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Electrochemical processing of alkaline nitrate and nitrite wastes

Description: Processing of high-level waste at the Savannah River Plant (SRP) will produce, as a by-product, a low-level, alkaline salt solution containing approximately 17% sodium nitrate and sodium nitrite. This solution will be incorporated into a cement formulation, saltstone, and placed in an engineered landfill. Electrochemical methods have been investigated to decrease the nitrate and nitrite in this solution in order to lower the leaching of nitrate and nitrite from saltstone and to reduce the volume of saltstone. Laboratory experiments have demonstrated the technical feasibility of electrolytically reducing the nitrate and nitrite in a synthetic salt solution similar in composition to that expected to be produced at SRP. Greater than 99% of the sodium nitrate and sodium nitrite can be reduced, producing ammonia, nitrogen, oxygen, and sodium hydroxide. In addition, significant reductions in the volume of saltstone may be realized if the sodium hydroxide produced by electrolysis can be recycled.
Date: January 1, 1986
Creator: Hobbs, D T & Ebra, M A
Partner: UNT Libraries Government Documents Department

DEVELOPMENT OF AN IMPROVED SODIUM TITANATE FOR THE PRETREATMENT OF NUCLEAR WASTE AT THE SAVANNAH RIVER SITE

Description: 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.
Date: January 22, 2008
Creator: Hobbs, D
Partner: UNT Libraries Government Documents Department

FY06 ANNUAL REPORT FOR ENVIRONMENTAL MANAGEMENT SCIENCE PROGRAM PROJECT #95061STRATEGIC DESIGN AND OPTIMIZATION OF INORGANIC SORBENTSFOR CESIUM, STRONTIUM AND ACTINIDES

Description: 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{sup +} 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.
Date: August 10, 2006
Creator: Hobbs, D
Partner: UNT Libraries Government Documents Department

A FAMILY OF PEROXO-TITANATE MATERIALS TAILORED FOR OPTIMAL STRONTIUM ANDACTINIDE SORPTION

Description: 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; {sup 90}Sr 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{sub v}Na{sub w}Ti{sub 2}O{sub 5}-(xH{sub 2}O)[yH{sub z}O{sub 2}] 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.
Date: August 7, 2006
Creator: Hobbs, D
Partner: UNT Libraries Government Documents Department

DEVELOPMENT OF AN IMPROVED SODIUM TITANATE FOR THE PRETREATMENT OF HIGH LEVEL NUCLEAR WASTE AT THE SAVANNAH RIVER SITE

Description: 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 {sup 137}Cs removal. 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.
Date: November 15, 2007
Creator: Hobbs, D
Partner: UNT Libraries Government Documents Department

DEVELOPMENT OF MONOSODIUM TITANATE (MST) PURCHASE SPECIFICATIONS

Description: 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{sup -1} strontium with 0.1 g L{sup -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.
Date: April 30, 2006
Creator: Hobbs, D
Partner: UNT Libraries Government Documents Department

FORM AND AGING OF PLUTONIUM IN SAVANNAH RIVER SITE WASTE TANK 18

Description: 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){sub 4(am)} or hydrous plutonium oxide, PuO{sub 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 Pu{sup 4+} that has been substituted for other metal ions in crystal lattice sites, Pu{sup 4+} occluded within hydrous metal oxide particles and Pu{sup 4+} adsorbed onto the surface of hydrous metal oxide particles. The adsorbed plutonium could include both inner sphere ...
Date: February 24, 2012
Creator: Hobbs, D.
Partner: UNT Libraries Government Documents Department

URANIUM AND PLUTONIUM LOADING ONTO MONOSODIUM TITANATE MST IN TANK 50H

Description: 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.
Date: August 31, 2006
Creator: Hobbs, D
Partner: UNT Libraries Government Documents Department

REVIEW OF EXPERIMENTAL STUDIES INVESTIGATING THE RATE OF STRONTIUM AND ACTINIDE ADSORPTION BY MONOSODIUM TITANATE

Description: 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.
Date: October 1, 2010
Creator: Hobbs, D.
Partner: UNT Libraries Government Documents Department

