility. The complexants tested included aceto-, benzo-, and salicylhydroxamic acids, N,N,N',N'-tetrakis(2-pyridylmethyl)ethylenediamine (TPEN), and ammonium thiocyanate (NH{sub 4}SCN). The hydroxamic acids were the least effective of the complexants tested. The separation factors for TPEN and NH{sub 4}SCN were higher, especially for the heaviest lanthanides in the series; however, no conditions were identified which resulted in separations factors which consistently approached those measured for the use of DTPA

Caustic Precipitation of Plutonium and Uranium with Gadolinium as a Neutron Poison

The caustic precipitation of plutonium (Pu) and uranium (U) from Pu and U containing waste solutions has been investigated to determine whether gadolinium (Gd) could be used as a neutron poison for precipitation with greater than a fissile mass containing both Pu and enriched U. Precipitation experiments were performed using both actual samples and simulant solutions with a range of 2.6-5.16 g/L U and 0-4.3 to 1 U to Pu. Analyses were performed on solutions at intermediate pH to determine the partitioning of elements for accident scenarios. When both Pu and U were present in the solution, precipitation began at pH 4.5 and by pH 7, 99 percent of Pu and U had precipitated. When complete neutralization was achieved at pH greater than 14 with 1.2 M excess OH-, greater than 99 percent of Pu, U, and Gd had precipitated. At pH greater than 14, the particles sizes were larger and the distribution was a single mode. The ratio of hydrogen to fissile atoms in the precipitate was determined after both settling and centrifuging and indicates that sufficient water was associated with the precipitates to provide the needed neutron moderation for Gd to prevent a criticality in solutions containing up to 4.3 to 1 U to Pu and up to 5.16 g/L U

Canyon dissolution of sand, slag, and crucible residues

An alternative to the FB-Line scrap recovery dissolver was desired for the dissolution of sand, slag, and crucible (SS{ampersand}C) residues from the plutonium reduction process due to the potential generation of hydrogen gas concentrations above the lower flammability limit. To address this concern, a flowsheet was developed for the F-Canyon dissolvers. The dissolvers are continually purged with nominally 33 SCFM of air; therefore the generation of flammable gas concentrations should not be a concern. Following removal of crucible fragments, small batches of the remaining sand fines or slag chunks containing less than approximately 350 grams of plutonium can be dissolved using the center insert in each of the four annular dissolver ports to address nuclear criticality safety concerns. Complete dissolution of the sand fines and slag chunks was achieved in laboratory experiments by heating between 75 and 85 degrees Celsius in a 9.3M nitric acid/0.013M (hydrogen) fluoride solution. Under these conditions, the sand and slag samples dissolved between 1 and 3 hours. Complete dissolution of plutonium and calcium fluorides in the slag required adjusting the dissolver solution to 7.5 wt% aluminum nitrate nonahydrate (ANN). Once ANN was added to a dissolver solution, further dissolution of any plutonium oxide (PuO2) in successive charges was not practical due to complexation of the fluoride by aluminum. During the laboratory experiments, well mixed solutions were necessary to achieve rapid dissolution rates. When agitation was not provided, sand fines dissolved very slowly. Measurement of the hydrogen gas generation rate during dissolution of slag samples was used to estimate the amount of metal in the chunks. Depending upon the yield of the reduction, the values ranged between approximately 1 (good yield) and 20% (poor yield). Aging of the slag will reduce the potential for hydrogen generation as calcium metal oxidizes over time. The potential for excessive corrosion in the dissolvers was evaluated using experimental data reported in the literature. Corrosion data at the exact flowsheet conditions were not available; however, the corrosion rate for 304L stainless steel (wrought material) corrosion coupons in 10M nitric acid/0.01M hydrofluoric acid at 95 degrees Celsius was reported as 21 mils per year. If the fluoride in the dissolver is complexed with aluminum, the corrosion rate will decrease to approximately 5 mils per year

PLUTONIUM SOLUBILITY IN SIMULATED SAVANNAH RIVER SITE WASTE SOLUTIONS

To address the accelerated disposition of the supernate and salt portions of Savannah River Site (SRS) high level waste (HLW), solubility experiments were performed to develop a predictive capability for plutonium (Pu) solubility. A statistically designed experiment was used to measure the solubility of Pu in simulated solutions with salt concentrations and temperatures which bounded those observed in SRS HLW solutions. Constituents of the simulated waste solutions included: hydroxide (OH{sup -}), aluminate (Al(OH){sub 4}{sup -}), sulfate (SO{sub 4}{sup 2-}), carbonate (CO{sub 3}{sup 2-}), nitrate (NO{sub 3}{sup -}), and nitrite (NO{sub 2}{sup -}) anions. Each anion was added to the waste solution in the sodium form. The solubilities were measured at 25 and 80 C. Five sets of samples were analyzed over a six month period and a partial sample set was analyzed after nominally fifteen months of equilibration. No discernable time dependence of the measured Pu concentrations was observed except for two salt solutions equilibrated at 80 C which contained OH{sup -} concentrations >5 mol/L. In these solutions, the Pu solubility increased with time. This observation was attributed to the air oxidation of a portion of the Pu from Pu(IV) to the more soluble Pu(V) or Pu(VI) valence states. A data driven approach was subsequently used to develop a modified response surface model for Pu solubility. Solubility data from this study and historical data from the literature were used to fit the model. The model predicted the Pu solubility of the solutions from this study within the 95% confidence interval for individual predictions and the analysis of variance indicated no statistically significant lack of fit. The Savannah River National Laboratory (SRNL) model was compared with predicted values from the Aqueous Electrolyte (AQ) model developed by OLI Systems, Inc. and a solubility prediction equation developed by Delegard and Gallagher for Hanford tank waste. The agreement between measured or values predicted by the SRNL model and values predicted by the OLI AG model was very poor. The much higher predicted concentrations by the OLI AQ model appears to be the result of the model predicting the predominate Pu oxidation state is Pu(V) which is reported as unstable below sodium hydroxide (NaOH) concentrations of 6 M. There was very good agreement between the predicted Pu concentrations using the SRNL model and the model developed by Delegard and Gallagher with the exception of solutions that had very high OH{sup -} (15 M) concentrations. The lower Pu solubilities in these solutions were attributed to the presence of NO{sub 3}{sup -} and NO{sub 2}{sup -} which limit the oxidation of Pu(IV) to Pu(V)