Skylab 4 visual observations project report

The Skylab 4 Visual Observations Project was undertaken to determine the ways in which man can contribute to future earth-orbital observational programs. The premission training consisted of 17 hours of lectures by scientists representing 16 disciplines and provided the crewmen information on observational and photographic procedures and the scientific significance of this information. During the Skylab 4 mission, more than 850 observations and 2000 photographs with the 70-millimeter Hasselblad and 35-millimeter Nikon cameras were obtained for many investigative areas. Preliminary results of the project indicate that man can obtain new and unique information to support satellite earth-survey programs because of his inherent capability to make selective observations, to integrate the information, and to record the data by describing and photographing the observational sites

The Role of Liquid Waste Pretreatment Technologies in Solving the Doe Clean-Up Mission

The objective of this report is to describe the pretreatment solutions that allow treatment to be tailored to specific wastes, processing ahead of the completion schedules for the main treatment facilities, and reduction of technical risks associated with future processing schedules. Wastes stored at Hanford and Savannah River offer challenging scientific and engineering tasks. At both sites, space limitations confound the ability to effectively retrieve and treat the wastes. Additionally, the radiation dose to the worker operating and maintaining the radiochemical plants has a large role in establishing the desired radioactivity removal. However, the regulatory requirements to treat supernatant and saltcake tank wastes differ at the two sites. Hanford must treat and remove radioactivity from the tanks based on the TriParty Agreement and Waste Incidental to Reprocessing (WIR) documentation. These authorizing documents do not specify treatment technologies; rather, they specify endstate conditions. Dissimilarly, Waste Determinations prepared at SRS in accordance with Section 3116 of the 2005 National Defense Authorization Act along with state operating permits establish the methodology and amounts of radioactivity that must be removed and may be disposed of in South Carolina. After removal of entrained solids and site-specific radionuclides, supernatant and saltcake wastes are considered to be low activity waste (LAW) and are immobilized in glass and disposed of at the Hanford Site Integrated Disposal Facility (IDF) or formulated into a grout for disposal at the Savannah River Site Saltstone Disposal Facility. Wastes stored at the Hanford Site or SRS comprise saltcake, supernate, and sludges. The supernatant and saltcake waste fractions contain primarily sodium salts, metals (e.g., Al, Cr), cesium-137 (Cs-137), technetium-99 (Tc-99) and entrained solids containing radionuclides such as strontium-90 (Sr-90) and transuranic elements. The sludges contain many of the transition metal hydroxides that precipitate when the spent acidic process solutions are rendered alkaline with sodium hydroxide. The sludges contain Sr-90 and transuranic elements. The wastes stored at each site have been generated and stored for over fifty years. Although the majority of the wastes were generated to support nuclear weapons production and reprocessing, the wastes differ substantially between the sites. Table 5 shows the volumes and total radioactivity (including decay daughters) of the waste phases stored in tanks at each site. At Hanford, there are 177 tanks that contain 56.5 Mgal of waste. SRS has 51 larger tanks, of which 2 are closed, that contain 36.5 Mgal. Mainly due to recovery operations, the waste stored at Hanford has less total curies than that stored at Savannah River. The total radioactivity of the Hanford wastes contains approximately 190 MCi, and the total radioactivity of the Savannah River wastes contains 400 MCi

