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Pulse jet mixed process vessels comprise a key element of the U.S. Department of Energy’s strategy to process millions of gallons of legacy nuclear waste slurries. Slurry suctioned into a pulse jet mixer (PJM) tube at the end of one pulse is pneumatically driven from the PJM toward the bottom of the vessel at the beginning of the next pulse, forming a jet. The jet front traverses the distance from nozzle outlet to the bottom of the vessel and spreads out radially. Varying numbers of PJMs are typically arranged in a ring configuration within the vessel at a selected radius and operated concurrently. Centrally directed radial flows from neighboring jets collide to create a central upwell that elevates the solids in the center of the vessel when the PJM tubes expel their contents. An essential goal of PJM operation is to elevate solids to the liquid surface to minimize stratification. Solids stratification may adversely affect throughput of the waste processing plant. Unacceptably high slurry densities at the base of the vessel may plug the pipeline through which the slurry exits the vessel. Additionally, chemical reactions required for processing may not achieve complete conversion. To avoid these conditions, a means of predicting the elevation to which the solids rise in the central upwell that can be used during vessel design remains essential. In this paper we present a simple criterion to evaluate the extent of solids elevation achieved by a turbulent upwell jet. The criterion asserts that at any location in the central upwell the local velocity must be in excess of a cutoff velocity to remain turbulent. We find that local velocities in excess of 0.6 m/s are necessary for turbulent jet flow through both Newtonian and yield stress slurries. By coupling this criterion with the free jet velocity equation relating the local velocity to elevation in the central upwell, we estimate the elevation at which turbulence fails, and consequently the elevation at which the upwell fails to further lift the slurry. Comparing this elevation to the vessel fill level predicts whether the jet flow will achieve the full vertical extent of the vessel at the center. This simple local-velocity criterion determines a minimum PJM nozzle velocity at which the full vertical extent of the central upwell in PJM vessels will be turbulent. The criterion determines a minimum because flow in regions peripheral to the central upwelling jet may not be turbulent, even when the center of the vessel in the upwell is turbulent, if the jet pulse duration is too short. The local-velocity criterion ensures only that there is sufficient wherewithal for the turbulent jet flow to drive solids to the surface in the center of the vessel in the central upwell.

Boehmite dissolution experiments were conducted at NaOH concentrations of 10 and 12 M to determine whether the increased aluminum solubility at high hydroxide concentration would offset the increase in added sodium, causing a decrease in the amount of sodium added during boehmite dissolution. A shrinking-core dissolution rate model with a reversible reaction was fitted to the test data. The resulting model included the effects of temperature, hydroxide concentration, and dissolved aluminum concentration. The rate was found to depend on the ∼1.5 power of hydroxide molarity. When the rate model was used to simulate batch boehmite dissolution, a concentration range of 7–9 M NaOH was found to minimize the mass of sodium needed to dissolve a given mass of aluminum, potentially reducing it by as much as two-thirds. The time required to dissolve the boehmite could be decreased by using hydroxide concentrations greater than ∼10 M.

This paper examines the electrochemical and direct internal steam–methane reforming performance of the solid oxide fuel cell when subjected to pressurization. Pressurized operation boosts the Nernst potential and decreases the activation polarization, both of which serve to increase cell voltage and power while lowering the heat load and operating temperature. A model considering the activation polarization in both the fuel and the air electrodes was adopted to address this effect on the electrochemical performance. The pressurized methane conversion kinetics and the increase in equilibrium methane concentration are considered in a new rate expression. The models were then applied in simulations to predict how the distributions of direct internal reforming rate, temperature, and current density are effected within stacks operating at elevated pressure. A generic 10 cm counter-flow stack model was created and used for the simulations of pressurized operation. The predictions showed improved thermal and electrical performance with increased operating pressure. The average and maximum cell temperatures decreased by 3% (20 °C) while the cell voltage increased by 9% as the operating pressure was increased from 1 to 10 atm.

The Hanford Waste Treatment and Immobilization Plant (WTP) is currently being designed and constructed to pretreat and vitrify a large portion of the waste in the 177 underground waste storage tanks at the Hanford Site. A number of technical issues related to the design of the pretreatment facility (PTF) of the WTP have been identified. These issues must be resolved prior to the U.S. Department of Energy (DOE) Office of River Protection (ORP) reaching a decision to proceed with engineering, procurement, and construction activities for the PTF. One of the issues is Technical Issue T1 - Hydrogen Gas Release from Vessels (hereafter referred to as T1). The focus of T1 is identifying controls for hydrogen release and completing any testing required to close the technical issue. In advance of selecting specific controls for hydrogen gas safety, a number of preliminary technical studies were initiated to support anticipated future testing and to improve the understanding of hydrogen gas generation, retention, and release within PTF vessels. These activities supported the development of a plan defining an overall strategy and approach for addressing T1 and achieving technical endpoints identified for T1. Preliminary studies also supported the development of a test plan for conductingmore » testing and analysis to support closing T1. Both of these plans were developed in advance of selecting specific controls, and in the course of working on T1 it was decided that the testing and analysis identified in the test plan were not immediately needed. However, planning activities and preliminary studies led to significant technical progress in a number of areas. This report summarizes the progress to date from the preliminary technical studies. The technical results in this report should not be used for WTP design or safety and hazards analyses and technical results are marked with the following statement: “Preliminary Technical Results for Planning – Not to be used for WTP Design or Safety Analyses.”« less

