Mark L. Crowder

Monosodium Titanate in Hydrous Titanium Oxide Spheres for the Removal of Strontium and Key Actinides from Salt Solutions at the Savannah River Site

Abstract Fine powders of monosodium titanate effectively remove strontium and plutonium from alkaline salt supernatant. At the Savannah River Site, larger, porous particles with monosodium titanate were desired for continuous column operations. The internal gelation process was used to make hydrous titanium oxide microspheres with 32 and 50 wt% monosodium titanate. With actual supernatant, the microspheres with 50 wt% monosodium titanate produced average batch distribution coefficients of 35,000 mL/g for plutonium and 99,000 mL/g for strontium. These microspheres were tested using a simulant and a flow rate of 5.3 bed volumes per hour. The plutonium removal dropped from 99% to 94% while the strontium removal remained nearly 100%. The microspheres exhibited good flow performance and no particle degradation.

Performance Testing of the Next-Generation CSSX Solvent with Actual SRS Tank Waste

Efforts are underway to qualify the Next-Generation Solvent for the Caustic Side Solvent Extraction (CSSX) process. Researchers at multiple national laboratories have been involved in this effort. As part of the effort to qualify the solvent extraction system at the Savannah River Site (SRS), Savannah River National Laboratory (SRNL) researchers performed a number of tests at various scales. A series of batch equilibrium, or Extraction-Scrub-Strip (ESS), tests were conducted first. These tests used ∼30 mL of Next-Generation Solvent and either actual SRS tank waste, or waste simulant solutions. The results from these cesium mass transfer tests were used to predict solvent behavior under a number of conditions. For larger-scale testing, twelve stages of 2-cm (diameter) centrifugal contactors were assembled. This rack of contactors is structurally similar to one tested in 2001 during the demonstration of the baseline CSSX process. No issues were encountered during assembly and mechanical testing. A non-radiological test was performed using 35 L of cesium-spiked caustic waste simulant followed by a test with 39 L of actual tank waste. Test results are discussed, particularly those related to the effectiveness of extraction.

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.

Thermal Release of 3He from Tritium Aged LaNi4.25Al0.75 Hydride

Abstract Recently, the demand for He-3 has increased dramatically due to widespread use in nuclear nonproliferation, cryogenic, and medical applications. Essentially all of the world’s supply of He-3 is created by the radiolytic decay of tritium. The Savannah River Site Tritium Facilities (SRS-TF) utilizes LANA.75 in the tritium process to store hydrogen isotopes. The vast majority of He-3 “born” from tritium stored in LANA.75 is trapped in the hydride metal matrix. The SRS-TF has multiple LANA.75 tritium storage beds that have been retired from service with significant quantities of He-3 trapped in the metal. To support He-3 recovery, the Savannah River National Laboratory (SRNL) conducted thermogravimetric analysis coupled with mass spectrometry (TGA-MS) on a tritium aged LANA.75 sample. TGA-MS testing was performed in an argon environment. Prior to testing, the sample was isotopically exchanged with deuterium to reduce residual tritium and passivated with air to alleviate pyrophoric concerns associated with handling the material outside of an inert glovebox. Analyses indicated that gas release from this sample was bimodal, with peaks near 220 and 490°C. The first peak consisted of both He-3 and residual hydrogen isotopes, the second was primarily He-3. The bulk of the gas was released by 600 °C.

Production of Neptunium Oxide from Impure Solutions by Anion Exchange Followed by Oxalate Precipitation

Abstract The stabilization of the neptunium solutions stored at the Savannah River Site (SRS) has generated additional recycle solutions that contain a different mix of impurities. A list of expected purities and the known laboratory and production data for purification from those impurities has been accumulated. An evaluation has been performed of the options for modifying the current process to ensure oxide product purity does not significantly change when recycle solutions are processed. This paper discusses the details of the reduction of the major impurities utilizing both known production quality analyses and laboratory flowsheet development data and proposes modifications to the anion wash volumes to remove higher levels of each impurity.

