[en] The overall goals of the Tank Farm Vadose Zone Project, led by Washington River Protection Solutions, are to define risks from past and future single-shell tank farm activities; identify and evaluate the efficacy of interim measures; and aid, via collection of geochemical information and data, the future decisions that must be made by the U.S. Department of Energy (DOE) regarding the near-term operations, future waste retrieval, and final closure activities for the single-shell tank waste management areas (WMAs). To meet the investigative goals of the Tank Farm Vadose Zone Project, the Environmental Sciences Laboratory performed geochemical analyses on vadose zone sediments collected within Waste Management Area C. Tier one analyses of UPR-200-E-86, which includes direct push probe holes C5952, C5958 and C5960, were performed between 3/25/08 and 4/14/08. Preliminary results were presented to CH2M Hill Hanford Group on 6/5/08. As a result of the tier one investigations, further tier two analyses were requested. Tier two investigations include particle size and mineralogy analyses on samples collected between 80 to 120 feet below ground surface that were found to contain high concentrations of chloride and sulfate. Tier one analyses on sediments retrieved near UPR-200-E-81, direct push probe hole C6394, were performed between 6/20/08 and 7/22/08. Preliminary results of the tier one analyses were presented on 8/15/08. As a result of the tier one investigations, further tier two analyses were requested. Tier two analyses include determining whether U-236 exists in samples at approximately 42 feet below the ground surface. Confirmation of U-236 will determine whether the U-238 seen in the leaches performed on samples at that depth is a result of contamination and not from leaching natural uranium. Using the water and acid extract U-238 concentrations from the tier one analysis, equilibrium Kd values were requested to be calculated. Additional tier two analysis includes particle size analysis on samples collected at approximately 135 feet below ground surface to investigate a moist layer containing high chloride and sulfate anion concentrations. Particle size analysis was also requested for a sample collected at approximately 42 feet below ground surface due to its high moisture content and nitrate concentrations. The ESL was also requested to examine Paleosols to determine where the paleosols from the Integrated Disposal Facility would extrapolate to in WMA C.

[en] Understanding mechanisms and kinetics of mineral carbonation reactions relevant to sequestering carbon dioxide as a supercritical fluid (scCO2) in geologic formations is crucial to accurately predicting long-term storage risks. Most attention so far has been focused on reactions occurring between silicate minerals and rocks in the aqueous dominated CO2-bearing fluid. However, water-bearing scCO2 also comprises a reactive fluid, and in this situation mineral carbonation mechanisms are poorly understood. Using in situ high-pressure x-ray diffraction, the carbonation of brucite [Mg(OH)2] in wet scCO2 was examined at pressure (82 bar) as a function of water concentration and temperature (50 C and 75 C). Exposing brucite to anhydrous scCO2 at either temperature resulted in little or no detectable reaction over three days. However, addition of trace amounts of water resulted in partial carbonation of brucite into nesquehonite [MgCO3 3H2O] within a few hours at 50 C. By increasing water content to well above the saturation level of the scCO2, complete conversion of brucite into nesquehonite was observed. Tests conducted at 75 C resulted in the conversion of brucite into magnesite [MgCO3] instead, apparently through an intermediate nesquehonite step. Raman spectroscopy applied to brucite reacted with 18O-labeled water in scCO2 show it was incorporated into carbonate at relatively high concentration. This supports a carbonation mechanism with at least one step involving a direct reaction between the mineral and water molecules without mediation by a condensed aqueous layer.