ALUMINUM AND CHROMIUM LEACHING WORKSHOP WHITEPAPER

A workshop was held on January 23-24, 2007 to discuss the status of processes to leach constituents from High Level Waste (HLW) sludges at the Hanford and Savannah River Sites. The objective of the workshop was to examine the needs and requirements for the HLW flowsheet for each site, discuss the status of knowledge of the leaching processes, communicate the research plans, and identify opportunities for synergy to address knowledge gaps. The purpose of leaching of non-radioactive constituents from the sludge waste is to reduce the burden of material that must be vitrified in the HLW melter systems, resulting in reduced HLW glass waste volume, reduced disposal costs, shorter process schedules, and higher facility throughput rates. The leaching process is estimated to reduce the operating life cycle of SRS by seven years and decrease the number of HLW canisters to be disposed in the Repository by 1000 [Gillam et al., 2006]. Comparably at Hanford, the aluminum and chromium leaching processes are estimated to reduce the operating life cycle of the Waste Treatment Plant by 20 years and decrease the number of canisters to the Repository by 15,000-30,000 [Gilbert, 2007]. These leaching processes will save the Department of Energy (DOE) billions of dollars in clean up and disposal costs. The primary constituents targeted for removal by leaching are aluminum and chromium. It is desirable to have some aluminum in glass to improve its durability; however, too much aluminum can increase the sludge viscosity, glass viscosity, and reduce overall process throughput. Chromium leaching is necessary to prevent formation of crystalline compounds in the glass, but is only needed at Hanford because of differences in the sludge waste chemistry at the two sites. Improving glass formulations to increase tolerance of aluminum and chromium is another approach to decrease HLW glass volume. It is likely that an optimum condition can be found by both performing leaching and improving formulations. Disposal of the resulting aluminum and chromium-rich streams are different at the two sites, with vitrification into Low Activity Waste (LAW) glass at Hanford, and solidification in Saltstone at SRS. Prior to disposal, the leachate solutions must be treated to remove radionuclides, resulting in increased operating costs and extended facility processing schedules. Interim storage of leachate can also add costs and delay tank closure. Recent projections at Hanford indicate that up to 40,000 metric tons of sodium would be needed to dissolve the aluminum and maintain it in solution, which nearly doubles the amount of sodium in the entire current waste tank inventory. This underscores the dramatic impact that the aluminum leaching can have on the entire system. A comprehensive view of leaching and the downstream impacts must therefore be considered prior to implementation. Many laboratory scale tests for aluminum and chromium dissolution have been run on Hanford wastes, with samples from 46 tanks tested. Three samples from SRS tanks have been tested, out of seven tanks containing high aluminum sludge. One full-scale aluminum dissolution was successfully performed on waste at SRS in 1982, but generated a very large quantity of liquid waste ({approx}3,000,000 gallons). No large-scale tests have been done on Hanford wastes. Although the data to date give a generally positive indication that aluminum dissolution will work, many issues remain, predominantly because of variable waste compositions and changes in process conditions, downstream processing, or storage limitations. Better approaches are needed to deal with the waste volumes and limitations on disposal methods. To develop a better approach requires a more extensive understanding of the kinetics of dissolution, as well as the factors that effect rates, effectiveness, and secondary species. Models of the dissolution rate that have been developed are useful, but suffer from limitations on applicable compositional ranges, mineral phases, and particle properties that are difficult to measure. The experimental bases for the models contain very few data points

SODIUM ALUMINOSILICATE SOLIDS AFFINITY FOR CESIUM AND ACTINIDES

Washed sodium-aluminosilicate (NAS) solids at initial concentrations of 3.55 and 5.4 g/L sorb or uptake virtually no cesium over 288 hours, nor do any NAS solids generated during that time. These concentrations of solids are believed to conservatively bound current and near-term operations. Hence, the NAS solids should not have affected measurements of the cesium during the mass transfer tests and there is minimal risk of accumulating cesium during routine operations (and hence posing a gamma radiation exposure risk in maintenance). With respect to actinide uptake, it appears that NAS solids sorb minimal quantities of uranium - up to 58 mg U per kg NAS solid. The behavior with plutonium is less well understood. Additional study may be needed for radioactive operations relative to plutonium or other fissile component sorption or trapping by the solids. We recommend this testing be incorporated in the planned tests using samples from Tank 25F and Tank 49H to extend the duration to bound expected inventory time for solution

Results of Caustic Dissolution of Aluminosilicate Scale and Characterization Data for Samples From the Evaporator Pot and Gravity Drain Line

The build-up of sodium aluminosilicate scale in the 2H Evaporator system continues to cause operational difficulties. The use of a nitric acid cleaning operation proved successful in 2001. However, the operation required additional facilities to support spent cleaning solution neutralization and was quite costly. A proposed caustic cleaning flowsheet has many advantages over the acid flowsheet. Therefore, samples were retrieved from the evaporator system (gravity drain line and pot) for both chemical and radiological characterization and dissolution testing. The characterization of these scale samples showed the presence of nitrated cancrinite along with a dehydrated zeolite. Small amounts of depleted uranium were also found in these samples as expected and the amount of uranium ranged from 0.5 wt% to 2 wt%. Dissolution in sodium hydroxide solutions of various caustic concentrations showed that the scale slowly dissolves at elevated temperature (90 C). Data from similar testing indicate that the scale removed from the GDL in 2005 dissolves slower than that removed in 1997. Differences in the particle size of these samples of scale may well explain the measured dissolution rate differences