One of the events postulated in the hazard analysis at the Waste Treatment and Immobilization Plant (WTP) and other U.S. Department of Energy (DOE) nuclear facilities is a breach in process piping that produces aerosols with droplet sizes in the respirable range. The current approach for predicting the size and concentration of aerosols produced in a spray leak involves extrapolating from correlations reported in the literature. These correlations are based on results obtained from small engineered spray nozzles using pure liquids with Newtonian fluid behavior. The narrow ranges of physical properties on which the correlations are based do not cover the wide range of slurries and viscous materials that will be processed in the WTP and across processing facilities in the DOE complex. To expand the data set upon which the WTP accident and safety analyses were based, an aerosol spray leak testing program was conducted by Pacific Northwest National Laboratory (PNNL). PNNL’s test program addressed two key technical areas to improve the WTP methodology (Larson and Allen 2010). The first technical area was to quantify the role of slurry particles in small breaches where slurry particles may plug the hole and prevent high-pressure sprays. The results from an effort more » to address this first technical area can be found in Mahoney et al. (2012a). The second technical area was to determine aerosol droplet size distribution and total droplet volume from prototypic breaches and fluids, including sprays from larger breaches and sprays of slurries for which literature data are largely absent. To address the second technical area, the testing program collected aerosol generation data at two scales, commonly referred to as small-scale and large-scale. The small-scale testing and resultant data are described in Mahoney et al. (2012b) and the large-scale testing and resultant data are presented in Schonewill et al. (2012). In tests at both scales, simulants were used to mimic the relevant physical properties projected for actual WTP process streams. « less

One of the events postulated in the hazard analysis at the Waste Treatment and Immobilization Plant (WTP) and other U.S. Department of Energy (DOE) nuclear facilities is a breach in process piping that produces aerosols with droplet sizes in the respirable range. The current approach for predicting the size and concentration of aerosols produced in a spray leak involves extrapolating from correlations reported in the literature. These correlations are based on results obtained from small engineered spray nozzles using pure liquids with Newtonian fluid behavior. The narrow ranges of physical properties on which the correlations are based do not cover the wide range of slurries and viscous materials that will be processed in the WTP and across processing facilities in the DOE complex. Two key technical areas were identified where testing results were needed to improve the technical basis by reducing the uncertainty due to extrapolating existing literature results. The first technical need was to quantify the role of slurry particles in small breaches where the slurry particles may plug and result in substantially reduced, or even negligible, respirable fraction formed by high-pressure sprays. The second technical need was to determine the aerosol droplet size distribution and volume from prototypicmore » breaches and fluids, specifically including sprays from larger breaches with slurries where data from the literature are scarce. To address these technical areas, small- and large-scale test stands were constructed and operated with simulants to determine aerosol release fractions and generation rates from a range of breach sizes and geometries. The properties of the simulants represented the range of properties expected in the WTP process streams and included water, sodium salt solutions, slurries containing boehmite or gibbsite, and a hazardous chemical simulant. The effect of anti-foam agents was assessed with most of the simulants. Orifices included round holes and rectangular slots. The round holes ranged in size from 0.2 to 4.46 mm. The slots ranged from (width × length) 0.3 × 5 to 2.74 × 76.2 mm. Most slots were oriented longitudinally along the pipe, but some were oriented circumferentially. In addition, a limited number of multi-hole test pieces were tested in an attempt to assess the impact of a more complex breach. Much of the testing was conducted at pressures of 200 and 380 psi, but some tests were conducted at 100 psi. Testing the largest postulated breaches was deemed impractical because of the large size of some of the WTP equipment. This report presents the experimental results and analyses for the aerosol measurements obtained in the small-scale test stand. It includes a description of the simulants used and their properties, equipment and operations, data analysis methodologies, and test results. The results of tests investigating the role of slurry particles in plugging small breaches are reported in Mahoney et al. (2012). The results of the aerosol measurements in the large-scale test stand are reported in Schonewill et al. (2012) along with an analysis of the combined results from both test scales.« less

