108 Matching Results

Search Results

Advanced search parameters have been applied.

Pourbaix diagram for the prediction of waste glass durability in geologic environments

Description: Dissolution of nuclear waste glass occurs by corrosion mechanisms similar to those of metallurgical and mineralogic systems albeit on different time scales. The effects of imposed pH and oxidation potential (Eh) conditions existing in natural environments on metals and minerals have been quantitatively and phenomenologically described in compendiums of Pourbaix (pH-potential) diagrams. Construction of Pourbaix diagrams to quantify the response of nuclear waste glasses to repository specific pH and Eh conditions is demonstrated. The expected long-term effects of groundwater contact on the durability of nuclear waste glasses can then be unified. 40 refs., 4 figs., 1 tab.
Date: January 1, 1987
Creator: Jantzen, C M
Partner: UNT Libraries Government Documents Department

Methods of simulating low redox potential (Eh) for a basalt repository

Description: Basalt groundwaters have inherently low redox potentials, approximately -0.4V, which can be measured with platinum electrodes, but are difficult to reproduce during leaching experiments. In the presence of deionized water, crushed basalt reaches the measured Eh-pH values of a basalt repository. Other waste package components, such as iron, will interact with groundwater in different ways under oxic or anoxic conditions since the presence of any redox active solid will affect the groundwater Eh. 26 references, 4 figures.
Date: January 1, 1983
Creator: Jantzen, C M
Partner: UNT Libraries Government Documents Department

Prediction of glass durability as a function of environmental conditions

Description: A thermodynamic model of glass durability is applied to natural, ancient, and nuclear waste glasses. The durabilities of over 150 different natural and man-made glasses, including actual ancient Roman and Islamic glasses (Jalame ca. 350 AD, Nishapur 10-11th century AD and Gorgon 9-11th century AD), are compared. Glass durability is a function of the thermodynamic hydration free energy, ..delta..G/sub hyd/, which can be calculated from glass composition and solution pH. The durability of the most durable nuclear waste glasses examined was /approximately/10/sup 6/ years. The least durable waste glass formulations were comparable in durability to the most durable simulated medieval window glasses of /approximately/10/sup 3/ years. In this manner, the durability of nuclear waste glasses has been interpolated to be /approximately/10/sup 6/ years and no less than 10/sup 3/ years. Hydration thermodynamics have been shown to be applicable to the dissolution of glass in various natural environments. Groundwater-glass interactions relative to geologic disposal of nuclear waste, hydration rind dating of obsidians, andor other archeological studies can be modeled, e.g., the relative durabilities of six simulated medieval window glasses have been correctly predicted for both laboratory (one month) and burial (5 years) experiments. Effects of solution pH on glass dissolution has been determined experimentally for the 150 different glasses and can be predicted theoretically by hydration thermodynamics. The effects of solution redox on dissolution of glass matrix elements such as SI and B have shown to be minimal. The combined effects of solution pH and Eh have been described and unified by construction of thermodynamically calculated Pourbaix (pH-Eh) diagrams for glass dissolution. The Pourbaix diagrams have been quantified to describe glass dissolution as a function of environmental conditions by use of the data derived from hydration thermodynamics. 56 refs., 7 figs.
Date: January 1, 1988
Creator: Jantzen, C M
Partner: UNT Libraries Government Documents Department

Prediction of glass durability as a function of glass composition and test conditions: Thermodynamics and kinetics

Description: The long-term durability of nuclear waste glasses can be predicted by comparing their performance to natural and ancient glasses. Glass durability is a function of the kinetic and thermodynamic stability of glass in solution. The relationship between the kinetic and thermodynamic aspects of glass durability can be understood when the relative contributions of glass composition and imposed test conditions are delineated. Glass durability has been shown to be a function of the thermodynamic hydration free energy which can be calculated from the glass composition. Hydration thermodynamics also furnishes a quantitative frame of reference to understand how various test parameters affect glass durability. Linear relationships have been determined between the logarithmic extent of hydration and the calculated hydration free energy for several different test geometries. Different test conditions result in different kinetic reactivity parameters such as the exposed glass surface area (SA), the leachant solution volume (V), and the length of time that the glass is in the leachant (t). Leachate concentrations are known to be a function of the kinetic test parameter (SAV)t. The relative durabilities of glasses, including pure silica, obsidians, nuclear waste glasses, medieval window glasses, and frit glasses define a plane in three dimensional ..delta..G/sub hyd/-concentration-(SAV)t space. At constant kinetic conditions, e.g., test geometry and test duration, the three dimensional plane is intersected at constant (SAV)t and the ..delta..G/sub hyd/-concentration plots have similar slopes. The slope represents the natural logarithm of the theoretical slope, (12.303 RT), for the rate of glass dissolution. 53 refs., 4 figs.
Date: January 1, 1988
Creator: Jantzen, C M
Partner: UNT Libraries Government Documents Department

