Tracy S. Rudisill

Thermal Stability of Formohydroxamic Acid

The thermal stability of formohydroxamic acid (FHA) was evaluated to address the potential for exothermic decomposition during storage and its use in the uranium extraction process. Accelerating rate calorimetry showed rapid decomposition at a temperature above 65°C; although, the rate of pressure rise was greater than two orders of magnitude less than the lower bound for materials which have no explosive properties with respect to transportation. An FHA solution in nitric acid did not reach runaway conditions until 150°C. Water appeared to temper the FHA decomposition. Analysis by differential scanning calorimetry showed that FHA melted at 67°C and thermally decomposed at 90°C with an enthalpy of −1924 J/g. The energics of the FHA thermal decomposition are comparable to those measured for aqueous solutions of hydroxylamine nitrate. Solid FHA should be stored in a location where the temperature does not exceed 20–25°C. As a best practice, the solid material should be stored in a climate-controlled environment such as a refrigerator or freezer. FHA solutions in water are not susceptible to degradation by acid hydrolysis and are the preferred way to handle FHA prior to use.

Dissolution of Fissile Materials Containing Plutonium and Beryllium Metals

Abstract 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 (HNO3) and potassium fluoride (KF). Since the dissolution of Pu metal in HNO3/fluoride (F−) 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 HNO3 concentration over the limited range of investigation and only a moderate to weak function of the F− 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 (H2) was sensitive to the HNO3 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 HNO3 solution containing 0.1–0.2 M KF. Gas samples collected during each experiment showed that the maximum H2 generation rate occurred at temperatures below 70–80°C. A Pu metal dissolution experiment was performed using a 4 M HNO3/0.1 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− by the dissolved Pu. The F− became unavailable to catalyze the dissolution of plutonium oxide (PuO2) formed on the surface of the metal which inhibited the dissolution rate. To compensate for the complexation of F−, an increase in the concentration to 0.15–0.2 M was recommended. Dissolution of the PuO2 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 H2 concentration below 25% of the lower flammability limit by purging the dissolver with air was also necessary to eliminate the flammability concern.

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−), aluminate , sulfate , carbonate , nitrate , and nitrite 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− 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− (15 M) concentrations. The lower Pu solubilities in these solutions were attributed to the presence of and which limit the oxidation of Pu(IV) to Pu(V).

DISSOLUTION AND OFF-GAS RESULTS FOR U-AL AND AL ALLOYS REPRESENTATIVE OF MURR-TYPE USED NUCLEAR FUELS

Dissolution experiments were performed to build on previous work and allow for modifications of the UNF dissolution flowsheet. The targeted UNF for dissolution at the Savannah River Site (SRS) are fuels similar to the University of Missouri Research Reactor (MURR) fuel. The UAlx-Al fuels are dissolved with HNO3 and Hg catalyst. The experiments initially performed used Al-1100 alloy coupons. The Al-1100 coupons were considered a representative surrogate (based on the fuel bundle and assembly material) that provided an upper bound on the generation of flammable gas during the dissolution process. Hydrogen generation profiles from Al-1100 dissolutions differed from previous work used for the technical basis for UNF dissolution performed with U-Al alloy. To resolve differences in the previous data, additional experiments were performed with U-Al alloys. Results from the initial phase of experimental work are presented including discussion of flammability calculations representative of H-Canyon fuel dissolutions.

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 (NH4SCN). The hydroxamic acids were the least effective of the complexants tested. The separation factors for TPEN and NH4SCN 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 and uranium from Pu- and U-containing waste solutions has been investigated to determine whether gadolinium 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 process solution samples and simulant solutions with a range of 2.6 to 5.16 g/l U and 0 to 4.3:1 U: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% of Pu and U had precipitated. When complete neutralization was achieved at pH >14 with 1.2 M excess OH-, greater than 99% of Pu, U, and Gd had precipitated. At pH >14, the particle 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 4.3:1 U:Pu and up to 5.16 g/l U.

Demonstration of the Feasibility of Recovering Americium and Curium Isotopes from a Lanthanide Borosilicate Glass

The intent of the study was to define an acceptable glass composition region where these properties met predefined acceptance criteria.

Low temperature dissolution flowsheet for Pu metal

ABSTRACT The Savannah River National Laboratory was requested to develop a Pu metal dissolution flowsheet at two reduced temperature ranges for implementation in the Savannah River Site H-Canyon facility. The dissolution and H2 generation rates during Pu metal dissolution were investigated using a dissolving solution at ambient temperature (20–30°C) and for an intermediate temperature of 50–60°C. The Pu metal dissolution rate measured at 57°C was approximately 20 times slower than at boiling (112–116°C). The dissolution rate at ambient temperature (24°C) was approximately 80 times slower than the dissolution rate at boiling. Hydrogen concentrations were less than detectable (<0.1 vol%).

Characterization of undissolved solids from the dissolution of North Anna reactor fuel

ABSTRACT Samples of undissolved solids (UDS) from the dissolution of North Anna reactor fuel were characterized to investigate the effects of using air or oxygen as the oxidant during tritium removal. The UDS composition data also support the development of a waste form for disposal. There was no discernible effect of the oxidant used during the tritium removal process or the size fraction on the UDS composition. Scanning electron microscopy (SEM) and energy dispersive (x-ray) spectroscopy were used to estimate the oxygen content of the UDS and it was found to be potentially significant, on the order of 30% by mass and 80% by atom.

Dissolution of used nuclear fuel using recycled nitric acid

ABSTRACT An evaluation was performed on the feasibility of using HB-Line anion exchange column waste streams from Alternate Feedstock 2 (AFS-2) processing for the dissolver solution for used nuclear fuel (UNF) processing. The targeted UNF for dissolution using recycled solution are fuels similar to the University of Missouri Research Reactor (MURR) fuel. The objectives of this experimental program were to validate the feasibility of using impure dissolver solutions with the MURR dissolution flowsheet to verify they would not significantly affect dissolution of the UNF in a detrimental manner.