RECENT ADVANCES IN THE DEVELOPMENT OF THE HYBRID SULFUR PROCESS FOR HYDROGEN PRODUCTION

Description: Thermochemical processes are being developed to provide global-scale quantities of hydrogen. A variant on sulfur-based thermochemical cycles is the Hybrid Sulfur (HyS) Process, which uses a sulfur dioxide depolarized electrolyzer (SDE) to produce the hydrogen. In the HyS Process, sulfur dioxide is oxidized in the presence of water at the electrolyzer anode to produce sulfuric acid and protons. The protons are transported through a cation-exchange membrane electrolyte to the cathode and are reduced to form hydrogen. In the second stage of the process, the sulfuric acid by-product from the electrolyzer is thermally decomposed at high temperature to produce sulfur dioxide and oxygen. The two gases are separated and the sulfur dioxide recycled to the electrolyzer for oxidation. The Savannah River National Laboratory (SRNL) has been exploring a fuel-cell design concept for the SDE using an anolyte feed comprised of concentrated sulfuric acid saturated with sulfur dioxide. The advantages of this design concept include high electrochemical efficiency and small footprint compared to a parallel-plate electrolyzer design. This paper will provide a summary of recent advances in the development of the SDE for the HyS process.
Date: July 22, 2010
Creator: Hobbs, D.
Partner: UNT Libraries Government Documents Department

OPTIMIZED MONOSODIUM TITANATE PHASE II SUPPLEMENTAL TESTING REPORT URANIUM ADSORPTION AND SHELF-LIFE MEASUREMENTS

Description: The DOE Office of Waste Processing recently funded supplemental Phase II testing to further investigate the uranium affinity and shelf-life of modified monosodium titanate (mMST). Testing results confirmed earlier findings that the mMST exhibits much lower affinity for uranium than the baseline monosodium titanate (MST) material. The loading of uranium onto the mMST sample measured more than an order of magnitude lower than that of the MST. This finding indicates that the use of mMST provides a significant advantage over MST in that the mMST will not concentrate enriched uranium to the degree that MST does. The reduced affinity of mMST for uranium allows more operational flexibility in treating waste solutions from a nuclear criticality safety perspective. Testing results also indicate that the mMST exhibits good shelf-life with no measurable loss in plutonium and neptunium removal upon storage of samples at ambient laboratory temperatures for up to 30-months. Testing did exhibit a change in strontium removal performance for both the mMST and MST samples at the most recent testing event. However, the decrease in strontium removal performance proved lower for the mMST than the MST sample. Given these positive findings SRNL recommends continued development of mMST as a replacement for MST in pretreatment facilities at the Savannah River Site (SRS).
Date: January 1, 2008
Creator: Hobbs, D
Partner: UNT Libraries Government Documents Department

PEROXOTITANATE- AND MONOSODIUM METAL-TITANATE COMPOUNDS AS INHIBITORS OF BACTERIAL GROWTH

Description: Sodium titanates are ion-exchange materials that effectively bind a variety of metal ions over a wide pH range. Sodium titanates alone have no known adverse biological effects but metal-exchanged titanates (or metal titanates) can deliver metal ions to mammalian cells to alter cell processes in vitro. In this work, we test a hypothesis that metal-titanate compounds inhibit bacterial growth; demonstration of this principle is one prerequisite to developing metal-based, titanate-delivered antibacterial agents. Focusing initially on oral diseases, we exposed five species of oral bacteria to titanates for 24 h, with or without loading of Au(III), Pd(II), Pt(II), and Pt(IV), and measuring bacterial growth in planktonic assays through increases in optical density. In each experiment, bacterial growth was compared with control cultures of titanates or bacteria alone. We observed no suppression of bacterial growth by the sodium titanates alone, but significant (p < 0.05, two-sided t-tests) suppression was observed with metal-titanate compounds, particularly Au(III)-titanates, but with other metal titanates as well. Growth inhibition ranged from 15 to 100% depending on the metal ion and bacterial species involved. Furthermore, in specific cases, the titanates inhibited bacterial growth 5- to 375-fold versus metal ions alone, suggesting that titanates enhanced metal-bacteria interactions. This work supports further development of metal titanates as a novel class of antibacterials.
Date: January 19, 2011
Creator: Hobbs, D.
Partner: UNT Libraries Government Documents Department

DECONTAMINATION FACTORS AND FILTRATION FLUX IMPACT TO ARP AT REDUCED MST CONCENTRATION

Description: 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%.
Date: June 27, 2012
Creator: Hobbs, D.
Partner: UNT Libraries Government Documents Department

Potential for the Precipitation of Uranium and Plutonium Solids upon Addition of Nitric Acid to Waste Solutions in a Caustic-Side Solvent Extraction Process

Description: Based on the data from those studies, the author led development of predictive solubility models. This report used these models to calculate uranium and plutonium solubility versus solution compositions before and after mixing waste solution with scrub acid.
Date: June 18, 2002
Creator: Hobbs, D.T.
Partner: UNT Libraries Government Documents Department