The role of test parameters on the kinetics and thermodynamics of glass leaching. [None]

Description: The relative durabilities of nuclear waste, natural, and ancient glasses have been assessed by standard laboratory leach tests. Different test conditions result in different glass surface areas (SA), leachant volumes (V), and test durations (t). Leachate concentrations are known to be a parabolic function of the kinetic test parameter SAV/center dot/t. Based on durability experiments of glass monoliths at low (SAV)/center dot/ glass durability has been shown to be a logarithmic function of the thermodynamic hydration free energy, ..delta..G/sub hyd/. The thermodynamic hydration free energy, ..delta..G/sub hyd/, can be calculated from glass composition and solution pH. In the repository environment high effective glass surface areas to solution volume ratios may occur as a result of slow groundwater flow rates. The application of hydration thermodynamics to crushed glass, high (SAV)/center dot/t, durability tests has been demonstrated. The relative contributions of the kinetic test parameters, (SAV)/center dot/t, and the thermodynamic parameter, ..delta..G/sub hyd/, have been shown to define a plane in ..delta..G/sub hyd/-concentration-(SAV)/center dot/t space. At constant test conditions, e.g. constant (SAV/center dot/t, the intersection with this surface indicates that all /delta G//sub hyd/-concentration plots should have similar slopes and predict the same relative durabilities for various glasses as a function of glass composition. Using this approach, the durability of nuclear waste glasses has been interpolated to be -- 10/sup 6/ years and no less than 10/sup 3/ years. 28 refs., 24 figs.
Date: January 1, 1988
Creator: Jantzen, C M
Partner: UNT Libraries Government Documents Department

Systems approach to nuclear waste glass development

Description: Development of a host solid for the immobilization of nuclear waste has focused on various vitreous wasteforms. The systems approach requires that parameters affecting product performance and processing be considered simultaneously. Application of the systems approach indicates that borosilicate glasses are, overall, the most suitable glasses for the immobilization of nuclear waste. Phosphate glasses are highly durable; but the glass melts are highly corrosive and the glasses have poor thermal stability and low solubility for many waste components. High-silica glasses have good chemical durability, thermal stability, and mechanical stability, but the associated high melting temperatures increase volatilization of hazardous species in the waste. Borosilicate glasses are chemically durable and are stable both thermally and mechanically. The borosilicate melts are generally less corrosive than commercial glasses, and the melt temperature miimizes excessive volatility of hazardous species. Optimization of borosilicate waste glass formulations has led to their acceptance as the reference nuclear wasteform in the United States, United Kingdom, Belgium, Germany, France, Sweden, Switzerland, and Japan.
Date: January 1, 1986
Creator: Jantzen, C M
Partner: UNT Libraries Government Documents Department

Investigation of lead-iron-phosphate glass for SRP waste

Description: Development of a host solid for immobilizing nuclear waste has focused on various vitreous wasteforms. Recently, lead-iron-phosphate (LIP) glasses have been proposed for solidifying all types of high-level liquid waste (HLLW). Investigation of this glass for vitrifying Savannah River Plant (SRP) waste demonstrated that the phosphate glass is incompatible with current borosilicate glass processing. Although the durability of the LIP glasses in deionized water was comparable to current borosilicate waste glass formulations, many of the defense waste constituents have low solubility in the phosphate melt, producing a nonhomogeneous or nonvitreous product. Although the LIP glass has a low melt-temperature, it is highly corrosive, which prevents the use of current melter materials such as Inconel and alumina, and requires more exotic materials of construction such as platinum.
Date: January 1, 1986
Creator: Jantzen, C M
Partner: UNT Libraries Government Documents Department