Results of the 2h Evaporator Acid Cleaning and in-Pot Neutralization

The estimated 200 gallons of sodium aluminosilicate scale (NAS) present in the 242-16H Evaporator pot prior to chemical cleaning was subjected to four batches of 1.5 M (9 wt%) nitric acid. Each batch was neutralized with 19 M (50 wt %) sodium hydroxide (caustic) before transfer to Tank 38. The chemical cleaning process began on November 20, 2006, and was terminated on December 10, 2006. An inspection of the pot's interior was performed and based on data gathered during that inspection; the current volume of scale in the pot is conservatively estimated to be 36.3 gallons, which is well below the 200 gallon limit specified in the Technical Safety Requirements. In addition, the performance during all aspects of cleaning agreed well with the flowsheet developed at the bench and pilot scale. There were some lessons learned during the cleaning outage and are detailed in appendices of this report

SORPTION OF URANIUM, PLUTONIUM AND NEPTUNIUM ONTO SOLIDS PRESENT IN HIGH CAUSTIC NUCLEAR WASTE STORAGE TANKS

Solids such as granular activated carbon, hematite and sodium phosphates, if present as sludge components in nuclear waste storage tanks, have been found to be capable of precipitating/sorbing actinides like plutonium, neptunium and uranium from nuclear waste storage tank supernatant liqueur. Thus, the potential may exists for the accumulation of fissile materials in such nuclear waste storage tanks during lengthy nuclear waste storage and processing. To evaluate the nuclear criticality safety in a typical nuclear waste storage tank, a study was initiated to measure the affinity of granular activated carbon, hematite and anhydrous sodium phosphate to sorb plutonium, neptunium and uranium from alkaline salt solutions. Tests with simulated and actual nuclear waste solutions established the affinity of the solids for plutonium, neptunium and uranium upon contact of the solutions with each of the solids. The removal of plutonium and neptunium from the synthetic salt solution by nuclear waste storage tank solids may be due largely to the presence of the granular activated carbon and transition metal oxides in these storage tank solids or sludge. Granular activated carbon and hematite also showed measurable affinity for both plutonium and neptunium. Sodium phosphate, used here as a reference sorbent for uranium, as expected, exhibited high affinity for uranium and neptunium, but did not show any measurable affinity for plutonium

USING WET AIR OXIDATION TECHNOLOGY TO DESTROY TETRAPHENYLBORATE

A bench-scale feasibility study on the use of a Wet Air Oxidation (WAO) process to destroy a slurry laden with tetraphenylborate (TPB) compounds has been undertaken. WAO is an aqueous phase process in which soluble and/or insoluble waste constituents are oxidized using oxygen or oxygen in air at elevated temperatures and pressures ranging from 150 C and 1 MPa to 320 C and 22 MPa. The products of the reaction are CO{sub 2}, H{sub 2}O, and low molecular weight oxygenated organics (e.g. acetate, oxalate). Test results indicate WAO is a feasible process for destroying TPB, its primary daughter products [triphenylborane (3PB), diphenylborinic acid (2PB), and phenylboronic acid (1PB)], phenol, and most of the biphenyl byproduct. The required conditions are a temperature of 300 C, a reaction time of 3 hours, 1:1 feed slurry dilution with 2M NaOH solution, the addition of CuSO{sub 4}.5H{sub 2}O solution (500 mg/L Cu) as catalyst, and the addition of 2000 mL/L of antifoam. However, for the destruction of TPB, its daughter compounds (3PB, 2PB, and 1PB), and phenol without consideration for biphenyl destruction, less severe conditions (280 C and 1-hour reaction time with similar remaining above conditions) are adequate

