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A THERMAL MODEL OF THE IMMOBILIZATION OF LOW-LEVEL RADIOACTIVE WASTE AS GROUT IN CONCRETE VAULTS

Description: Salt solution will be mixed with cement and flyash/slag to form a grout which will be immobilized in above ground concrete vaults. The curing process is exothermic, and a transient thermal model of the pouring and curing process is herein described. A peak temperature limit of 85 C for the curing grout restricts the rate at which it can be poured into a vault. The model is used to optimize the pouring.
Date: October 27, 2008
Creator: Shadday, M
Partner: UNT Libraries Government Documents Department

HIGH LEVEL WASTE MECHANCIAL SLUDGE REMOVAL AT THE SAVANNAH RIVER SITE F TANK FARM CLOSURE PROJECT

Description: The Savannah River Site F-Tank Farm Closure project has successfully performed Mechanical Sludge Removal (MSR) using the Waste on Wheels (WOW) system for the first time within one of its storage tanks. The WOW system is designed to be relatively mobile with the ability for many components to be redeployed to multiple waste tanks. It is primarily comprised of Submersible Mixer Pumps (SMPs), Submersible Transfer Pumps (STPs), and a mobile control room with a control panel and variable speed drives. In addition, the project is currently preparing another waste tank for MSR utilizing lessons learned from this previous operational activity. These tanks, designated as Tank 6 and Tank 5 respectively, are Type I waste tanks located in F-Tank Farm (FTF) with a capacity of 2,840 cubic meters (750,000 gallons) each. The construction of these tanks was completed in 1953, and they were placed into waste storage service in 1959. The tank's primary shell is 23 meters (75 feet) in diameter, and 7.5 meters (24.5 feet) in height. Type I tanks have 34 vertically oriented cooling coils and two horizontal cooling coil circuits along the tank floor. Both Tank 5 and Tank 6 received and stored F-PUREX waste during their operating service time before sludge removal was performed. DOE intends to remove from service and operationally close (fill with grout) Tank 5 and Tank 6 and other HLW tanks that do not meet current containment standards. Mechanical Sludge Removal, the first step in the tank closure process, will be followed by chemical cleaning. After obtaining regulatory approval, the tanks will be isolated and filled with grout for long-term stabilization. Mechanical Sludge Removal operations within Tank 6 removed approximately 75% of the original 95,000 liters (25,000 gallons). This sludge material was transferred in batches to an interim storage tank to prepare for ...
Date: January 15, 2008
Creator: Jolly, R & Bruce Martin, B
Partner: UNT Libraries Government Documents Department

AMMONIA CONCENTRATION IN SALTSTONE HEADSPACE SUMMARY REPORT

Description: The Saltstone Facility Documented Safety Analysis (DSA) is under revision to accommodate changes in the Composite Lower Flammability Limit (CLFL) from the introduction of Isopar into Tank 50. Saltstone samples were prepared with an 'MCU' type salt solution spiked with ammonia. The ammonia released from the saltstone was captured and analyzed. The ammonia concentration found in the headspace of samples maintained at 95 C and 1 atm was, to 95% confidence, less than or equal to 3.9 mg/L. Tank 50 is fed by several influent streams. The salt solution from Tank 50 is pumped to the salt feed tank (SFT) in the Saltstone Production Facility (SPF). The premix materials cement, slag and fly ash are blended together prior to transfer to the grout mixer. The premix is fed to the grout mixer in the SPF and the salt solution is incorporated into the premix in the grout mixer, yielding saltstone slurry. The saltstone slurry drops into a hopper and then is pumped to the vault. The Saltstone Facility Documented Safety Analysis (DSA) is under revision to accommodate changes in the Composite Lower Flammability Limit (CLFL) from the introduction of Isopar{reg_sign} L into Tank 50. Waste Solidification-Engineering requested that the Savannah River National Laboratory (SRNL) perform testing to characterize the release of ammonia in curing saltstone at 95 C. The test temperature represents the maximum allowable temperature in the Saltstone Disposal Facility (SDF). Ammonia may be present in the salt solution and premix materials, or may be produced by chemical reactions when the premix and salt solution are combined. A final report (SRNS-STI-2008-00120, Rev. 0) will be issued that will cover in more depth the information presented in this report.
Date: September 26, 2008
Creator: Zamecnik, J & Alex Cozzi, A
Partner: UNT Libraries Government Documents Department