Effects of Eh (oxidation potential) on borosilicate waste glass durability

Description: Solid state materials can be used to control the oxidation potential (Eh) of waste glass leachants. These materials are chosen based on known Eh-pH relations for redox species in the presence of dissolved silica. Experiments under oxidized conditions have enabled evaluation of waste glass durability in the presence of waste package components which affect Eh. 18 references, 4 figures.
Date: January 1, 1984
Creator: Jantzen, C M
Partner: UNT Libraries Government Documents Department

Stabilization of Savannah River National Laboartory (SRNL) Aqueous Waste by Fluidized Bed Steam Reforming (FBSR)

Description: The Savannah River National Laboratory (SRNL) is a multidisciplinary laboratory operated by Westinghouse Savannah River Company (WSRC) in Aiken, South Carolina. Research and development programs have been conducted at SRNL for {approx}50 years generating non-radioactive (hazardous and non-hazardous) and radioactive aqueous wastes. Typically the aqueous effluents from the R&D activities are disposed of from each laboratory module via the High Activity Drains (HAD) or the Low Activity Drains (LAD) depending on whether they are radioactive or not. The aqueous effluents are collected in holding tanks, analyzed and shipped to either H-Area (HAD waste) or the F/H Area Effluent Treatment Facility (ETF) (LAD waste) for volume reduction. Because collection, analysis, and transport of LAD and HAD waste is cumbersome and since future treatment of this waste may be curtailed as the F/H-Area evaporators and waste tanks are decommissioned, SRNL laboratory operations requested several proof of principle demonstrations of alternate technologies that would define an alternative disposal path for the aqueous wastes. Proof of principle for the disposal of SRNL HAD waste using a technology known as Fluidized Bed Steam Reforming (FBSR) is the focus of the current study. The FBSR technology can be performed either as a batch process, e.g. in each laboratory module in small furnaces with an 8'' by 8'' footprint, or in a semi-continuous Bench Scale Reformer (BSR). The proof of principle experiments described in this study cover the use of the FBSR technology at any scale (pilot or full scale). The proof of principle experiments described in this study used a non-radioactive HAD simulant.
Date: November 1, 2004
Creator: Jantzen, C
Partner: UNT Libraries Government Documents Department

FLUIDIZED BED STEAM REFORMER MONOLITH FORMATION

Description: Fluidized Bed Steam Reforming (FBSR) is being considered as an alternative technology for the immobilization of a wide variety of aqueous high sodium containing radioactive wastes at various DOE facilities in the United States. The addition of clay, charcoal, and a catalyst as co-reactants converts aqueous Low Activity Wastes (LAW) to a granular or ''mineralized'' waste form while converting organic components to CO{sub 2} and steam, and nitrate/nitrite components, if any, to N{sub 2}. The waste form produced is a multiphase mineral assemblage of Na-Al-Si (NAS) feldspathoid minerals with cage-like structures that atomically bond radionuclides like Tc-99 and anions such as SO{sub 4}, I, F, and Cl. The granular product has been shown to be as durable as LAW glass. Shallow land burial requires that the mineralized waste form be able to sustain the weight of soil overburden and potential intrusion by future generations. The strength requirement necessitates binding the granular product into a monolith. FBSR mineral products were formulated into a variety of monoliths including various cements, Ceramicrete, and hydroceramics. All but one of the nine monoliths tested met the <2g/m{sup 2} durability specification for Na and Re (simulant for Tc-99) when tested using the Product Consistency Test (PCT; ASTM C1285). Of the nine monoliths tested the cements produced with 80-87 wt% FBSR product, the Ceramicrete, and the hydroceramic produced with 83.3 wt% FBSR product, met the compressive strength and durability requirements for an LAW waste form.
Date: December 22, 2006
Creator: Jantzen, C
Partner: UNT Libraries Government Documents Department