One of the events postulated in the hazard analysis at the Waste Treatment and Immobilization Plant (WTP) and other U.S. Department of Energy (DOE) nuclear facilities, is a breach in process piping that produces aerosols with droplet sizes in the respirable range. The current approach for predicting the size and concentration of aerosols produced in a spray leak involves extrapolating from correlations published in the literature. These correlations are based on results obtained from small engineered spray nozzles using pure liquids with Newtonian fluid behavior. The narrow ranges of physical properties on which the correlations are based do not cover the wide range of slurries and viscous materials present in the WTP and across processing facilities in the DOE complex. Two key technical areas were identified where testing results were needed to improve the technical basis by reducing the uncertainty introduced by extrapolating existing literature results. The first technical need was to quantify the role of slurry particles in small breaches in which the slurry particles may plug and result in substantially reduced, or even negligible, respirable fraction formed by high pressure sprays. The second technical need was to determine the aerosol droplet size distribution and volume from prototypic breaches more » and fluids, specifically including sprays from larger breaches with slurries where data from the literature are largely absent. To address these technical areas, small- and large-scale test stands were constructed and operated with simulants to determine the aerosol release fractions and aerosol generation rates from a range of breach sizes and geometries. The properties of the simulants represented the range of properties expected in the WTP process streams and included water, sodium salt solutions, slurries containing boehmite or gibbsite, and a hazardous chemical simulant. The effect of anti-foam agents (AFA) was assessed with most of the simulants. Orifices included round holes and rectangular slots. Much of the testing was conducted at pressures of 200 and 380 psi, but some tests were conducted at 100 psi. Testing the largest postulated breaches was deemed impractical because of the large size of some of the WTP equipment. The purpose of the study described in this report is to provide experimental data for the first key technical area, potential plugging of small breaches, by performing small-scale tests with a range of orifice sizes and orientations representative of the WTP conditions. The simulants used were chosen to represent the range of process stream properties in the WTP. Testing conducted after the plugging tests in the small- and large-scale test stands addresses the second key technical area, aerosol generation. The results of the small-scale aerosol generation tests are included in Mahoney et al. 2012. The area of spray generation from large breaches is covered by large-scale testing in Schonewill et al. 2012. « less

The Hanford Site in Washington State manages 177 underground storage tanks containing approximately 250,000 m3 of waste generated during past defense reprocessing and waste management operations. These tanks contain a mixture of sludge, saltcake and supernatant liquids. The insoluble sludge fraction of the waste consists of metal oxides and hydroxides and contains the bulk of many radionuclides such as the transuranic components and 90Sr. The saltcake, generated by extensive evaporation of aqueous solutions, consists primarily of dried sodium salts. The supernates consist of concentrated (5-15 M) aqueous solutions of sodium and potassium salts. The 177 storage tanks include 149 single-shell tanks (SSTs) and 28 double -hell tanks (DSTs). Ultimately the wastes need to be retrieved from the tanks for treatment and disposal. The SSTs contain minimal amounts of liquid wastes, and the Tank Operations Contractor is continuing a program of moving solid wastes from SSTs to interim storage in the DSTs. The Hanford DST system provides the staging location for waste feed delivery to the Department of Energy (DOE) Office of River Protection’s (ORP) Hanford Tank Waste Treatment and Immobilization Plant (WTP). The WTP is being designed and constructed to pretreat and then vitrify a large portion of the wastesmore » in Hanford’s 177 underground waste storage tanks.« less

Boehmite leaching tests were carried out at NaOH concentrations of 10 M and 12 M, temperatures of 85°C and 60°C, and a range of initial aluminate concentrations. These data, and data obtained during earlier 100°C tests using 1 M and 5 M NaOH, were used to establish the dependence of the boehmite dissolution rate on hydroxide concentration, temperature, and initial aluminate concentration. A semi-empirical kinetic model for boehmite leaching was fitted to the data and used to calculate the NaOH additions required for leaching at different hydroxide concentrations. The optimal NaOH concentration for boehmite leaching at 85°C was estimated, based on minimizing the amount of Na that had to be added in NaOH to produce a given boehmite conversion.

The current System Plan for the Hanford Tank Farms uses relaxed buoyant displacement gas release event (BDGRE) controls for deep sludge (i.e., high level waste [HLW]) tanks, which allows the tank farms to use more storage space, i.e., increase the sediment depth, in some of the double-shell tanks (DSTs). The relaxed BDGRE controls are based on preliminary analysis of a gas release model from van Kessel and van Kesteren. Application of the van Kessel and van Kesteren model requires parametric information for the sediment, including the lateral earth pressure at rest and shear modulus. No lateral earth pressure at rest and shear modulus in situ measurements for Hanford sludge are currently available. The two chemical sludge simulants will be used in follow-on work to experimentally measure the van Kessel and van Kesteren model parameters, lateral earth pressure at rest, and shear modulus.

