Savannah River Site High-level Waste Research Articles

Three-dimensional fluorinated pyrazinium-based cationic organic polymers with high charge density for enhanced ReO4− removal

Ultrafast separation of hazardous and radioactive anions from strongly alkaline nuclear wastewater is urgently needed for both the public health and environment, but also extremely challenging. Herein, we report a 3D fluorinated pyrazinium-based cationic organic polymer (TBPM-Fpz) with high charge density for ultrafast and selective removal of ReO4−. Using fluorinated pyrazine as the linker and rigid tetraphenylmethane as the building block, a high charge density and relatively hydrophobic diamond-like topology skeleton of TBPM-Fpz was constructed, which is beneficial to improve the affinity for ReO4− and promote thorough exposure of active sites. Batch sorption experiments revealed that TBPM-Fpz is a rare example that can integrate ultrafast sorption kinetics (equilibrium within 3 min), high capacity (833.01 mg g−1), and excellent selectivity towards ReO4− (92.34% in 1000 times excess of SO42−) in one adsorbent. In addition, TBPM-Fpz achieve high ReO4− removal efficiency in simulated Savannah River Site High-Level Waste stream (95.32%, solid-to-liquid ratio: 40:1) and simulated Hanford Low Activity Waste wastewater (89.76%, solid-to-liquid ratio: 1:1). Experimental and computational results indicate that the superior performance is attributed to fluorination that increases charge density and decrease the steric hindrance. This study shows the enormous potential of fluorinated cationic organic polymer for enhanced ReO4− removal from strongly alkaline wastewater, providing valuable insights into designing high-performance cationic organic polymers for alkaline nuclear waste treatment.

CTAB assisted evaporation-induced self-assembly to construct imidazolium-based hierarchical porous covalent organic polymers for ReO4-/TcO4- removal

Evaporation-induced self-assembly method (EISA) was a facile and reliable method to synthesize porous materials. Herein, we report a kind of hierarchical porous ionic liquid covalent organic polymers (HPnDNH2) under cetyltrimethylammonium bromide (CTAB) assisted by EISA for ReO4-/TcO4- removal. Unlike covalent organic frameworks (COFs), which usually needed to be prepared in a closed environment or with a long reaction time, HPnDNH2 in this study was prepared within 1 h in an open environment. It was worth noting that CTAB not only served as a soft template for forming pore, but also induced ordered structure, which was verified by SEM, TEM, and Gas sorption. Benefit from its hierarchical pore structure, HPnDNH2 exhibited higher adsorption capacity (690.0 mg g−1 for HP1DNH2 and 808.7 mg g−1 for HP1.5DNH2) and faster kinetics for ReO4-/TcO4- than 1DNH2 (without employing CTAB). Additionally, the material used to remove TcO4- from alkaline nuclear waste was seldom reported, because combining features of alkali resistance and high uptake selectivity was not easy to achieve. In this study, in the case of HP1DNH2, it displayed outstanding adsorption efficiency toward aqueous ReO4-/TcO4- in 1 mol L−1 NaOH solution (92%) and simulated Savannah River Site High-level waste (SRS HLW) melter recycle stream (98%), which could be a potentially excellent nuclear waste adsorbing material.

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).

Survival in Nuclear Waste, Extreme Resistance, and Potential Applications Gleaned from the Genome Sequence of Kineococcus radiotolerans SRS30216

Kineococcus radiotolerans SRS30216 was isolated from a high-level radioactive environment at the Savannah River Site (SRS) and exhibits γ-radiation resistance approaching that of Deinococcus radiodurans. The genome was sequenced by the U.S. Department of Energy's Joint Genome Institute which suggested the existence of three replicons, a 4.76 Mb linear chromosome, a 0.18 Mb linear plasmid, and a 12.92 Kb circular plasmid. Southern hybridization confirmed that the chromosome is linear. The K. radiotolerans genome sequence was examined to learn about the physiology of the organism with regard to ionizing radiation resistance, the potential for bioremediation of nuclear waste, and the dimorphic life cycle. K. radiotolerans may have a unique genetic toolbox for radiation protection as it lacks many of the genes known to confer radiation resistance in D. radiodurans. Additionally, genes involved in the detoxification of reactive oxygen species and the excision repair pathway are overrepresented. K. radiotolerans appears to lack degradation pathways for pervasive soil and groundwater pollutants. However, it can respire on two organic acids found in SRS high-level nuclear waste, formate and oxalate, which promote the survival of cells during prolonged periods of starvation. The dimorphic life cycle involves the production of motile zoospores. The flagellar biosynthesis genes are located on a motility island, though its regulation could not be fully discerned. These results highlight the remarkable ability of K radiotolerans to withstand environmental extremes and suggest that in situ bioremediation of organic complexants from high level radioactive waste may be feasible.