FLUIDIZED BED STEAM REFORMER (FBSR) PRODUCT: MONOLITH FORMATION AND CHARACTERIZATION

Description: The most important requirement for Hanford's low activity waste (LAW) form for shallow land disposal is the chemical durability of the product. A secondary, but still essential specification, is the compressive strength of the material with regards to the strength of the material under shallow land disposal conditions, e.g. the weight of soil overburden and potential intrusion by future generations, because the term ''near-surface disposal'' indicates disposal in the uppermost portion, or approximately the top 30 meters, of the earth's surface. The THOR{reg_sign} Treatment Technologies (TTT) mineral waste form for LAW is granular in nature because it is formed by Fluidized Bed Steam Reforming (FBSR). As a granular product it has been shown to be as durable as Hanford's LAW glass during testing with ASTM C-1285-02 known as the Product Consistency Test (PCT) and with the Single Pass Flow Through Test (SPFT). Hanford Envelope A and Envelope C simulants both performed well during PCT and SPFT testing and during subsequent performance assessment modeling. This is partially due to the high aluminosilicate content of the mineral product which provides a natural aluminosilicate buffering mechanism that inhibits leaching and is known to occur in naturally occurring aluminosilicate mineral analogs. In order for the TTT Na-Al-Si (NAS) granular mineral product to meet the compressive strength requirements (ASTM C39) for a Hanford waste form, the granular product needs to be made into a monolith or disposed of in High Integrity Containers (HIC's). Additionally, the Hanford intruder scenario for disposal in the Immobilized Low Activity Waste (ILAW) trench is mitigated as there is reduced intruder exposure when a waste form is in a monolithic form. During the preliminary testing of a monolith binder for TTT's FBSR mineral product, four parameters were monitored: (1) waste loading (not optimized for each waste form tested); (2) density; (3) ...
Date: September 13, 2006
Creator: Jantzen, C
Partner: UNT Libraries Government Documents Department

DEVELOPMENT OF GLASS MATRICES FOR HLW RADIOACTIVE WASTES

Description: Vitrification is currently the most widely used technology for the treatment of high level radioactive wastes (HLW) throughout the world. Most of the nations that have generated HLW are immobilizing in either borosilicate glass or phosphate glass. One of the primary reasons that glass has become the most widely used immobilization media is the relative simplicity of the vitrification process, e.g. melt waste plus glass forming frit additives and cast. A second reason that glass has become widely used for HLW is that the short range order (SRO) and medium range order (MRO) found in glass atomistically bonds the radionuclides and governs the melt properties such as viscosity, resistivity, sulphate solubility. The molecular structure of glass controls contaminant/radionuclide release by establishing the distribution of ion exchange sites, hydrolysis sites, and the access of water to those sites. The molecular structure is flexible and hence accounts for the flexibility of glass formulations to waste variability. Nuclear waste glasses melt between 1050-1150 C which minimizes the volatility of radioactive components such as Tc{sup 99}, Cs{sup 137}, and I{sup 129}. Nuclear waste glasses have good long term stability including irradiation resistance. Process control models based on the molecular structure of glass have been mechanistically derived and have been demonstrated to be accurate enough to control the world's largest HLW Joule heated ceramic melter in the US since 1996 at 95% confidence.
Date: March 18, 2010
Creator: Jantzen, C.
Partner: UNT Libraries Government Documents Department

DURABILITY TESTING OF FLUIDIZED BED STEAM REFORMER (FBSR) WASTE FORMS

Description: Fluidized Bed Steam Reforming (FBSR) is being considered as a potential technology for the immobilization of a wide variety of high sodium aqueous radioactive wastes. The addition of clay and a catalyst as co-reactants converts high sodium aqueous low activity wastes (LAW) such as those existing at the Hanford and Idaho DOE sites to a granular ''mineralized'' waste form that may be made into a monolith form if necessary. Simulant Hanford and Idaho high sodium wastes were processed in a pilot scale FBSR at Science Applications International Corporation (SAIC) Science and Technology Applications Research (STAR) facility in Idaho Falls, ID. Granular mineral waste forms were made from (1) a basic Hanford Envelope A low-activity waste (LAW) simulant and (2) an acidic INL simulant commonly referred to as sodium-bearing waste (SBW). The FBSR waste forms were characterized and the durability tested via ASTM C1285 (Product Consistency Test), the Environmental Protection Agency (EPA) Toxic Characteristic Leaching Procedure (TCLP), and the Single Pass Flow Through (SPFT) test. The durability of the FBSR waste form products was tested in order to compare the measured durability to previous FBSR waste form testing on Hanford Envelope C waste forms that were made by THOR Treatment Technologies (TTT) and to compare the FBSR durability to vitreous LAW waste forms, specifically the Hanford low activity waste (LAW) glass known as the Low-activity Reference Material (LRM). The durability of the FBSR waste form is comparable to that of the LRM glass for the test responses studied.
Date: January 6, 2006
Creator: Jantzen, C
Partner: UNT Libraries Government Documents Department