Pacific Northwest National Laboratory (PNNL) was tasked by Bechtel National Inc. (BNI) on the River Protection Project, Hanford Tank Waste Treatment and Immobilization Plant (RPP-WTP) project to conduct testing to demonstrate the performance of the WTP Pretreatment Facility (PTF) leaching and ultrafiltration processes at an engineering-scale. In addition to the demonstration, the testing was to address specific technical issues identified in Issue Response Plan for Implementation of External Flowsheet Review Team (EFRT) Recommendations - M12, Undemonstrated Leaching Processes.( ) Testing was conducted in a 1/4.5-scale mock-up of the PTF ultrafiltration system, the Pretreatment Engineering Platform (PEP). Parallel laboratory testing was conducted in various PNNL laboratories to allow direct comparison of process performance at an engineering-scale and a laboratory-scale. This report presents and discusses the results of those tests.

Testing Summary Pacific Northwest National Laboratory (PNNL) has been tasked by Bechtel National Inc. (BNI) on the River Protection Project-Hanford Tank Waste Treatment and Immobilization Plant (RPP-WTP) project to perform research and development activities to resolve technical issues identified for the Pretreatment Facility (PTF). The Pretreatment Engineering Platform (PEP) was designed and constructed and is to be operated as part of a plan to respond to issue M12, “Undemonstrated Leaching Processes.” The PEP is a 1/4.5-scale test platform designed to simulate the WTP pretreatment caustic leaching, oxidative leaching, ultrafiltration solids concentration, and slurry washing processes. The PEP replicates the WTP leaching processes using prototypic equipment and control strategies. The PEP also includes non-prototypic ancillary equipment to support the core processing. Two operating scenarios are currently being evaluated for the ultrafiltration process (UFP) and leaching operations. The first scenario has caustic leaching performed in the UFP-2 ultrafiltration feed vessels (i.e., vessel UFP-VSL-T02A in the PEP and vessels UFP-VSL-00002A and B in the WTP PTF). The second scenario has caustic leaching conducted in the UFP-1 ultrafiltration feed preparation vessels (i.e., vessels UFP-VSL-T01A and B in the PEP; vessels UFP-VSL-00001A and B in the WTP PTF). In both scenarios, 19-M sodium hydroxide solution (NaOH, caustic) is added to the waste slurry in the vessels to leach solid aluminum compounds (e.g., gibbsite, boehmite). Caustic addition is followed by a heating step that uses direct injection of steam to accelerate the leaching process. Following the caustic leach, the vessel contents are cooled using vessel cooling jackets and/or external heat exchangers. The main difference between the two scenarios is that for leaching in UFP-1, the 19-M NaOH is added to un-concentrated waste slurry (3 to 8 wt% solids), while for leaching in UFP-2, the slurry is concentrated to nominally 20 wt% solids using cross-flow ultrafiltration before adding caustic. The work described in this report addresses the kinetics of caustic leach under WTP conditions, based on tests performed with a Hanford waste simulant. The tests were completed at the lab-scale and in the PEP, which is a 1/4.5-scale mock-up of key PTF process equipment. The purpose of this report is to summarize the results from both scales that are related to caustic leach chemistry to support a scale-up factor for the submodels to be used in the G2 model, which predicts WTP operating performance. The scale-up factor will take the form of an adjustment factor for the rate constant in the boehmite leach kinetic equation in the G2 model.

The U.S. Department of Energy (DOE) Office of River Protection’s Waste Treatment Plant (WTP) will process and treat radioactive waste that is stored in tanks at the Hanford Site. The waste treatment process in the pretreatment facility will mix both Newtonian and non-Newtonian slurries in large process tanks. Process vessels mixing non-Newtonian slurries will use pulse jet mixers (PJMs), air sparging, and recirculation pumps. An anti-foam agent (AFA) will be added to the process streams to prevent surface foaming, but may also increase gas holdup and retention within the slurry. The work described in this report addresses gas retention and release in simulants with AFA through testing and analytical studies. Gas holdup and release tests were conducted in a 1/4-scale replica of the lag storage vessel operated in the Pacific Northwest National Laboratory (PNNL) Applied Process Engineering Laboratory using a kaolin/bentonite clay and AZ-101 HLW chemical simulant with non-Newtonian rheological properties representative of actual waste slurries. Additional tests were performed in a small-scale mixing vessel in the PNNL Physical Sciences Building using liquids and slurries representing major components of typical WTP waste streams. Analytical studies were directed at discovering how the effect of AFA might depend on gas composition and predicting the effect of AFA on gas retention and release in the full-scale plant, including the effects of mass transfer to the sparge air. The work at PNNL was part of a larger program that included tests conducted at Savannah River National Laboratory (SRNL) that is being reported separately. SRNL conducted gas holdup tests in a small-scale mixing vessel using the AZ-101 high-level waste (HLW) chemical simulant to investigate the effects of different AFAs, their components, and of adding noble metals. Full-scale, single-sparger mass transfer tests were also conducted at SRNL in water and AZ-101 HLW simulant to provide data for PNNL’s WTP gas retention and release modeling.

