[en] The basic science goal in this project identifies structure/affinity relationships for selected radionuclides and existing sorbents. The task will apply this knowledge to the design and synthesis of new sorbents that will exhibit increased affinity for cesium, strontium and actinide separations. The target problem focuses on the treatment of high-level nuclear wastes. The general approach can likewise be applied to nonradioactive separations. During the fifth year of the project our studies focused along the following paths: (1) determination of Cs⁺ ion exchange mechanism in sodium titanium silicates with sitinikite topology and the influence of crystallinity on ion exchange, (2) synthesis and characterization of novel peroxo-titanate materials for strontium and actinide separations, and (3) further refinements in computational models for the CST and polyoxoniobate materials

[en] Tank Farm and Closure Engineering is evaluating changes to the Actinide Removal Process facility operations to decrease the MST concentration from 0.4 g/L to 0.2 g/L and the contact time from 12 hours to between 6 and 8 hours. For this evaluation, SRNL reviewed previous datasets investigating the performance of MST at 0.2 g/L in salt solutions ranging from 4.5 to 7.5 M in sodium concentration. In general, reducing the MST concentration from 0.4 to 0.2 g/L and increasing the ionic strength from 4.5 to 7.5 M in sodium concentration will decrease the measured decontamination factors for plutonium, neptunium, uranium and strontium. The decontamination factors as well as single standard deviation values for each sorbate are reported. These values are applicable within the sorbate and sodium concentrations used in the experimental measurements. Decreasing the MST concentration in the ARP from 0.4 g/L to 0.2 g/L will produce an increase in the filter flux, and could lead to longer operating times between filter cleaning. The increase in flux is a function of a number of operating parameters, and is difficult to quantify. However, it is estimated that the reduction in MST could result in a reduction of filtration time of up to 20%.

[en] A possible disposition pathway for the residue from the abandoned In-Tank Precipitation (ITP) sends the material from Tank 48H in increments to Saltstone via aggregation in Tank 50H. After entering Tank 50H, the amount of fissile material sorbed on MST may increase as a result of contacting waste solutions with dissolved uranium and plutonium. SRNL recommends that nuclear criticality safety evaluations use uranium and plutonium loadings onto MST of 14.0 ± 1.04 weight percent (wt %) for uranium and 2.79 ± 0.197 wt % for plutonium given the assumed streams defined in this report. These values derive from recently measured for conditions relevant to the Actinide Removal Process (ARP) and serve as conservative upper bounds for uranium and plutonium loadings during the proposed transfers of MST from Tank 48H into Tank 50H

[en] Achieving global optimization of inorganic sorbent efficacy, as well as tailoring sorbent specificity for target sorbates would facilitate increased wide-spread use of these materials in applications such as producing potable water or nuclear waste treatment. Sodium titanates have long been known as sorbents for radionuclides; ⁹⁰Sr and transuranic elements in particular. We have developed a related class of materials, which we refer to as peroxo-titanates: these are sodium titanates or hydrous titanates synthesized in the presence of or treated post-synthesis with hydrogen peroxide. Peroxo-titanates show remarkable and universal improved sorption behavior with respect to separation of actinides and strontium from Savannah River Site (SRS) nuclear waste simulants. Enhancement in sorption kinetics can potentially result in as much as an order of magnitude increase in batch processing throughput. Peroxo-titanates have been produced by three different synthetic routes: post-synthesis peroxide-treatment of a commercially produced monosodium titanate, an aqueous-peroxide synthetic route, and an isopropanol-peroxide synthetic route. The peroxo-titanate materials are characteristically yellow to orange, indicating the presence of protonated or hydrated Ti-peroxo species; and the chemical formula can be generally written as H_vNa_wTi₂O₅-(xH₂O)[yH_zO₂] where (v+w) = 2, z = 0-2, and total volatile species accounts for 25-50 wt % of the solid. Further enhancement of sorption performance is achieved by processing, storing and utilizing the peroxo-titanate as an aqueous slurry rather than a dry powder, and post-synthesis acidification. All three synthesis modifications; addition of hydrogen peroxide, use of a slurry form and acidification can be applied more broadly to the optimization of other metal oxide sorbents and other ion separations processes