DURABLE GLASS FOR THOUSANDS OF YEARS

Description: The durability of natural glasses on geological time scales and ancient glasses for thousands of years is well documented. The necessity to predict the durability of high level nuclear waste (HLW) glasses on extended time scales has led to various thermodynamic and kinetic approaches. Advances in the measurement of medium range order (MRO) in glasses has led to the understanding that the molecular structure of a glass, and thus the glass composition, controls the glass durability by establishing the distribution of ion exchange sites, hydrolysis sites, and the access of water to those sites. During the early stages of glass dissolution, a 'gel' layer resembling a membrane forms through which ions exchange between the glass and the leachant. The hydrated gel layer exhibits acid/base properties which are manifested as the pH dependence of the thickness and nature of the gel layer. The gel layer ages into clay or zeolite minerals by Ostwald ripening. Zeolite mineral assemblages (higher pH and Al{sup 3+} rich glasses) may cause the dissolution rate to increase which is undesirable for long-term performance of glass in the environment. Thermodynamic and structural approaches to the prediction of glass durability are compared versus Ostwald ripening.
Date: December 4, 2009
Creator: Jantzen, C.
Partner: UNT Libraries Government Documents Department

MINERALIZATION OF RADIOACTIVE WASTES BY FLUIDIZED BED STEAM REFORMING (FBSR): COMPARISONS TO VITREOUS WASTE FORMS, AND PERTINENT DURABILITY TESTING

Description: The Savannah River National Laboratory (SRNL) was requested to generate a document for the Washington State Department of Ecology and the U.S. Environmental Protection Agency that would cover the following topics: (1) A description of the mineral structures produced by Fluidized Bed Steam Reforming (FBSR) of Hanford type Low Activity Waste (LAW including LAWR which is LAW melter recycle waste) waste, especially the cage structured minerals and how they are formed. (2) How the cage structured minerals contain some contaminants, while others become part of the mineral structure (Note that all contaminants become part of the mineral structure and this will be described in the subsequent sections of this report). (3) Possible contaminant release mechanisms from the mineral structures. (4) Appropriate analyses to evaluate these release mechanisms. (5) Why the appropriate analyses are comparable to the existing Hanford glass dataset. In order to discuss the mineral structures and how they bond contaminants a brief description of the structures of both mineral (ceramic) and vitreous waste forms will be given to show their similarities. By demonstrating the similarities of mineral and vitreous waste forms on atomic level, the contaminant release mechanisms of the crystalline (mineral) and amorphous (glass) waste forms can be compared. This will then logically lead to the discussion of why many of the analyses used to evaluate vitreous waste forms and glass-ceramics (also known as glass composite materials) are appropriate for determining the release mechanisms of LAW/LAWR mineral waste forms and how the durability data on LAW/LAWR mineral waste forms relate to the durability data for LAW/LAWR glasses. The text will discuss the LAW mineral waste form made by FBSR. The nanoscale mechanism by which the minerals form will be also be described in the text. The appropriate analyses to evaluate contaminant release mechanisms will be discussed, as ...
Date: December 26, 2008
Creator: Jantzen, C
Partner: UNT Libraries Government Documents Department

Disposition of Tank 48H Organics by Fluidized Bed Steam Reforming (FBSR)