FLUOR HANFORD DECOMMISSIONING UPDATE

Description: Fluor Hanford is completing D&D of the K East Basin at the U.S. Department of Energy's (DOE's) Hanford Site in southeastern Washington State this spring, with demolition expected to begin in June. Located about 400 yards from the Columbia River, the K East Basin is one of two indoor pools that formerly contained irradiated nuclear fuel, radioactive sludge and tons of contaminated debris. In unique and path-breaking work, workers finished removing the spent fuel from the K Basins in 2004. In May 2007, workers completed vacuuming the sludge into containers in the K East Basin, and transferring it into containers in the K West Basin. In December, they finished vacuuming the remainder of K West Basin sludge into these containers. The K East Basin was emptied of its radioactive inventory first because it was more contaminated than the K West Basin, and had leaked in the past. In October 2007, Fluor Hanford began physical D&D of the 8,400-square foot K East Basin by pouring approximately 14-inches of grout into the bottom of it. Grout is a type of special cement used for encasing waste. Two months later, Fluor Hanford workers completed sluicing contaminated sand from the large filter that had sieved contaminants from the basin water for more than 50 years. Next, they poured grout into the filter housing and the vault that surrounds the filter, as well as into ion exchange columns that also helped filter basin water. For a six-week period in February and March, personnel drained the approximately one million gallons of contaminated water from the K East Basin. The effort required more than 200 tanker truck loads that transported the water to an effluent treatment facility for treatment and then release. A thin fixative was also applied to the basin walls as the water was removed ...
Date: April 21, 2008
Creator: MS, GERBER
Partner: UNT Libraries Government Documents Department

RESULTS OF IONSIV® IE-95 STUDIES FOR THE REMOVAL OF RADIOACTIVE CESIUM FROM K-EAST BASIN SPENT NUCLEAR FUEL POOL DURING DECOMMISSIONING ACTIVITIES

Description: This report delineates the results obtained from laboratory testing of IONISIV{reg_sign} IE-95 to determine the efficacy of the zeolite for the removal of radioactive cesium from the KE Basin water prior to transport to the Effluent Treatment Facility, as described in RPP-PLAN-36158, IONSIV{reg_sign} IE-95 Studies for the removal of Radioactive Cesium from KE Basin Spent Nuclear Fuel Pool during Decommissioning Activities. The spent nuclear fuel was removed from KE Basin and the remaining sludge was layered with a grout mixture consisting of 26% Lehigh Type I/II portland cement and 74% Boral Mohave type F fly ash with a water-to-cement ratio of 0.43. The first grout pour was added to the basin floor to a depth of approximately 14 in. covering an area of 12,000 square feet. A grout layer was also added to the sludge containers located in the attached Weasel and Technical View pits.
Date: July 7, 2008
Creator: JB, DUNCAN & SP, BURKE
Partner: UNT Libraries Government Documents Department

CLOSURE OF HLW TANKS PHASE 2 FULL SCALE COOLING COILS GROUT FILL DEMONSTATIONS

Description: This report documents the Savannah River National Laboratory (SRNL) support for the Tank Closure and Technology Development (TCTD) group's strategy for closing high level radioactive waste (HLW) tanks at the Savannah River Site (SRS). Specifically, this task addresses the ability to successfully fill intact cooling coils, presently within the HLW tanks, with grout that satisfies the fresh and cured grout requirements [1] under simulated field conditions. The overall task was divided into two phases. The first phase was the development of a grout formulation that satisfies the processing requirements for filling the HLW tank cooling coils [5]. The second phase of the task, which is documented in this report, was the filling of full scale cooling coils under simulated field conditions using the grout formulation developed in the first phase. SRS Type I tank cooling coil assembly design drawings and pressure drop calculations were provided by the Liquid Waste (LW) customer to be used as the basis for configuring the test assemblies. The current concept for closing tanks equipped with internal cooling coils is to pump grout into the coils to inhibit pathways for infiltrating water. Access to the cooling coil assemblies is through the existing supply/return manifold headers located on top of the Type I tanks. The objectives for the second phase of the testing, as stated in the Task Technical and Quality Assurance plan (TTQAP) [2], were to: (1) Perform a demonstration test to assess cooling coil grout performance in simulated field conditions, and (2) Measure relevant properties of samples prepared under simulated field conditions. SRNL led the actual work of designing, fabricating and filling two full-scale cooling coil assemblies which were performed at Clemson Engineering Technologies Laboratory (CETL) using the South Carolina University Research and Education Foundation (SCUREF) program. A statement of work (SOW) was issued to ...
Date: June 19, 2008
Creator: Hansen, E & Alex Cozzi, A
Partner: UNT Libraries Government Documents Department