CHARACTERIZATION OF URANIUM SOLIDS PRECIPITATED WITH ALUMINOSILICATES

At the Savannah River Site (SRS), the High-Level Waste (HLW) Tank Farms store and process high-level liquid radioactive wastes from the Canyons and recycle water from the Defense Waste Processing Facility. The waste is concentrated using evaporators to minimize the volume of space required for HLW storage. Recently, the 2H Evaporator was shutdown due to the crystallization of sodium aluminosilicate (NAS) solids (such as cancrinite and sodalite) that contained close to 10 weight percent of elementally-enriched uranium (U). Prior to extensive cleaning, the evaporator deposits resided on the evaporator walls and other exposed internal surfaces within the evaporator pot. Our goal is to support the basis for the continued safe operation of SRS evaporators and to gain more information that could be used to help mitigate U accumulation during evaporator operation. To learn more about the interaction between U(VI) and NAS in HLW salt solutions, we performed several fundamental studies to examine the mechanisms of U accumulation with NAS in highly caustic solutions. This larger group of studies focused on the following processes: co-precipitation/structural incorporation, sorption, and precipitation (with or without NAS), which will be reviewed in this presentation. We will present and discuss local atomic structural characterization data about U that has been co-precipitated with NAS solids (such as amorphous zeolite precursor material and sodalite or Na 8 (AlSiO 4 ) 6 ·nH 2 O (s) ) using X-ray absorption fine-structure (XAFS) spectroscopic techniques. Our results indicate that U uptake from solution is greater during the precipitation of sodalite and amorphous zeolite precursor material than during the precipitation of zeolite A. The XAFS data indicate that U exists in several forms, such as U(VI) (uranyl-and uranate-type) oxide and oxyhydroxides (such as clarkeite). Crystalline forms of U(VI)-silicate were not resolved from the XAFS spectra but the presence of Si in the outer coordination shell of U indicate that the U is probably associated with amorphous silica. Mass balance determinations for U in these materials indicate that during formation, the structural incorporation of U within these structures is not a likely mechanism for accumulation. However, uptake of U was greatest during the precipitation of amorphous zeolite precursor material. Additionally, removal of U from solution by surface sorption on the NAS solids (a process which could have occurred after these solids were formed) probably had a minor role with respect to U accumulation in the 2H Evaporator. Processes most likely to largely influence on U accumulation are precipitation as U(VI) (as uranyl/uranate) oxide/oxyhydroxides and formation of an amorphous U-silica material. INTRODUCTION Uranium accumulation during the evaporation of HLW is a potential criticality risk if the incoming waste is enriched in 235 U. Little is known about the interactions between U and NAS in caustic, high Na + HLW salt solutions at room and at elevated temperature. To examine these interactions during NAS formation, we conducted studies that focused on potential mechanisms of U accumulation with NAS in the evaporators and in other process areas at the SRS that may concentrate U in the presence of silicates, Al and NAS. It is intended that the information gained from these studies will help support the basis for the continued safe operation of SRS evaporators and that this fundamental information will be used to help mitigate U accumulation during evaporator operation. Potential Routes of U Accumulation with NAS During the evaporation of caustic Na + -rich solutions, several processes could potentially contribute to the accumulation of U-containing solids. Uptake processes by solids can occur by several mechanisms: structural incorporation, ion exchange (electrostatic or outer-sphere) sorption, specific adsorption and surface precipitation/polymerization. WSRC-MS-2003-00898 3 Ion Exchange in a more restrictive sense as used in this study is an electrostatic process involving the replacement of one readily exchangeable hydrated ion by another similarly exchangeable ion Specific Adsorption (often referred to as Chemisorption or Inner Sphere sorption) involves the formation of predominantly covalent bonds with the surface, but the bonds can have some ionic behavior. These adsorbed metals typically have one or more atoms from the participating surface in the second coordination shell In HLW, U may be concentrated by sorption to the surfaces of the NAS, precipitation within NAS structures and precipitation as U phases. Sorption can be divided into two types of molecular scale processes (outer sphere and specific adsorption) that involve the uptake of atoms near or at a participating sorptive surface. An element such as U could co-precipitate with the NAS and related solids. [For zeolites, the term co-precipitation could be further divided to include uptake into zeolite channels and any isomorphic substitution (i.e., of U for Si or Al) in the zeolite structure Uranium may also interact with silica sols, which have no defined crystal structure because of their amorphous nature. At an atom-or molecular-scale basis, this type of interaction with U may be best be described by structural incorporation in Review of U(VI) Chemistry and Uptake Studies with U(VI) and Zeolites In oxidized systems, dissolved U exists as the highly soluble uranyl [U(VI)O 2 2+ ] species with two axial U=O double bonds at ~1.8 Å. This form of U(VI) can exist in U solids. However, U(VI) can also exist in solids as the less common uranate form, which has at least three single U-O bonds and no short axial double bonds. This form of U(VI) is very small in size (~0.72-0.8 Å) relative to the large uranyl ion group (~3.6 Å). 