The U.S. Department of Energy (DOE) Office of River Protection’s Waste Treatment and Immobilization Plant (WTP) will process and treat radioactive waste that is stored in tanks at the Hanford Site. Piping, pumps, and mixing vessels have been selected to transport, store, and mix the high-level waste slurries in the WTP. This report addresses the analyses performed by the Rheology Working Group (RWG) and Risk Assessment Working Group composed of Pacific Northwest National Laboratory (PNNL), Bechtel National Inc. (BNI), CH2M HILL, DOE Office of River Protection (ORP) and Yasuo Onishi Consulting, LLC staff on data obtained from documented Hanford waste analyses to determine a best-estimate of the rheology of the Hanford tank wastes and their settling behavior. The actual testing activities were performed and reported separately in referenced documentation. Because of this, many of the required topics below do not apply and are so noted.

“Designing in Safety” is a desired part of the development of any new potentially hazardous system, process, or facility. It is a required part of nuclear safety activities as specified in the U.S. Department of Energy (DOE) Order 420.1B, Facility Safety. This order addresses the design of nuclear related facilities developed under federal regulation 10CFR830, Nuclear Safety Management. 10CFR830 requires that safety basis documentation be provided to identify how nuclear safety is being adequately addressed as a condition for system operation (e.g., the safety basis). To support the development of the safety basis, a safety analysis is performed. Although the concept of developing a design that addresses safety is simple, the execution can be complex and challenging. This paper addresses those complexities and challenges for the design activity of a system to treat sludge, a corrosion product of spent nuclear fuel, at DOE’s Hanford Site in Washington State. The system being developed is referred to as the Sludge Treatment Project (STP). This paper describes the portion of the safety analysis that addresses the selection of design basis events using the experience gained from the STP and the development of design requirements for safety features associated with those events. Specifically, the paper describes the safety design process and the application of the process for two types of potential design basis accidents associated with the operation of the system, 1) flashing spray leaks and 2) splash and splatter leaks. Also presented are the technical challenges that are being addressed to develop effective safety features to deal with these design basis accidents.

The toxicological source terms used for potential accident assessment in the Hanford Tank Farms DSA are based on toxicological sums-of-fractions (SOFs) that were calculated based on the Best Basis Inventory (BBI) from May 2002, using a method that depended on thermodynamic equilibrium calculations of the compositions of liquid and solid phases. The present report describes a simplified SOF-calculation method that is to be used in future toxicological updates and assessments and compares its results (for the 2002 BBI) to those of the old method.

Recent studies conducted for the Hanford Waste Treatment Plant (WTP) found that the centrifuged solids from tanks 241-AW-101, 241-AN-107, and 241-C-104 contained two carcinogenic chemicals, 1-naphthylamine and 2,2-bis(4-chlorophenyl)-1,1-dichloroethene (DDE), that had never been detected in tank headspaces. This report estimates the potential headspace concentrations associated with these two compounds. The calculation was based on a comparison of headspace concentrations and waste concentrations for sixteen other organic compounds in the passively-ventilated tank 241-C-104. An approximate relation was found between the compounds? solubilities in water and their headspace concentrations. The relation was used to estimate the 241-C-104 headspace concentration that would result from the maximum measured waste concentrations of DDE and 1-naphthylamine. On the basis of the assumptions made in this report about organic compound transport and equilibration, the DDE concentration was estimated at well below one part per trillion, below standard analytical detection limits. However, 1-naphthylamine could potentially be present in the headspace of a passively ventilated tank at the 30 ppb level.

Hanford Site processes associated with the chemical separation of plutonium from uranium and other fission products produced a variety of volatile, semivolatile, and nonvolatile organic and inorganic waste chemicals that were sent to high-level waste tanks. These chemicals have undergone and continue to undergo radiolytic and thermal reactions in the tanks to produce a wide variety of degradation reaction products. The origins of the organic wastes, the chemical reactions they undergo, and their reaction products have recently been examined by Stock (2004). Stock gives particular attention to explaining the presence of various types of volatile and semivolatile organic species identified in headspace air samples. This report complements the Stock report by examining the storage of volatile and semivolatile species in the waste, their transport through any overburden of waste to the tank headspaces, the physical phenomena affecting their concentrations in the headspaces, and their eventual release into the atmosphere above the tanks.