Impacts of Sodium Oxalate on High-Level Waste Processing at the Savannah River Site

Description: This report documents results from tests conducted to evaluate the impacts of elevated levels of oxalate on operations within the SRS High-Level Waste System. These operations include sludge washing, evaporation, mixing of supernates and wash waters and pretreatment of supernates to remove strontium and actinides by monosodium titanate.
Date: May 2, 2003
Creator: Hobbs, D.T.
Partner: UNT Libraries Government Documents Department

FY03 Annual Report for Environmental Management Science Program - Strategic Design and Optimization of Inorganic Sorbents for Cesium, Strontium, and Actinides

Description: 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.
Date: August 7, 2003
Creator: Hobbs, D.T.
Partner: UNT Libraries Government Documents Department

Strategic Research: In-Tank Generation of Corrosion Inhibitors

Description: Prevention of stress corrosion cracking and pitting corrosion in high-level waste (HLW) tanks requires the periodic addition of corrosion inhibitors, sodium hydroxide and sodium nitrite. These inhibitor ions can be generated electrochemically from the nitrate present in the waste. Thus, a continuously operated electrochemical reactor placed in the top of the tank could generate nitrite and hydroxide. In-tank generation would eliminate the need to continually add process chemicals resulting in cost savings associated with the procurement, pretreatment and disposal of these chemicals. Experiments examined whether both nitrite and hydroxide could be generated simultaneously from a simple waste simulant in a single electrolytic cell. Results indicated that hydroxide, but not nitrite, formed at a rate that would be effective for in-tank generation. Nitrate reduction proceeded beyond the production of nitrite to produce other nitrogen-containing products. We recommend additional testing to identify an optimum cathode material for nitrite production. Alternatively, the in-tank generator may feature a divided cell configuration or dual electrochemical cells in which one cell generates hydroxide and the second cell generates nitrite.
Date: August 21, 2002
Creator: Hobbs, D.T.
Partner: UNT Libraries Government Documents Department

Electrochemical Destruction of Nitrates and Organics FY1995 Progress Report

Description: Production of nuclear materials within the DOE complex has yielded large volumes of high-level waste containing hazardous species such as nitrate, nitrite, chromium, and mercury. Processes being developed for the permanent disposal of these wastes are aimed at separating the bulk of the radioactivity, primarily 137-Cs and 90-Sr, into a small volume for incorporation into a vitrified wasteform, with the remainder being incorporated into a low-level wasteform.
Date: May 30, 1995
Creator: Hobbs, D.T.
Partner: UNT Libraries Government Documents Department

Evaluation of nitrate and nitrite destruction/separation technologies

Description: This report describes and evaluates four types of nitrate and nitrite destruction and separation technologies that could be used to treat the aqueous, alkaline, nitrate-bearing mixed waste that is generated by the In-Tank Precipitation (ITP) process at the Savannah River Site (SRS). The technologies considered in this report include thermal, hydrothermal, chemical, and electrochemical technologies.
Date: August 29, 1997
Creator: Hobbs, D.T.
Partner: UNT Libraries Government Documents Department

Preliminary Report on Monosodium Titanate Adsorption Kinetics

Description: The Salt Disposition Systems Engineering Team identified the adsorption kinetics of actinides and strontium onto monosodium titanate (MST) as a technical risk for several of the processing alternatives selected for additional evaluation in Phase III of their effort. The Flow Sheet Team requested that the Savannah River Technology Center (SRTC) examine the adsorption kinetics of MST for several process alternatives.This study consisted of a statistically designed set of tests to determine the rate of adsorption of strontium, uranium, neptunium and plutonium as a function of temperature, MST concentration, and concentrations of sodium, strontium, uranium, neptunium and plutonium. Additional tests incorporated into the design assess the effects of mixing as well as the influence from the presence of sludge solids and sodium tetraphenylborate.
Date: December 11, 1998
Creator: Hobbs, D. T.
Partner: UNT Libraries Government Documents Department

Characterization of the Tank 41H Saltcake Insoluble Solids

Description: The particle sizes of the insoluble solids from two of the Tank 41H saltcake samples have been determined by scanning electron microscopy. Settling velocities of the solids have been calculated using Stokes Law. Thus, it is concluded that the formation of a solid phase consisting of uranium without significant amounts of Cr, Fe, Mn, and Zn is not possible during the dissolution of saltcake.
Date: October 31, 1994
Creator: Hobbs, D.T.
Partner: UNT Libraries Government Documents Department