[en] Savannah River National Laboratory (SRNL) evaluated the previous monosodium titanate (MST) purchase specifications for particle size and strontium decontamination factor. Based on the measured particle size and filtration performance characteristics of several MST samples with simulated waste solutions and various filter membranes we recommend changing the particle size specification as follows. The recommended specification varies with the size and manufacturer of the filter membrane as shown below. We recommend that future batches of MST received at SRS be tested for particle size and filtration performance. This will increase the available database and provide increased confidence that particle size parameters are an accurate prediction of filtration performance. Testing demonstrated the feasibility of a non-radiochemical method for evaluating strontium removal performance of MST samples. Using this analytical methodology we recommend that the purchase specification include the requirement that the MST exhibits a strontium DF factor of >1.79 upon contact with a simulated waste solution with composition as reported for simulated waste solution SWS-7-2005-1 in Table 1 and containing 5.2 to 5.7 mg L^-1 strontium with 0.1 g L^-1 of the MST. We also recommend performing additional tests with these simulants and MST samples and, if available, new MST samples, to determine the reproducibility and increase the available database for the measurements by the ICP-ES instrument. These measurements will provide increased confidence that the non-radiological method provides a reliable method for evaluating the strontium and actinide removal performance for MST samples

[en] A number of laboratory studies have been conducted to determine the influence of mixing and mixing intensity, solution ionic strength, initial sorbate concentrations, temperature, and monosodium titanate (MST) concentration on the rates of sorbate removal by MST in high-level nuclear waste solutions. Of these parameters, initial sorbate concentrations, ionic strength, and MST concentration have the greater impact on sorbate removal rates. The lack of a significant influence of mixing and mixing intensity on sorbate removal rates indicates that bulk solution transport is not the rate controlling step in the removal of strontium and actinides over the range of conditions and laboratory-scales investigated. However, bulk solution transport may be a significant parameter upon use of MST in a 1.3 million-gallon waste tank such as that planned for the Small Column Ion Exchange (SCIX) program. Thus, Savannah River National Laboratory (SRNL) recommends completing the experiments in progress to determine if mixing intensity influences sorption rates under conditions appropriate for this program. Adsorption models have been developed from these experimental studies that allow prediction of strontium (Sr), plutonium (Pu), neptunium (Np) and uranium (U) concentrations as a function of contact time with MST. Fairly good agreement has been observed between the predicted and measured sorbate concentrations in the laboratory-scale experiments.

[en] High-level nuclear waste produced from fuel reprocessing operations at the Savannah River Site (SRS) requires pretreatment to remove Cs-137, Sr-90 and alpha-emitting radionuclides (i.e., actinides) prior to disposal onsite as low level waste. Separation processes planned at SRS include sorption of Sr-90 and alpha-emitting radionuclides onto monosodium titanate (MST) and caustic side solvent extraction, for Cs-137 removal. The MST and separated Cs-137 will be encapsulated into a borosilicate glass waste form for eventual entombment at the federal repository. The predominant alpha-emitting radionuclides in the highly alkaline waste solutions include plutonium isotopes Pu-238, Pu-239 and Pu-240. This paper describes recent results to produce an improved sodium titanate material that exhibits increased removal kinetics and capacity for Sr-90 and alpha-emitting radionuclides compared to the baseline MST material