The objectives of this study were to evaluate mixing of liquid waste from Tank 241-S-112 with waste in Tank 241-SY-101 and to determine the properties of the resulting waste for the cross-site transfer to avoid potential double-shell tank corrosion and pipeline plugging. We applied the time-varying, three-dimensional computer code TEMPEST to Tank SY-101 as it received the S-112 liquid waste. The model predicts that temperature variations in Tank SY-101 generate a natural convection flow that is very slow, varying from about 7 x 10{sup -5} to 1 x 10{sup -3} ft/sec (0.3 to about 4 ft/hr) in most areas. Thus, natural convection would eventually mix the liquid waste in SY-101 but would be very slow to achieve nearly complete mixing. These simulations indicate that the mixing of S-112 and SY-101 wastes in Tank SY-101 is a very slow process, and the density difference between the two wastes would further limit mixing. It is expected to take days or weeks to achieve relatively complete mixing in Tank SY-101.

CH2M HILL Hanford Group, Inc. (CH2M HILL) is in the process of identifying and developing supplemental process technologies to accelerate the tank waste cleanup mission. A range of technologies is being evaluated to allow disposal of Hanford waste types, including transuranic (TRU) process wastes. Ten Hanford single-shell tanks (SSTs) have been identified whose contents may meet the criteria for designation as TRU waste: the B-200 series (241-B-201, -B-202, -B 203, and B 204), the T-200 series (241-T-201, T 202, -T-203, and -T-204), and Tanks 241-T-110 and -T-111. CH2M HILL has requested vendor proposals to develop a system to transfer and package the contact-handled TRU (CH-TRU) waste retrieved from the SSTs for subsequent disposal at the Waste Isolation Pilot Plant (WIPP). Current plans call for a modified ''dry'' retrieval process in which a liquid stream is used to help mobilize the waste for retrieval and transfer through lines and vessels. This retrieval approach requires that a significant portion of the liquid be removed from the mobilized waste sludge in a ''dewatering'' process such as centrifugation prior to transferring to waste packages in a form suitable for acceptance at WIPP. In support of CH2M HILL's effort to procure a TRU waste handlingmore » and packaging process, Pacific Northwest National Laboratory (PNNL) developed waste simulant formulations to be used in evaluating the vendor's system. For the SST CH-TRU wastes, the suite of simulants includes (1) nonradioactive chemical simulants of the liquid fraction of the waste, (2) physical simulants that reproduce the important dewatering properties of the waste, and (3) physical simulants that can be used to mimic important rheological properties of the waste at different points in the TRU waste handling and packaging process. To validate the simulant formulations, their measured properties were compared with the limited data for actual TRU waste samples. PNNL developed the final simulant formulations through a series of screening and scoping experimental studies, ultimately completing a set of physical simulant validation studies that were conducted in accordance with NQA-1 requirements for quality assurance. Only simple, commercially available feedstock is used in the simulant formulations. Readily available kaolin and bentonite clays mixed with water in specific proportions are the basis of the physical simulants, and the liquid simulants contain no exotic chemicals.« less

CH2M HILL Hanford Group, Inc. (CH2M HILL) is in the process of identifying and developing supplemental process technologies to accelerate the tank waste cleanup mission. One technology targets disposal of Hanford transuranic (TRU) process wastes stored in single-shell tanks (SSTs). Ten Hanford SSTs are candidates for designation as contact-handled TRU waste type: the B-200 series tanks (241-B-201, -B-202, -B-203, and -B-204), the T-200 series tanks (241-T-201, -T-202, -T-203, and -T-204), and Tanks 241-T-110 and T-111. Current plans identify a process in which these wastes are retrieved from the tanks, either dry or with a recycled liquid stream to help mobilize the waste in the tank and through transfer lines and vessels, dewatered to remove excess liquid, and transferred to waste packages in a form suitable for disposal. CH2M HILL seeks to procure a process for dewatering, handling, and packaging the contact-handled TRU waste after it is retrieved. An understanding of waste physical properties is needed to support design of the SST TRU handling and packaging system and to produce suitable physical simulants to test such a process. Pacific Northwest National Laboratory (PNNL) has been tasked with developing these waste simulants. This report summarizes PNNL's assessment of available waste physical propertymore » information for the 10 candidate TRU SSTs. Data sources include the Hanford Tank Waste Information Network System (TWINS) database, technical reports, and visual observations from the review of photographs and videotape recordings taken during the extrusion of various SST TRU waste core samples. While the retrieval process is expected to alter certain waste physical properties such as shear strength, the effects of this process on waste properties cannot yet be quantified. Therefore, the scope of this report is to describe the properties of SST TRU wastes as they are known for unprocessed wastes or, in some cases, for diluted waste samples. The report focuses on the waste rheology, settling characteristics, particle properties, bulk density, and water content. Because some physical properties of the waste are related to the waste chemistry (e.g., particle hardness), a brief overview of the process and chemistry resulting in the TRU waste is provided. Other properties, including the potential for gas retention in the waste, are discussed briefly. Qualitative descriptions of how waste properties might be affected by retrieval and subsequent processing are also noted.« less