[en] Recent field investigation that collected and characterized vadose zone sediments from beneath inactive liquid disposal facilities at the Hanford 200 Areas show lower than expected concentrations of a long-term risk driver, Tc-99. Therefore laboratory studies were performed to re-create one of the three processes that were used to separate the plutonium from spent fuel and that created most of the wastes disposed or currently stored in tanks at Hanford. The laboratory simulations were used to compare with current estimates based mainly on flow sheet estimates and spotty historical data. Three simulations of the bismuth phosphate precipitation process show that less that 1% of the Tc-99, Cs-135/137, Sr-90, I-129 carry down with the Pu product and thus these isotopes should have remained within the metals waste streams that after neutralization were sent to single shell tanks. Conversely, these isotopes should not be expected to be found in the first and subsequent cycle waste streams that went to cribs. Measurable quantities (∼20 to 30%) of the lanthanides, yttrium, and trivalent actinides (Am and Cm) do precipitate with the Pu product, which is higher than the 10% estimate made for current inventory projections. Surprisingly, Se (added as selenate form) also shows about 10% association with the Pu/bismuth phosphate solids. We speculate that the incorporation of some Se into the bismuth phosphate precipitate is caused by selenate substitution into crystal lattice sites for the phosphate. The bulk of the U daughter product Th-234 and Np-237 daughter product Pa-233 also associate with the solids. We suspect that the Pa daughter products of U (Pa-234 and Pa-231) would also co-precipitate with the bismuth phosphate induced solids. No more than 1 % of the Sr-90 and Sb-125 should carry down with the Pu product that ultimately was purified. Thus the current scheme used to estimate where fission products end up being disposed overestimates by one order of magnitude the partitioning Sr-90, Cs-137, and Sb-125 and by at least two orders of magnitude the portioning of Tc-99 to the first and subsequent cycle waste streams that went to cribs. Conversely, the current scheme underestimates the lanthanide and yttrium fission product quantities that went to cribs by a factor of about 3.

[en] This report presents the results of laboratory characterization, testing, and analysis for a composite sample (designated 20578) of residual waste collected from single-shell tank C-108 during the waste retrieval process after modified sluicing. These studies were completed to characterize concentration and form of contaminant of interest in the residual waste; assess the leachability of contaminants from the solids; and develop release models for contaminants of interest. Because modified sluicing did not achieve 99% removal of the waste, it is expected that additional retrieval processing will take place. As a result, the sample analyzed here is not expected to represent final retrieval sample.

[en] Static batch experiments were used to examine the sorption of dissolved perrhenate [Re(VII)], as a surrogate for pertechnetate [Tc(VII)], on corrosion products of A-516 carbon steel coupons contacted with synthetic groundwater or dilute water. After 109 days of contact time, the concentration of dissolved Re(VII) in the synthetic groundwater matrix decreased by approximately 26%; the dilute water matrix experienced a 99% decrease in dissolved Re(VII) over the same time period. Bulk XRD results for the corroded steel coupons showed that the corrosion products consisted primarily of maghemite, lepidocrocite, and goethite. Analyses of the coupons by SEM/EDS indicated that Re was present with the morphologically complex assemblages of Fe oxide/hydroxide corrosion products for samples spiked with the highest dissolved Re(VII) concentration (1.0 mmol/L) used for these experiments. Analyses of corroded steel coupons contacted with solutions containing 1.0 mmol/L Re(VII) by synchrotron-based methods confirmed the presence of Re sorbed with the corrosion product on the steel coupons. Analyses showed that the Re sorbed on these corroded coupons was in the +7 oxidation state, suggesting that the Re(VII) uptake mechanism did not involve reduction of Re to a lower oxidation state, such as +4. The results of our studies using Re(VII) as an analogue for Tc(VII)-99 suggest that Tc(VII)-99 would also be sorbed with steel corrosion products and that the inventory of Tc(VII)-99 released from breached waste packages would be lower than what is now conservatively estimated.