[en] This report provides a summary of the effects of aging on and the expected forms of plutonium in Tank 18 waste residues. The findings are based on available information on the operational history of Tank 18, reported analytical results for samples taken from Tank 18, and the available scientific literature for plutonium under alkaline conditions. These findings should apply in general to residues in other waste tanks. However, the operational history of other waste tanks should be evaluated for specific conditions and unique operations (e.g., acid cleaning with oxalic acid) that could alter the form of plutonium in heel residues. Based on the operational history of other tanks, characterization of samples from the heel residues in those tanks would be appropriate to confirm the form of plutonium. During the operational period and continuing with the residual heel removal periods, Pu(IV) is the dominant oxidation state of the plutonium. Small fractions of Pu(V) and Pu(VI) could be present as the result of the presence of water and the result of reactions with oxygen in air and products from the radiolysis of water. However, the presence of Pu(V) would be transitory as it is not stable at the dilute alkaline conditions that currently exists in Tank 18. Most of the plutonium that enters Savannah River Site (SRS) high-level waste (HLW) tanks is freshly precipitated as amorphous plutonium hydroxide, Pu(OH)_4(am) or hydrous plutonium oxide, PuO_2(am,hyd) and coprecipitated within a mixture of hydrous metal oxide phases containing metals such as iron, aluminum, manganese and uranium. The coprecipitated plutonium would include Pu⁴⁺ that has been substituted for other metal ions in crystal lattice sites, Pu⁴⁺ occluded within hydrous metal oxide particles and Pu⁴⁺ adsorbed onto the surface of hydrous metal oxide particles. The adsorbed plutonium could include both inner sphere coordination and outer sphere coordination of the plutonium. PuO_2(am,hyd) is also likely to be present in deposits and scales that have formed on the steel surfaces of the tank. Over the operational period and after closure of Tank 18, Ostwald ripening has and will continue to transform PuO_2(am,hyd) to a more crystalline form of plutonium dioxide, PuO_2(c). After bulk waste removal and heel retrieval operations, the free hydroxide concentration decreased and the carbonate concentration in the free liquid and solids increased. Consequently, a portion of the PuO_2(am,hyd) has likely been converted to a hydroxy-carbonate complex such as Pu(OH)₂(CO₃)_(s). or PuO(CO₃) · xH₂O_(am). Like PuO_2(am,hyd), Ostwald ripening of Pu(OH)₂(CO₃)_(s) or PuO(CO₃) · xH₂O_(am) would be expected to occur to produce a more crystalline form of the plutonium carbonate complex. Due to the high alkalinity and low carbonate concentration in the grout formulation, it is expected that upon interaction with the grout, the plutonium carbonate complexes will transform back into plutonium hydroxide. Although crystalline plutonium dioxide is the more stable thermodynamic state of Pu(IV), the low temperature and high water content of the waste during the operating and heel removal periods in Tank 18 have limited the transformation of the plutonium into crystalline plutonium dioxide. During the tank closure period of thousands of years, transformation of the plutonium into a more crystalline plutonium dioxide form would be expected. However, the continuing presence of water, reaction with water radiolysis products, and low temperatures will limit the transformation, and will likely maintain an amorphous Pu(OH)₄ or PuO_2(am,hyd) form on the surface of any crystalline plutonium dioxide produced after tank closure. X-ray Absorption Spectroscopic (XAS) measurements of Tank 18 residues are recommended to confirm coordination environments of the plutonium. If the presence of PuO(CO₃)_(am,hyd) is confirmed by XAS, it is recommended that experiments be conducted to determine if plutonium carbonates are transformed back into PuO_2(am,hyd) upon contact with grout.

[en] Electrochemical treatment processes are being evaluated and developed for the destruction of organic compounds and nitrates/nitrites and the removal of other hazardous species from liquid wastes stored throughout the DOE complex. This activity consists of five major tasks: (1) evaluation of different electrochemical reactors for the destruction and removal of hazardous waste components, (2) development and validation of engineering process models, (3) radioactive laboratory-scale tests, (4) demonstration of the technology in an engineering-scale size reactor, and (5) analysis and evaluation of testing data. The development program team is comprised of individuals from federal, academic, and private industry. Work is being carried out in DOE, academic, and private industrial laboratories

[en] The basic science goal in this project identifies structure/affinity relationships for selected radionuclides and existing sorbents. The task will apply this knowledge to the design and synthesis of new sorbents that will exhibit increased cesium, strontium and actinide removal. The target problem focuses on the treatment of high-level nuclear wastes. The general approach can likewise be applied to non-radioactive separations