Measurement of Tritium during Voloxidation of Zircaloy-2 Cladding

Abstract A straightforward method to determine the tritium content of Zircaloy-2 cladding hulls via oxidation of the hulls and capture of the volatilized tritium in liquids has been demonstrated. Hull samples were heated in air inside a thermogravimetric analyzer (TGA). The TGA was rapidly heated to 1000°C to oxidize the hulls and to release absorbed tritium. To capture tritium, the TGA off-gas was bubbled through a series of liquid traps. The concentrations of tritium in bubbler solutions indicated that nearly all of the tritiated water vapor was captured. The average tritium content measured in the hulls was 19% of the amount of tritium produced by the fuel, according to ORIGEN2 isotope generation and depletion calculations. Published experimental data show that there is an initial, nonlinear oxidation rate for Zircaloy-2 followed by a faster, linear rate after “breakaway” of the oxide film and that the linear rate follows an Arrhenius model. This study demonstrates that the linear oxidation rate of Zircaloy samples at 974°C is faster than predicted by the extrapolation of data from lower temperatures.

Dissolution of Plutonium Metal Using Nitric Acid Solutions Containing Potassium Fluoride

Abstract The deinventory and deactivation of the Department of Energys (DOEs) FB-Line facility at the Savannah River Site (SRS) required the disposition of approximately 2000 items from the facilitys vaults. Plutonium (Pu) scraps and residues which do not meet criteria for conversion to a mixed oxide fuel will be dissolved and the solution stored for subsequent disposition. Some of the items scheduled for dissolution are composite materials containing Pu and tantalum (Ta) metals. The preferred approach for handling this material is to dissolve the Pu metal, rinse the Ta metal with water to remove residual acid, and burn the Ta metal. The use of a 4 M nitric acid (HNO3) solution containing 0.2 M potassium fluoride (KF) was initially recommended for the dissolution of approximately 500 g of Pu metal. However, prior to the use of the flowsheet in the SRS facility, a new processing plan was proposed in which the feed to the dissolver could contain up to 1250 g of Pu metal. To evaluate the use of a larger batch size and subsequent issues associated with the precipitation of plutonium-containing solids from the dissolving solution, scaled experiments were performed using Pu metal and samples of the composite material. In the initial experiment, incomplete dissolution of a Pu metal sample demonstrated that a 1250 g batch size was not feasible in the HB-Line dissolver. Approximately 45% of the Pu was solubilized in 4 h. The remaining Pu metal was converted to plutonium oxide (PuO2). Based on this work, the dissolution of 500 g of Pu metal using a 4–6 h cycle time was recommended for the HB-Line facility. Three dissolution experiments were subsequently performed using samples of the Pu/Ta composite material to demonstrate conditions which reduced the risk of precipitating a double fluoride salt containing Pu and K from the dissolving solution. In these experiments, the KF concentration was reduced from 0.2 M to either 0.15 or 0.175 M. With the use of 4 M HNO3 and a reduction in the KF concentration to 0.175 M, the dissolution of 300 g of Pu metal is expected to be essentially complete in 6 h. The dissolution of larger batch sizes would result in the formation of PuO2 solids. Incomplete dissolution of the PuO2 formed from the metal is not a solubility limitation, but can be attributed to a combination of reduced acidity and complexation of fluoride which slows the dissolution kinetics and effectively limits the mass of Pu dissolved.

EVALUATION OF FLOWSHEET CHANGES FOR THE HIGHLY ENRICHED URANIUM BLENDDOWN PROGRAM

Abstract At the Savannah River Site, highly enriched uranium (HEU) is dissolved, purified, and blended with natural uranium to make low enriched uranium solutions sufficiently pure for conversion to power reactor fuel. The process to dissolve and purify aluminum-clad HEU fuel at SRS is well-established. However, for the dissolution and recovery of metal scrap, flowsheet changes were proposed. This study evaluates the proposed changes. Specifically, solvent extraction modeling calculations were performed which indicated that one solvent extraction cycle would be sufficient to purify the metal scrap solution by removing boron, which is added as a neutron poison. In addition, stability constants from the literature and Savannah River National Laboratory corrosion studies were documented to demonstrate that boron complexation of fluoride in nitric acid solutions, at the levels anticipated, is sufficient to prevent excessive corrosion in stainless steel vessels. Downstream from the purification process, limitations on the boron concentration in waste evaporators were recommended to prevent formation of boron-containing solids.