Appraisal of Mixed Amorphous Manganese Oxide/Titanium Oxide Sorbents for the Removal of Strontium-90 from Solutions, with Special Reference to Savannah River Site and Chernobyl Radioactive Waste Simulants

Hydrous, amorphous manganese oxide/titanium oxide composites (AMTO) obtained via template synthesis using alkali metal ions as templates have been characterized. Their applicability in remediation technologies has been tested, specifically for 90Sr removal from Savannah River Site high-level waste supernatants (SHLW) and Chernobyl 4th Block Shelter water (CSW) simulants. Sorption isotherms were used to establish the efficiency of the templating process, while the proton-exchange capacity of AMTO was determined to show that the solids were really protonated. Use of the 90Sr isotope showed that there was little dependence of the distribution coefficients for AMTO on pH, and at pH < 7, i.e., in neutral and acidic media, AMTO had a much greater affinity for strontium than commercial sorbents. Tests with SHLW and CSW simulants revealed that AMTO is capable of removing more than 99% of 90Sr from such wastes. Hence, it could be useful for solving decontamination problems.

Sorption Modeling of Strontium, Plutonium, Uranium, and Neptunium Adsorption on Monosodium Titanate

:We examined the ability of various equilibrium isotherms to replicate the available data for the adsorption of strontium (Sr), plutonium (Pu), uranium (U), and neptunium (Np) on monosodium titanate (MST) during the treatment of simulated and actual Savannah River Site high-level waste. The data come from numerous experimental studies conducted between 1999 and 2002. The analysis considered 29 isotherm models from the literature. As part of this study, we developed a general method for selecting the best isotherm equation. The selection criteria for rating the isotherm equations considered the relative error in predicting the experimental data, the complexity of the mathematical expressions, the thermodynamic validity of the expressions, and statistical significance for the expressions.The Fowler Guggenheim-Jovanovic Freundlich (FG-JF), the Fowler Guggenheim-Langmuir Freundlich (FG-LF) and the Dubinin-Astashov (DA) isotherms each reliably predicted the actinide and Sr adsorption on MST. The first two models describe the adsorption process by single layer formation and lateral interactions between adsorbed sorbates, while the DA model assumes volume filling of micropores (by osmotic pressure difference). These two mechanisms include mutually exclusive assumptions. However, we cannot determine which model best represents the various adsorption mechanisms on MST. Based on our analysis, the DA model predicted the data well. The DA model assumes that an initial sorption layer forms after which networking begins in the pore spaces, filling the volume by a second mechanism. If this mechanism occurs in MST, as the experimental data suggest, then we expect all the empty and closed spaces of MST to contain actinides and Sr when saturated. Prior microstructure analyses determined that the MST surface is best described as heterogeneous (i.e., a semicrystalline outer layer on an amorphous core) or composite material for adsorption. Therefore, we expect the empty spaces (of nanometer size) between the crystalline units in the fibrous material to provide sorption area for the actinides and Sr. Additional conclusions from this study follow.Since each of the three models work reliably, we recommend use of the computationally simplest model as the primary tool until future work can differentiate between the two mechanisms. The DA model possesses a simpler mathematical form with fewer parameters and operations.The experimental data for actual and simulated wastes generally showed consistent agreement. However, the data sets do include considerable variance from a number of causes including the following:• The Pu sorption data appear most consistent (e.g., between actual and simulated waste) and most easily predicted. Since Pu removal efficiency proves most important for the process design efforts, this consistency of the data proves especially beneficial.• Extremely high mass loadings of U on MST result in multilayer sorption behavior and divergence from classical single monolayer isotherm forms. Prior X-ray studies demonstrate that U begins to network, or form dimers, which agrees with this interpretation. This U behavior also shows a complex interaction, and a direct correlation, with sorption data for the other radionuclides. We believe these data suggest nucléation (e.g., precipitation) of the actinides in the micropore space for both Np and Pu. For Sr, the high U loadings appear to inhibit the sorption of Sr.• Nearly all the solutions contained U as the radionuclide with the highest mass concentration. These data show the widest variance.• The composite data set indicates a notable variance in sorption for different batches of MST. The sorption of Sr with different batches of MST shows the largest variance among the four radionuclides for different batches of MST. This variance remains a relatively unexplored aspect of the process design.• Similarly, the experimental data included a wide variety of solution compositions. As such, the mathematical expressions implicitly account for variances in solution chemistry typical of that anticipated within the Salt Waste Processing Facility and Actinide Removal Process. The reader must consider the ranges of these concentrations when applying the expressions.• Increasing temperature decreases Sr, Pu, and, to a lesser extent, U sorption on MST. The opposite effect occurs with Np. This temperature variance further suggests a nucléation behavior for Np.• Nearly all the data used in developing the sorption models came from experiments using solutions with all the principle radionuclides of interest present simultaneously. We modeled the data without invoking competition between the actinides and Sr despite the large concentrations of both U and Np. Since the model does not explicitly invoke competition, the optimized parameters implicitly carry the impact of interaction within the concentration ranges of the original data. Hence, extrapolation of the models to concentrations markedly outside those ranges may result in poorer predictive ability.

