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Description: Low-level radioactive liquid wastes at the U.S. Department of Energy Savannah River Site are treated by mixing the wastes with Saltstone grout to generate the Saltstone waste form that is poured into the concrete vaults for long-term disposal. The formula for Saltstone includes {approx}25 wt% slag to create a reducing environment for mitigating the sub-surface transport of several radionuclides, including Tc-99. A two-dimensional reactive transport model was developed to estimate how long the Z-Area Saltstone will maintain a reducing environment, and therefore its ability to request Tc-99. The model predicted that {approx}16% of the Saltstone reduction capacity would be consumed after 213,000 years. Independent calculations published by other researchers yielded nearly identical results. The general modeling approach and the study results are presented in this paper.
Date: December 18, 2006
Creator: Hang, T & Daniel Kaplan, D
Partner: UNT Libraries Government Documents Department

Effects of Temperature and CSSX Organics on Saltstone Processing Properties

Description: This task was performed to determine whether the two variables, ''mix temperature'' and ''quantity of organics'' introduced into the decontaminated salt solution by the caustic side solvent extraction (CSSX) process, need to be included in the upcoming Saltstone Variability Study. Because the amount and types of organics introduced through the CSSX process do not significantly impact the fresh properties of Saltstone, the ''quantity of organics'' variable will not be included in the Saltstone Variability Study. The Saltstone Variability Study should include the variable of ''mix temperature'' in the experimental design. Examples are presented in this report that clearly demonstrate a pronounced dependence of the fresh grout properties on ''mix temperature''. One example, using mixes made with the Deliquification, Dissolution and Adjustment (DDA) simulant, shows that the properties of gel time and bleed water are highly mix temperature dependent. The gel time increased from 15 minutes at 10 C to 90 minutes at 35 C with most of the change occurring between 20 and 30 C. That is, gel time is highly sensitive to mix temperature, especially in the temperature range over which processing is most likely. The volume percent bleed water for these mixes increased from {approx}1 % at 10 C to 13 % at 35 C. The gel times and volume percent bleed water are correlated such that the longer the gel time, the greater the amount of bleed water. In another example, and in contrast to the DDA results, gel times decreased with increasing temperatures for mixes made using the Modular CSSX Unit (MCU) simulants. In this case the gel time decreased from 150 minutes at 10 C to 20 minutes at 38 C. The rheological properties of these mixes were shown to be dependent on temperature over the range of 10 to 40 C. The plastic viscosity ...
Date: February 6, 2006
Creator: Harbour, J
Partner: UNT Libraries Government Documents Department

ISOPAR L Release Rates from Saltstone Using Simulated Salt Solutions

Description: The Modular Caustic-Side Solvent Extraction (CSSX) Unit (MCU) and the Salt Waste Processing Facility (SWPF) will produce a Deactivated Salt Solution (DSS) that will go to the Saltstone Production Facility (SPF). Recent information indicates that solvent entrainment in the DSS is larger than expected. The main concern is with Isopar{reg_sign} L, the diluent in the solvent mixture, and its flammability in the saltstone vault. If it is assumed that all the Isopar{reg_sign} L is released instantaneously into the vault from the curing grout before each subsequent pour; the Isopar{reg_sign} L in the vault headspace is well mixed; and each pour displaces an equivalent volume of headspace, the allowable concentration of Isopar{reg_sign} L in the DSS sent to SPF has been calculated at approximately 4 ppm. The amount allowed would be higher, if the release from grout were significantly less. The Savannah River National Laboratory was tasked with determining the release of Isopar{reg_sign} L from saltstone prepared with a simulated DSS with Isopar{reg_sign} L concentrations ranging from 50 mg/L to 200 mg/L in the salt fraction and with test temperatures ranging from ambient to 95 C. The results from the curing of the saltstone showed that the Isopar{reg_sign} L release data can be treated as a percentage of initial concentration in the concentration range studied. The majority of the Isopar{reg_sign} L that was released over the test duration was released in the first few days. The release of Isopar{reg_sign} L begins immediately and the rate of release decreases over time. At higher temperatures the immediate release is larger than at lower temperatures. In one test at 95 C essentially all of the Isopar{reg_sign} L was released in three months. Initial curing temperature was found to be very important as slight variations during the first few days affected the final Isopar{reg_sign} L ...
Date: February 6, 2006
Creator: Bronikowski, M
Partner: UNT Libraries Government Documents Department