2-or NO 3 -solutions), U(VI) typically has a low affinity for certain solids, like the Fe oxides Use of XAFS Techniques to Characterize Metal Uptake by Surfaces The local environment of metals associated with surfaces can be investigated with analytical techniques such as XAFS spectroscopy. It is an X-ray-based technique that is non-destructive and provides average information on bulk and surface behavior. The XAFS spectroscopic techniques are among the best for providing detailed chemical speciation information in environmental samples-particularly when information from multiple characterization techniques is available. The term XAFS is applicable to both X-ray absorption near-edge structure (XANES) and extended X-ray absorption fine-structure (EXAFS) spectroscopic techniques. XAFS spectra give robust local structural information on coordination number (CN), bonding symmetry, neighbor and near-neighbor atomic distances and bond disorder (as the root mean square deviations of distances about the average values). Additionally, the information gained is atom specific-making it a versatile technique for structural determinations of atom clusters Experimental Objectives The primary objective of this research was to obtain information on speciation of U [added as U(VI)] associated with NAS solids that were synthesized with dissolved U using XAFS. Uranium-XAFS analyses were also conducted on solids that had been washed with solutions of DI water (only) and after washing with DI water and Na 2 CO 3 Washing U-loaded solids with Na 2 CO 3 solutions has been shown to remove sorbed forms of U(VI), in addition to dissolving the readily soluble (i.e., rapidly dissolving) solid phase forms of U(VI) MATERIALS AND EXPERIMENTAL METHODS Sample Preparation The NAS solids (amorphous zeolite, sodalite and zeolite A) were synthesized according to modified methods supplied by A. The difference between the amorphous zeolite and sodalite syntheses was temperature, in that the amorphous zeolite was made at 40 o C and the sodalite was made at 80 o C. Zeolite A was made at 90 o C. After preparation, the solids were washed three times in DI water, filtered with a 0.25 µm nylon filter, and dried in air. The air-dried solids were then washed three times with 0.4 M Na 2 CO 3 , filtered with a 0.25 µm nylon filter, and then air-dried. The air-dried solids were then provided to us for XAFS analyses. Sub-samples of the solids were digested in acid to determine the U concentrations after synthesis (using inductively-coupled argon plasma mass spectrometry) after each of the two washing steps. The results of the sample digests are shown in Additionally, to determine when the U should be added during these NAS syntheses, precipitation timing studies were done with U, Al and caustic salt solutions using the same experimental conditions (such as temperature) as those required for the individual NAS syntheses [ the reference samples as listed in 24]. No Si was added to avoid making NAS for each of these reference U materials. The U added to these solutions underwent precipitation and the unwashed solids were supplied for XAFS analyses. The solids in these "reference" samples may be representative of solids that can form in heated caustic solutions that are low in Si but contain high Al. EXAFS Data Collection and Analyses The XAFS data were collected on beamline X23a2 at the National Synchrotron Light Source (NSLS, Brookhaven National Laboratory, Upton, NY). Uranium-XAFS data were collected at the U L 3 -edge (17,166 eV) on the airdried filtered U-containing NAS solids. The XAFS data were collected in fluorescence mode using an unfocussed X-ray beam and a fixed-exit Si(311) monochromator. Ion chambers were used to collect incident (Io), transmission (It) and reference (Ir) signals. Gas ratios in Io were 100 % Ar. A Lytle detector was used to collect fluorescence X-WM'04 CONFERENCE, FEBRUARY 29-MARCH 4, 2004, TUSCON, AZ. WSRC-MS-2003-00898 6 rays (If). The monochromator energy was maximized using a piezo stack feedback energy stabilization system, with a settling time of 0.3 seconds per change in energy. An X-ray beam size of 2 by 28 mm 2 was used. Energy calibration was done using foils of Pt (L 1 -edge of 13,880 eV), Zr (K-edge, 17,998 eV), and Mo (K-edge, 20,000 eV). In simple terms, chi data (the plot of the wavevector in reciprocal space) show the oscillatory component (with both constructive and destructive interferences) of the atoms in the neighbor environment of the element of interest. The chi data represent part of the photoelectron wave that can be defined by the EXAFS equation Chi of k is the square root of [(2m / • 2 ) * (E -E 0 )]. S 0 2 is the amplitude reduction factor, which is associated with central atom shake-up and shake-off effects. SIGMA 2 or σ 2 is the Debye-Waller Factor or Relative Mean Square Disorder in bond length. "•" is Plank's constant and R pertains to mean atom position or bond distance (radial distance in Å). "m" is the mass of the photoelectron, E 0 is the EXAFS defined edge energy in electron volts or eV (not equal to edge energy as defined by XANES but is equal to the energy of the photoelectron at k = 0. "F of k" is the backscattering amplitude of the atom. N is the coordination number and δ(K) represents the electronic phase shifts due to atomic potentials. The background contribution to the EXAFS spectra was removed using an algorithm (AUTOBK) developed by RESULTS Background on the XAFS Characterization of Behavior of U on Surface