CH2M HILL Hanford Group, Inc. is identifying and developing supplemental process technologies to accelerate the Hanford tank waste cleanup mission. Bulk vitrification, containerized grout, and steam reforming are three technologies under consideration for treatment of the radioactive saltcake wastes in 68 single-shell tanks. To support development and testing of these technologies, Pacific Northwest National Laboratory (PNNL) was tasked with developing a cold dissolved saltcake simulant formulation to be representative of an actual saltcake waste stream, preparing 25- and 100-L batches of the simulant, and analyzing the composition of the batches to ensure conformance to formulation targets. Lacking a defined composition for dissolved actual saltcake waste, PNNL used available tank waste composition information and an equilibrium chemistry model (Environmental Simulation Program [ESP{trademark}]) to predict the concentrations of analytes in solution. Observations of insoluble solids in initial laboratory preparations for the model-predicted formulation prompted reductions in the concentration of phosphate and silicon in the final simulant formulation. The analytical results for the 25- and 100-L simulant batches, prepared by an outside vendor to PNNL specifications, agree within the expected measurement accuracy ({approx}10%) of the target concentrations and are highly consistent for replicate measurements, with a few minor exceptions. In parallel with themore » production of the 2nd simulant batch (100-L), a 1-L laboratory control sample of the same formulation was carefully prepared at PNNL to serve as an analytical standard. The instrumental analyses indicate that the vendor prepared batches of solution adequately reflect the as-formulated simulant composition. In parallel with the simulant development effort, a nominal 5-M (molar) sodium actual waste solution was prepared at the Hanford Site from a limited number of tank waste samples. Because this actual waste solution w as also to be used for testing the supplemental treatment technologies, the modeled simulant formulation was predicated on the composite of waste samples used to prepare it. Subsequently, the actual waste solution was filtered and pretreated to remove radioactive cesium at PNNL and then analyzed using the same instrumentation and procedures applied to the simulant samples. The overall agreement of measured simulant and actual waste solution compositions is better than {+-}10% for the most concentrated species including sodium, nitrate, hydroxide, carbonate, and nitrite. While the magnitude of the relative difference in the simulant and actual waste composition is large (>20% difference) for a few analytes (aluminum, chromium, fluoride, potassium, and total organic carbon), the absolute differences in concentration are in general not appreciable. Our evaluation is that these differences in simulant and actual waste solutions should have a negligible impact on bulk vitrification and containerized grout process testing, while the impact of the low aluminum concentration on steam reforming is yet to be determined.« less

The waste in Hanford tanks generates a mixture of flammable gases and releases it into the tank headspace. The potential hazard resulting from flammable gas generation requires that controls be established to prevent ignition and halt operations if gas concentrations reach levels of concern. In cases where only hydrogen is monitored, a control limit of 6,250 ppm hydrogen has been in use at Hanford for several years. The hydrogen-based control limit is intended to conservatively represent 25% of the lower flammability limit of a gas mixture, accounting for the presence of flammable gases other than hydrogen, with ammonia being the primary concern. This report reviews the technical basis of the current control limit based on observed and projected concentrations of hydrogen and ammonia representing a range of gas release scenarios. The conclusion supports the continued use of the current 6,250 ppm hydrogen control limit

This report was prepared as part of a task supporting the deployment of the retained gas sampler (RGS) system in Flammable Gas Watch List Tanks. The emphasis of this report is on presenting supplemental information about the ammonia measurements resulting from retained gas sampling of Tanks 241-AW-101, A-101, AN-105, AN-104, AN-103, U-103, S-106, BY-101, BY-109, SX-106, AX-101, S-102, S-111, U-109, and SY-101. This information provides a better understanding of the accuracy of past RGS ammonia measurements, which will assist in determining flammable and toxicological hazards.

This report presents an overview of what is known about the flammability of the gases generated and retained in Hanford waste tanks in terms of the gas composition, the flammability and detonability limits of the gas constituents, and the availability of ignition sources. The intrinsic flammability (or nonflammability) of waste gas mixtures is one major determinant of whether a flammable region develops in the tank headspace; other factors are the rate, surface area, volume of the release, and the tank ventilation rate, which are not covered in this report.