IMPACT OF ALUMINATE IONS ON THE PROPERTIES OF SALTSTONE GROUT MIXES

Description: It is important to identify and control the operational and compositional variables that impact the important processing and performance properties of Saltstone grout mixes. The grout that is produced at the Saltstone Production Facility (SPF) is referred to as Saltstone and is a waste form that immobilizes low concentrations of radionuclides as well as certain toxic metals. The Saltstone will be disposed of in vaults at Savannah River Site (SRS). An effort referred to as the Saltstone Variability Study has been initiated to achieve this goal. The protocols developed in this variability study are also ideally suited as a tool to assess the impact of proposed changes to the processing flow sheet for Liquid Waste Operations at SRS. One such proposal that is currently under consideration is to introduce a leaching step in the treatment of the High Level Waste (HLW) sludge to remove aluminum prior to vitrification at the Defense Waste Processing Facility (DWPF). This leachate would significantly increase the soluble aluminate concentration in the salt feed that will be processed at the SPF. Consequently, an initial study of the impact of increased aluminate concentration on the Saltstone grout properties was performed. Prior work by Lukens (1) showed that aluminate in the salt solutions increases the amount of heat generation.
Date: February 21, 2008
Creator: Harbour, J; Tommy Edwards, T; Erich Hansen, E & Vickie Williams, V
Partner: UNT Libraries Government Documents Department

HYDRAULIC AND PHYSICAL PROPERTIES OF SALTSTONE GROUTS AND VAULT CONCRETES

Description: The Saltstone Disposal Facility (SDF), located in the Z-Area of the Savannah River Site (SRS), is used for the disposal of low-level radioactive salt solution. The SDF currently contains two vaults: Vault 1 (6 cells) and Vault 4 (12 cells). Additional disposal cells are currently in the design phase. The individual cells of the saltstone facility are filled with saltstone. Saltstone is produced by mixing the low-level radioactive salt solution, with blast furnace slag, fly ash, and cement (dry premix) to form a dense, micro-porous, monolithic, low-level radioactive waste form. The saltstone is pumped into the disposal cells where it subsequently solidifies. Significant effort has been undertaken to accurately model the movement of water and contaminants through the facility. Key to this effort is an accurate understanding of the hydraulic and physical properties of the solidified saltstone. To date, limited testing has been conducted to characterize the saltstone. The primary focus of this task was to estimate the hydraulic and physical properties of three types of saltstone and two vault concretes. The saltstone formulations included saltstone premix batched with (1) Deliquification, Dissolution, and Adjustment (DDA) salt simulant (w/pm 0.60), (2) Actinide Removal Process (ARP)/Modular Caustic Side Solvent Extraction Unit (MCU) salt simulant (w/pm 0.60), and (3) Salt Waste Processing Facility (SWPF) salt simulant (w/pm 0.60). The vault concrete formulations tested included the Vault 1/4 concrete and two variations of the Vault 2 concrete (Mix 1 and Mix 2). Wet properties measured for the saltstone formulations included yield stress, plastic viscosity, wet unit weight, bleed water volume, gel time, set time, and heat of hydration. Hydraulic and physical properties measured on the cured saltstone and concrete samples included saturated hydraulic conductivity, moisture retention, compressive strength, porosity, particle density, and dry bulk density. These properties were determined following a minimum 28 day ...
Date: November 25, 2008
Creator: Dixon, K; John Harbour, J & Mark Phifer, M
Partner: UNT Libraries Government Documents Department