Description: A salt solution (doped with Tc-99), that simulates the salt waste stream to be processed at the Saltstone Production Facility, was immobilized in grout waste forms with and without (1) ground granulated blast furnace slag and (2) pretreatment with iron salts. The degree of immobilization of Tc-99 was measured through monolithic and crushed grout leaching tests. Although Fe (+2) was shown to be effective in reducing Tc-99 to the +4 state, the strong reducing nature of the blast furnace slag present in the grout formulation dominated the reduction of Tc-99 in the cured grouts. An effective diffusion coefficient of 4.75 x 10{sup -12} (Leach Index of 11.4) was measured using the ANSI/ANS-16.1 protocol. The leaching results show that, even in the presence of a concentrated salt solution, blast furnace slag can effectively reduce pertechnetate to the immobile +4 oxidation state. The measured diffusivity was introduced into a flow and transport model (PORFLOW) to calculate the release of Tc-99 from a Saltstone Vault as a function of hydraulic conductivity of the matrix.
Date: December 11, 2006
Creator: Harbour, J
Partner: UNT Libraries Government Documents Department

Hanford Tank 241-C-106: Impact of Cement Reactions on Release of Contaminants from Residual Waste

Description: The CH2M HILL Hanford Group, Inc. (CH2M HILL) is producing risk/performance assessments to support the closure of single-shell tanks at the U.S. Department of Energy's Hanford Site. As part of this effort, staff at Pacific Northwest National Laboratory 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. Initial work to produce release models was conducted on residual tank sludge using pure water as the leaching agent. The results were reported in an earlier report. The decision has now been made to close the tanks after waste retrieval with a cementitious grout to minimize infiltration and maintain the physical integrity of the tanks. This report describes testing of the residual waste with a leaching solution that simulates the composition of water passing through the grout and contacting the residual waste at the bottom of the tank.
Date: September 1, 2006
Creator: Deutsch, William J.; Krupka, Kenneth M.; Lindberg, Michael J.; Cantrell, Kirk J.; Brown, Christopher F. & Schaef, Herbert T.
Partner: UNT Libraries Government Documents Department

Diffusion of Iodine and Rhenium in Category 3 Waste Encasement Concrete and Soil Fill Material

Description: Assessing long-term performance of Category 3 waste cement grouts for radionuclide encasement requires knowledge of the radionuclide-cement interactions and mechanisms of retention (i.e. sorption or precipitation). This understanding will enable accurate prediction of radionuclide fate when the waste forms come in contact with groundwater. A set of diffusion experiments using carbonated and non-carbonated concrete-soil half cells was conducted under unsaturated conditions (4% and 7% by wt moisture content). Spiked concrete half-cell specimens were prepared with and without colloidal metallic iron addition and were carbonated using supercritical carbon dioxide. Spikes of I and Re were added to achieve measurable diffusion profile in the soil part of the half-cell. In addition, properties of concrete materials likely to influence radionuclide migration such as carbonation were evaluated in an effort to correlate these properties with the release of iodine and rhenium.
Date: December 15, 2006
Creator: Wellman, Dawn M.; Mattigod, Shas V.; Whyatt, Greg A.; Powers, Laura; Parker, Kent E. & Wood, Marcus I.
Partner: UNT Libraries Government Documents Department


Description: Savannah River National Laboratory (SRNL) personnel were requested to confirm the Tank 50 Batch 0 grout formulation per Technical Task Request, SSF-TTR-2006-0001 (task 1 of 2) [1]. Earlier Batch 0 formulation testing used a Tank 50 sample collected in September 2005 and is described elsewhere [2]. The current testing was performed using a sample of Tank 50 waste collected in May 2006. This work was performed according to the Technical Task and Quality Assurance Plan (TT/QAP), WSRC-RP-2006-00594 [3]. The salt solution collected from Tank 50 in May 2006 contained approximately 3 weight percent more solids than the sample collected in September 2005. The insoluble solids took longer to settle in the new sample which was interpreted as indicating finer particles in the current sample. The saltstone formulation developed for the September 2005 Tank 50 Batch 0 sample was confirmed for the May 2006 sample with one minor exception. Saltstone prepared with the Tank 50 sample collected in May 2006 required 1.5 times more Daratard 17 set retarding admixture than the saltstone prepared with the September In addition, a sample prepared with lower shear mixing (stirring with a spatula) had a higher plastic viscosity (57 cP) than samples made with higher shear mixing in a blender (23cP). The static gel times of the saltstone slurries made with low shear mixing were also shorter ({approx}32 minutes) than those for comparable samples made in the blender ({approx}47 minutes). The addition of the various waste streams (ETP, HEU-HCAN, and GPE-HCAN) to Tank 50 from September 2005 to May 2006 has increased the amount of set retarder, Daratard 17, required for processing saltstone slurries through the Saltstone facility. If these streams are continued to be added to Tank 50, the quantity of admixtures required to maintain the same processing conditions for the Saltstone facility will ...
Date: June 5, 2006
Creator: Langton, C.
Partner: UNT Libraries Government Documents Department