The flammable gas hazard in Hanford waste tanks was made an issue by the behavior of double-shell Tank (DST) 241-SY-101 (SY-101). Shortly after SY-101 was filled in 1980, the waste level began rising periodically, due to the generation and retention of gases within the slurry, and then suddenly dropping as the gases were released. An intensive study of the tank's behavior revealed that these episodic releases posed a safety hazard because the released gas was flammable, and, in some cases, the volume of gas released was sufficient to exceed the lower flammability limit (LFL) in the tank headspace (Allemann et al. 1993). A mixer pump was installed in SY-101 in late 1993 to prevent gases from building up in the settled solids layer, and the large episodic gas releases have since ceased (Allemann et al. 1994; Stewart et al. 1994; Brewster et al. 1995). However, the surface level of SY-101 has been increasing since at least 1995, and in recent months the level growth has shown significant and unexpected acceleration. Based on a number of observations and measurements, including data from the void fraction instrument (VFI), we have concluded that the level growth is caused largely by increased gas retention in the floating crust. In September 1998, the crust contained between about 21 and 43% void based on VFI measurements (Stewart et al. 1998). Accordingly, it is important to understand the dominant mechanisms of gas retention, why the gas retention is increasing, and whether the accelerating level increase will continue, diminish or even reverse. It is expected that the retained gas in the crust is flammable, with hydrogen as a major constituent. This gas inventory would pose a flammable gas hazard if it were to release suddenly. In May 1997, the mechanisms of bubble retention and release from crust material were the subject of a workshop. The evaluation of the crust and potential hazards assumed a more typical void of roughly 15% gas. It could be similar to percolati on in single-shell tank (SST) waste forms. The much higher void being currently observed in SY-101 represents essentially a new crust configuration, and the mechanisms for sudden gas release need to be evaluated. The purpose of this study is to evaluate the situation of gas bubbles in crust based on the previous work on gas bubble retention, migration, and release in simulants and actual waste. We have also conducted some visual observations of bubble migration through simulated crusts to help understand the interaction of the various mechanisms.

Four double-contained receiver tanks (DCRTs) at Hanford will be used to store salt-well pumped liquids from tanks on the Flammable Gas Watch List. This document was created to serve as a reference document describing the current knowledge of flammable gas issues in DCRTs. The document identifies, describes, evaluates, and attempts to quantify potential gas carryover and release mechanisms. It estimates several key parameters needed for these calculations, such as initial aqueous concentrations and ventilation rate, and evaluates the uncertainty in those estimates. It justifies the use of the Schumpe model for estimating vapor-liquid equilibrium constants. It identifies several potential waste compatibility issues (such as mixing and pH or temperature changes) that could lead to gas release and provides a basis for calculating their effects. It evaluates the potential for gas retention in precipitated solids within a DCRT and whether retention could lead to a buoyant displacement instability (rollover) event. It discusses rates of radiolytic, thermal, and corrosive hydrogen generation within the DCRT. It also describes in detail the accepted method of calculating the lower flammability limit (LFL) for mixtures of flammable gases. The report incorporates these analyses into two models for calculating headspace flammability, one based on instantaneous equilibrium between dissolved gases and the headspace and one incorporating limited release rates based on mass-transfer considerations. Finally, it demonstrates the use of both models to estimate headspace flammable gas concentrations and minimum ventilation rates required to maintain concentrations below 25% of the LFL.

Four double-contained receiver tanks (DCRTs) at Hanford will be used to store salt-well pumped liquids from tanks on the Flammable Gas Watch List. This document was created to serve as a technical basis or reference document for flammable gas issues in DCRTs. The document identifies, describes, evaluates, and attempts to quantify potential gas carryover and release mechanisms. It estimates several key parameters needed for these calculations, such as initial aqueous concentrations and ventilation rate, and evaluates the uncertainty in those estimates. It justifies the use of the Schumpe model for estimating vapor-liquid equilibrium constants. It identifies several potential waste compatibility issues (such as mixing and pH or temperature changes) that could lead to gas release and provides a basis for calculating their effects. It evaluates the potential for gas retention in precipitated solids within a DCRT and whether retention could lead to a buoyant displacement instability (rollover) event. It discusses rates of radiolytic, thermal, and corrosive hydrogen generation within the DCRT. It also describes in detail the accepted method of calculating the lower flammability limit (LFL) for mixtures of flammable gases.

This report describes the current understanding of flammable gas retention and release in Hanford single-shell waste tanks based on theory, experimental results, and observations of tank behavior. The single-shell tanks likely to pose a flammable gas hazard are listed and described, and photographs of core extrusions and the waste surface are included. The credible mechanisms for significant flammable gas releases are described, and release volumes and rates are quantified as much as possible. The only mechanism demonstrably capable of producing large ({approximately}100 m{sup 3}) spontaneous gas releases is the buoyant displacement, which occurs only in tanks with a relatively deep layer of supernatant liquid. Only the double-shell tanks currently satisfy this condition. All release mechanisms believed plausible in single-shell tanks have been investigated, and none have the potential for large spontaneous gas releases. Only small spontaneous gas releases of several cubic meters are likely by these mechanisms. The reasons several other postulated gas release mechanisms are implausible or incredible are also given.