PARTNERSHIP FOR THE DEVELOPMENT OF NEXT GENERATION SIMULATION TOOLS TO EVALUATE CEMENTITIOUS BARRIERS AND MATERIALS USED IN NUCLEAR APPLICATION - 8388

Description: The US DOE has initiated a multidisciplinary cross cutting project to develop a reasonable and credible set of tools to predict the structural, hydraulic and chemical performance of cement barriers used in nuclear applications over extended time frames (e.g., > 100 years for operating facilities and > 1000 years for waste management). A partnership that combines DOE, NRC, academia, private sector, and international expertise has been formed to accomplish the project objectives by integrating existing information and realizing advancements where necessary. The set of simulation tools and data developed under this project will be used to evaluate and predict the behavior of cementitious barriers used in near surface engineered waste disposal systems, e.g., waste forms, containment structures, entombments and environmental remediation, including decontamination and decommissioning (D&D) activities. The simulation tools will also support analysis of structural concrete components of nuclear facilities (spent fuel pools, dry spent fuel storage units, and recycling facilities, e.g., fuel fabrication, separations processes). Simulation parameters will be obtained from prior literature and will be experimentally measured under this project, as necessary, to demonstrate application of the simulation tools for three prototype applications (waste form in concrete vault, high level waste tank grouting, and spent fuel pool). Test methods and data needs to support use of the simulation tools for future applications will be defined. This is a national issue that affects all waste disposal sites that use cementitious waste forms and structures, decontamination and decommissioning activities, service life determination of existing structures, and design of future public and private nuclear facilities. The problem is difficult because it requires projecting conditions and responses over extremely long times. Current performance assessment analyses show that engineered barriers are typically the primary control to prevent the release of radionuclides from nuclear facilities into the environment. In the absence of an ...
Date: January 29, 2008
Creator: Langton, C & Richard Dimenna, R
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 Decontaminated 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 maximum concentration of Isopar{reg_sign} L in the DSS to assure 25% of the lower flammable limit is not exceeded has been determined to be about 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 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 amount of Isopar{reg_sign} L released versus time can be treated as a percentage of initial amount present; there was no statistically significant dependence of the release rate on the initial concentration. 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 rate is larger than at lower temperatures. Initial curing temperature was found to be very important as slight variations during the first few hours ...
Date: July 31, 2008
Creator: Zamecnik, J; Michael Bronikowski, M; Alex Cozzi, A; Russell Eibling, R & Charles Nash, C
Partner: UNT Libraries Government Documents Department

MEASUREMENT OF SPECIFIC HEAT CAPACITY OF SALTSTONE

Description: One of the goals of the Saltstone variability study is to identify (and quantify the impact of) the operational and compositional variables that control or influence the important processing and performance properties of Saltstone grout mixtures. The heat capacity of the Saltstone waste form is one of the important properties of Saltstone mixes that was last measured at SRNL in 1997. It is therefore important to develop a core competency for rapid and accurate analysis of the specific heat capacity of the Saltstone mixes in order to quantify the impact of compositional and operational variations on this property as part of the variability study. The heat capacity, coupled with the heat of hydration data obtained from isothermal calorimetry for a given Saltstone mix, can be used to predict the maximum temperature increase in the cells within the vaults of the Saltstone Disposal Facility (SDF). The temperature increase controls the processing rate and the pour schedule. The maximum temperature is also important to the performance properties of the Saltstone. For example, in mass pours of concrete or grout of which Saltstone is an example, the maximum temperature increase and the maximum temperature difference (between the surface and the hottest location) are controlled to ensure durability of the product and prevent or limit the cracking caused by the thermal gradients produced during curing. This report details the development and implementation of a method for the measurement of the heat capacities of Saltstone mixes as well as the heat capacities of the cementitious materials of the premix and the simulated salt solutions used to batch the mixes. The developed method utilizes the TAM Air isothermal calorimeter and takes advantage of the sophisticated heat flow measurement capabilities of the instrument. Standards and reference materials were identified and used to validate the procedure and ensure ...
Date: September 29, 2008
Creator: Harbour, J & Vickie Williams, V
Partner: UNT Libraries Government Documents Department