Experimental Test Plan for Grouting H-3 Calcine

Description: Approximately 4400 cubic meters of solid high-level waste called calcine are stored at the Idaho Nuclear Technology and Engineering Center. Under the Idaho Cleanup Project, dual disposal paths are being investigated. The first path includes calcine retrieval, package "as-is", and ship to the Monitored Geological Repository (MGR). The second path involves treatment of the calcine with such methods as vitrification or grouting. This test plan outlines the hot bench scale tests to grout actual calcine and verify that the waste form properties meet the waste acceptance criteria. This is a necessary sequential step in the process of qualifying a new waste form for repository acceptance. The archive H-3 calcine samples at the Contaminated Equipment Maintenance Building attached to New Waste Calcining Facility will be used in these tests at the Remote Analytical Laboratory. The tests are scheduled for the second quarter of fiscal year 2007.
Date: January 1, 2006
Creator: Herbst, Alan K.
Partner: UNT Libraries Government Documents Department

Adiabatic Heat of Hydration Calorimetric Measurements for Reference Saltstone Waste

Description: The production of nuclear materials for weapons, medical, and space applications from the mid-1950's through the late-1980's at the Savannah River Site (SRS) generated approximately 35 million gallons of liquid high-level radioactive waste, which is currently being processed into vitrified glass for long-term storage. Upstream of the vitrification process, the waste is separated into three components: high activity insoluble sludge, high activity insoluble salt, and very low activity soluble salts. The soluble salt represents 90% of the 35 million gallons of overall waste and is processed at the SRS Saltstone Facility, where it mixed with cement, blast furnace slag, and flyash, creating a grout-like mixture. The resulting grout is pumped into aboveground storage vaults, where it hydrates into concrete monoliths, called saltstone, thus immobilizing the low-level radioactive salt waste. As the saltstone hydrates, it generates heat that slowly diffuses out of the poured material. To ensure acceptable grout properties for disposal and immobilization of the salt waste, the grout temperature must not exceed 95 C during hydration. Adiabatic calorimetric measurements of the heat generated for a representative sample of saltstone were made to determine the time-dependent heat source term. These measurements subsequently were utilized as input to a numerical conjugate heat transfer model to determine the expected peak temperatures for the saltstone vaults.
Date: January 12, 2006
Creator: Bollinger, James
Partner: UNT Libraries Government Documents Department

Report for the HWMA/RCRA Post Closure Permit for the INTEC Waste Calcining Facility at the INL Site

Description: The Waste Calcining Facility (WCF) is located at the Idaho Nuclear Technology and Engineering Center. In 1998, the WCF was closed under an approved Hazardous Waste Management Act/Resource Conservation and Recovery Act (HWMA/RCRA) Closure Plan. Vessels and spaces were grouted and then covered with a concrete cap. The Idaho Department of Environmental Quality issued a final HWMA/RCRA post-closure permit on September 15, 2003, with an effective date of October 16, 2003. This permit sets forth procedural requirements for groundwater characterization and monitoring, maintenance, and inspections of the WCF to ensure continued protection of human health and the environment. The post-closure permit also includes semiannual reporting requirements under Permit Conditions III.H. and I.U. These reporting requirements have been combined into this single semiannual report.
Date: June 1, 2006
Creator: Project, Idaho Cleanup
Partner: UNT Libraries Government Documents Department

Cleanup Verification Package for the 600-259 Waste Site

Description: This cleanup verification package documents completion of remedial action for the 600-259 waste site. The site was the former site of the Special Waste Form Lysimeter, consisting of commercial reactor isotope waste forms in contact with soils within engineered caissons, and was used by Pacific Northwest National Laboratory to collect data regarding leaching behavior for target analytes. A Grout Waste Test Facility also operated at the site, designed to test leaching rates of grout-solidified low-level radioactive waste.
Date: February 9, 2006
Creator: Capron, J. M.
Partner: UNT Libraries Government Documents Department

Environmental Molecular Sciences Institute support for the Center for Environmental Molecular Sciences

Description: This project is investigating the chemical processes that govern actinide sequestration in grout materials with the goal of determining the long-term behavior of grouts used to stabilize actinides in source-terms such as high level waste tank heals. Two grouts contained portland cement, blast furnace slag and fly ash, with one formulation containing zeolite and the other fluorapatite. Earlier experimental work was conducted with funds from DOE/West Valley. CEMS funding allowed further exploration of grout behavior, beyond the scope of the original work which consisted of both batch and flow-through column experiments. The primary focus was the late stage behavior of actinides in the grout system when it is expected to be open to the atmosphere and groundwater, resulting in decreases of pH and interactions of U (and other elements) with dissolved carbonate.
Date: November 1, 2006
Creator: Dodge, C.; Fitts, J.; Francis, A.J.; Fuhrmann, M.; Gillow, J.; Kalb, P. et al.
Partner: UNT Libraries Government Documents Department