[en] This report describes the development of release models for key contaminants that are present in residual sludge remaining after closure of Hanford Tanks 241-C-203 (C-203) and 241-C-204 (C-204). The release models were developed from data generated by laboratory characterization and testing of samples from these two tanks. Key results from this work are (1) future releases from the tanks of the primary contaminants of concern (99Tc and 238U) can be represented by relatively simple solubility relationships between infiltrating water and solid phases containing the contaminants; and (2) high percentages of technetium-99 in the sludges (20 wt% in C-203 and 75 wt% in C-204) are not readily water leachable, and, in fact, are very recalcitrant. This is similar to results found in related studies of sludges from Tank AY-102. These release models are being developed to support the tank closure risk assessments performed by CH2M HILL Hanford Group, Inc., for the U.S. Department of Energy

[en] This report was revised in May 2007 to correct 90Sr values in Chapter 3. The changes were made on page 3.9, paragraph two and Table 3.10; page 3.16, last paragraph on the page; and Tables 3.21 and 3.31. The rest of the text remains unchanged from the original report issued in October 2004. This report describes the development of release models for key contaminants that are present in residual sludge remaining after closure of Hanford Tanks 241-C-203 (C-203) and 241-C-204 (C-204). The release models were developed from data generated by laboratory characterization and testing of samples from these two tanks. Key results from this work are (1) future releases from the tanks of the primary contaminants of concern (99Tc and 238U) can be represented by relatively simple solubility relationships between infiltrating water and solid phases containing the contaminants; and (2) high percentages of technetium-99 in the sludges (20 wt% in C-203 and 75 wt% in C-204) are not readily water leachable, and, in fact, are very recalcitrant. This is similar to results found in related studies of sludges from Tank AY-102. These release models are being developed to support the tank closure risk assessments performed by CH2M HILL Hanford Group, Inc., for the U.S. Department of Energy.

[en] This report presents the results of laboratory characterization and testing of two samples (designated 20406 and 20407) of residual waste collected from tank S-112 after final waste retrieval. These studies were completed to characterize the residual waste and assess the leachability of contaminants from the solids. This is the first report from this PNNL project to describe the composition and leach test data for residual waste from a salt cake tank. All previous PNNL reports (Cantrell et al. 2008; Deutsch et al. 2006, 2007a, 2007b, 2007c) describing contaminant release models, and characterization and testing results for residual waste in single-shell tanks were based on samples from sludge tanks

[en] This report was revised in May 2007 to correct values in Section 3.4.1.7, second paragraph, last sentence; 90Sr values in Tables 3.22 and 3.32; and 99Tc values Table 4.3 and in Chapter 5. In addition, the tables in Appendix F were updated to reflect corrections to the 90Sr values. The rest of the text remains unchanged from the original report issued in May 2005. CH2M HILL is producing risk/performance assessments to support the closure of single-shell tanks at the DOE's Hanford Site. As part of this effort, staff at PNNL were asked to develop release models for contaminants of concern that are present in residual sludge remaining in tank 241-C-106 (C-106) after final retrieval of waste from the tank. This report provides the information developed by PNNL.

[en] This data package documents the experimentally derived input data on the representative waste glasses LAWABP1 and HLP-31 that will be used for simulations of the immobilized lowactivity waste disposal system with the Subsurface Transport Over Reactive Multiphases (STORM) code. The STORM code will be used to provide the near-field radionuclide release source term for a performance assessment to be issued in March of 2001. Documented in this data package are data related to (1) kinetic rate law parameters for glass dissolution, (2) alkali-H ion exchange rate, (3) chemical reaction network of secondary phases that form in accelerated weathering tests, and (4) thermodynamic equilibrium constants assigned to these secondary phases. The kinetic rate law and Na+-H+ ion exchange rate were determined from single-pass flow-through experiments. Pressurized unsaturated flow and vapor hydration experiments were used for accelerated weathering or aging of the glasses. The majority of the thermodynamic data were extracted from the thermodynamic database package shipped with the geochemical code EQ3/6. However, several secondary reaction products identified from laboratory tests with prototypical LAW glasses were not included in this database, nor are the thermodynamic data available in the open literature. One of these phases, herschelite, was determined to have a potentially significant impact on the release calculations and so a solubility product was estimated using a polymer structure model developed for zeolites. Although this data package is relatively complete, final selection of ILAW glass compositions has not been done by the waste treatment plant contractor. Consequently, revisions to this data package to address new ILAW glass formulations are to be regularly expected.