Investigation of Plutonium and Uranium Precipitation Behavior with Gadolinium as a Neutron Poison

:The caustic precipitation of plutonium (Pu)-containing solutions has been investigated to determine whether the presence of 1 :3 Pu : uranium (U) in solutions stored in the H-Canyon Facility at the U.S. Department of Energy’s Savannah River Site (SRS) would adversely impact the use of gadolinium nitrate [GcHNO3)3] as a neutron poison. In the past, this disposition strategy has been successfully used to discard solutions containing approximately 100 kg of Pu to the SRS high-level waste (HLW) system. In the current experiments, gadolinium [as Gd(NO3)3] was added to samples of a 1 : 3 Pu : U solution, a surrogate 3 g/L U solution and a surrogate 3 g/L U with 1 g/L Pu solution. A series of experiments was then performed to observe and characterize the precipitate at selected pH values. Solids formed at pH 4.5 and were found to contain at least 50% of the U and 94% of the Pu but only 6% of the Gd. As the pH of the solution increased [e.g., pH >14 with 1.2 or 3.6M sodium hydroxide (NaOH) excess], the precipitate contained greater than 99% of the Pu, U, and Gd. After the pH >14 systems were undisturbed for one week, no significant changes were found in the composition of the solid or supernale for each sample. The solids were characterized by X-ray diffraction (XRD) which found sodium diuranate (Na2U2O7) and gadolinium hydroxide [Gd(OH)3] at pH > 14. Thermal gravimetric analysis (TGA) indicated sufficient water molecules were present in the solids to thermalize the neutrons, a requirement for the use of Gd as a neutron poison. Scanning electron microscopy (SEM) was also performed, and the accompanying back-scattering electron analysis (BSE) found Pu, U, and Gd compounds in all PH >14 precipitate samples. The rheological properties o! the slurries at pH >14 were also investigated by performing precipitate settling rate studies and measuring the viscosity and density of the materials. Based on the results of these experiments, poisoning the Pu-U solutions with Gil and subsequent neutralization is a viable process for discarding the Pu to the SRS HLW system.

Cesium Removal from Savannah River Site Radioactive Waste using the Caustic-Side Solvent Extraction (CSSX) Process

:Researchers at the Savannah River National Laboratory successfully demonstrated the Caustic-Side Solvent Extraction (CSSX) process flow sheet using a 33-stage, 2-cm centrifugal contactor apparatus in two 24 h tests using actual high-level waste. The CSSX process for removal of cesium from alkaline solutions is the reference process for decontamination of high-level waste (HLW) at the Savannah River Site (SRS). The solvent consists of a calix[4]arene-crown-6 extractant, an alkyl-phenoxy alcohol modifier, and trioctylamine dissolved in an inert hydrocarbon matrix. Previously, we demonstrated the solvent extraction process with actual SRS HLW supernatant solution using a nonoptimized solvent formulation. Following that test, the solvent system was optimized to enhance extractant solubility in the diluent by increasing the modifier concentration. We now report results of two tests with the new and optimized solvent. The first test used a composite of supernatant solutions from two waste tanks, and the second test used a solution derived from dissolved salt cake. Test results showed that the CSSX process using the optimized solvent reduces 137Cs in HLW supernatant solutions to concentrations below the waste acceptance criterion of 45nCi/g for disposal as low-level waste (called “saltstone”). Waste decontamination factors as high as three million were achieved during testing. Test durations exceeded 24 h of uninterrupted operation and demonstrated hydraulic stability of the contactor array while operating with the optimized solvent. Carryover of organic solvent in aqueous streams (and aqueous in organic streams) was found to be less Ilian 1 %. The concentration factor (i.e., the ratio of the cesium concentration in the strip raftinate to the concentration in the waste) averaged approximately 13 during both tests, slightly below the process requirement of 15. Uncertainties in process flow-rate measurement and control prevented the test from achieving the target of 15.

Control of Radioactive Waste Glass Melters: I, Preliminary General Limits at Savannah River

A ceramic, Joule‐heated, slurry‐fed melter will be used to vitrify Savannah River Site high‐level waste at the Defense Waste Processing Facility (DWPF) in Aiken, SC. The entire vitrification process must be performed remotely and the borosilicate host glass must accommodate a variety of waste compositions. The control bases for such a melter are unique and complex. DWPF operations are tightly constrained by processing parameters and by product specifications. Part I discusses the melter control considerations which have been identified to facilitate compliance with these constraints. These considerations include melter temperature, glass composition, product durability, waste loading limits, glass redox control, and glass cooling requirements.