MEASUREMENT OF WASTE LOADING IN SALTSTONE

Description: One of the goals of the Saltstone variability study is to identify the operational and compositional variables that control or influence the important processing and performance properties of Saltstone grout mixtures. One of those properties of importance is the Waste Loading (WL) of the decontaminated salt solution (DSS) in the Saltstone waste form. Waste loading is a measure of the amount of waste that can be incorporated within a waste form. The value of the Saltstone waste loading ultimately determines the number of vaults that will be required to disposition all of the DSS. In this report, the waste loading is defined as the volume in milliliters of DSS per liter of Saltstone waste form. The two most important parameters that determine waste loading for Saltstone are water to cementitious material (w/cm) ratio and the cured grout density. Data are provided that show the dependence of waste loading on the w/cm ratio for a fixed DSS composition using the current premix material (45% Blast Furnace Slag (BFS), 45% Fly Ash (FA) and 10% Ordinary Portland Cement (OPC)). The impact of cured grout density on waste loading was also demonstrated. Mixes (at 0.60 w/cm) made with a Modular Caustic side extraction Unit (MCU) simulant and either OPC or BFS have higher cured grout densities than mixes made with premix and increase the WL to 709 mL/L for the OPC mix and 689 mL/L for the BFS mix versus the value of 653 mL/L for MCU in premix at 0.60 w/cm ratio. Bleed liquid reduces the waste loading and lowers the effective w/cm ratio of Saltstone. A method is presented (and will be used in future tasks) for correcting the waste loading and the w/cm ratio of the as-batched mixes in those cases where bleed liquid is present. For example, the Deliquification, ...
Date: July 18, 2008
Creator: Harbour, J & Vickie Williams, V
Partner: UNT Libraries Government Documents Department

Uranium Metal Reaction Behavior in Water, Sludge, and Grout Matrices

Description: This report summarizes information and data on the reaction behavior of uranium metal in water, in water-saturated simulated and genuine K Basin sludge, and in grout matrices. This information and data are used to establish the technical basis for metallic uranium reaction behavior for the K Basin Sludge Treatment Project (STP). The specific objective of this report is to consolidate the various sources of information into a concise document to serve as a high-level reference and road map for customers, regulators, and interested parties outside the STP (e.g., external reviewers, other DOE sites) to clearly understand the current basis for the corrosion of uranium metal in water, sludge, and grout.
Date: September 25, 2008
Creator: Delegard, Calvin H. & Schmidt, Andrew J.
Partner: UNT Libraries Government Documents Department

CLOSURE OF HLW TANKS FORMULATION FOR A COOLING COIL GROUT

Description: The Tank Closure and Technology Development Groups are developing a strategy for closing the High Level Waste (HLW) tanks at the Savannah River Site (SRS). Two Type IV tanks, 17 and 20 in the F-Area Tank Farm, have been successfully filled with grout. Type IV tanks at SRS do not contain cooling coils; on the other hand, the majority of the tanks (Type I, II, III and IIIA) do contain cooling coils. The current concept for closing tanks equipped with cooling coils is to pump grout into the cooling coils to prevent pathways for infiltrating water after tank closure. This task addresses the use of grout to fill intact cooling coils present in most of the remaining HLW tanks on Site. The overall task was divided into two phases. Phase 1 focused on the development of a grout formulation (mix design) suitable for filling the HLW tank cooling coils. Phase 2 will be a large-scale demonstration of the filling of simulated cooling coils under field conditions using the cooling coil grout mix design recommended from Phase 1. This report summarizes the results of Phase 1, the development of the cooling coil grout formulation. A grout formulation is recommended for the full scale testing at Clemson Environmental Technology Laboratory (CETL) that is composed by mass of 90% Masterflow (MF) 816 (a commercially available cable grout) and 10% blast furnace slag, with a water to cementitious material (MF 816 + slag) ratio of 0.33. This formulation produces a grout that meets the fresh and cured grout requirements detailed in the Task Technical Plan (2). The grout showed excellent workability under continuous mixing with minimal change in rheology. An alternative formulation using 90% MF 1341 and 10% blast furnace slag with a water to cementitious material ratio of 0.29 is also acceptable and ...
Date: May 23, 2008
Creator: Harbour, J; Vickie Williams, V & Erich Hansen, E
Partner: UNT Libraries Government Documents Department