Description: In order to make space in the Savannah River Site Tank farm, the Tank 48H waste must be removed. Therefore, the Tank 48H waste must be processed to reduce or eliminate levels of nitrates, nitrites, and sodium tetraphenyl borate in order to reduce impacts of these species before it is vitrified. Fluidized Bed Steam Reforming is being considered as a candidate technology for destroying the nitrates and the NaTPB prior to melting. The Idaho National Engineering and Environmental Laboratory was tasked to perform a proof-of-concept steam reforming test to evaluate the technical feasibility for pretreating the Tank 48H waste. The crucible (bench scale) tests conducted at the Savannah River Technology Center were initiated to optimize and augment the parameters subsequently tested at the pilot scale at INEEL. The purposes of the current study, organic destruction and downstream processing of T48H waste slurry were fulfilled. TPB was destroyed in all 19 samples tested with the simulated FB SR process at operational temperatures 650-725 degrees Celsius. A test temperature of 650 degrees Celsius optimized NO3 destruction during the formation of an Na2CO3 FBSR product. A test temperature of 725 degrees Celsius optimized NO3 destruction during formation of a sodium silicate FBSR product. Destruction of nitrate at greater than 99 per cent was achieved with addition of sugar as a reductant at 1X stoichiometry and total organic carbon analyses indicated that excess reductant was not present in the FBSR product. The use of sugar at 1X stoichiometry appears to ensure that excess reductant is not contained in the FBSR product that would alter the REDuction/OXidation equilibrium of the DWPF melter, while simultaneously assuring that NO3 is destroyed adequately. Destruction of antifoam with the simulated FBSR process was also achieved at operating temperatures between 650-725 degrees Celsius. based on measured total organic carbon.
Date: December 2, 2003
Creator: Jantzen, C.M.
Partner: UNT Libraries Government Documents Department

Impact of Zeolite Transferred from Tank 19F to Tank 18F on DWPF Vitrification of Sludge Batch 3

Description: The Defense Waste Processing Facility (DWPF) is planning to initiate vitrification of Sludge Batch 3 (SB3) in combination with Sludge Batch 2 (SB2) in the spring of 2004. The contents of Sludge Batch 3 will be a mixture of the heel remaining from Sludge Batch 1B, sludge from Tank 7F (containing coal, sand, and sodium oxalate), and sludge materials from Tank 18F. The sludge materials in Tank 18F contain part of a mound of zeolitic material transferred there from Tank 19F. This mound was physically broken up and transfers were made from Tank 19F to Tank 18F for vitrification into SB3. In addition, excess Pu and Am/Cm materials were transferred to Tank 51H to be processed through the DWPF as part of SB3. Additional Pu material and a Np stream from the Canyons are also planned to be added to SB3 before processing of this batch commences at DWPF. The primary objective of this task was to assess the impacts of the excess zeolite mound material in Tank 19F on the predicted glass and processing properties of interest when the zeolite becomes part of SB3. The two potential impacts of the Tank 19F zeolite mound on DWPF processing relates to (1) the samples taken for determination of the acceptability of a macrobatch of DWPF feed and (2) the achievable waste loading. The potential effects of the large size of the zeolite particles found in the Tank 19F solids, as reported in this study, are considered minimal for processing of SB3 in DWPF. Other findings about the zeolite conversion mechanism via a process of Ostwald ripening are discussed in the text and in the conclusions.
Date: January 7, 2004
Creator: Jantzen, C.M.
Partner: UNT Libraries Government Documents Department

Thermodynamic Modeling of the SRS Evaporators: Part II. The 3H System

Description: Accumulations of two solid phases have formed scale deposits in the Savannah River Site 2H Evaporator system since late 1996. The aluminosilicate scale deposits caused the evaporator pot to become inoperable in October 1999. Accumulations of the diuranate phase have caused criticality concerns in the SRS 2H Evaporator. In order to ensure that similar deposits are not forming in the SRS 3H Evaporator, thermodynamically derived activity diagrams specific to the feeds processed from the 3H Evaporator feed tank (Tank 32) were evaluated.
Date: May 31, 2002
Creator: Jantzen, C.M.
Partner: UNT Libraries Government Documents Department

Impact of Zeolite Transferred from Tank 19F to Tank 18F on DWPF Vitrification of Sludge Batch 3

Description: The Defense Waste Processing Facility (DWPF) is planning to initiate vitrification of Sludge Batch 3 (SB3) in combination with Sludge Batch 2 (SB2) in the spring of 2004. The contents of Sludge Batch 3 will be a mixture of the heel remaining from Sludge Batch 1B, sludge from Tank 7F (containing coal, sand, and sodium oxalate), and sludge materials from Tank 18F. The sludge materials in Tank 18F contain part of a mound of zeolitic material transferred there from Tank 19F. This mound was physically broken up and transfers were made from Tank 19F to Tank 18F for vitrification into SB3. In addition, excess Pu and Am/Cm materials were transferred to Tank 51H to be processed through the DWPF as part of SB3. Additional Pu material and a Np stream from the Canyons are also planned to be added to SB3 before processing of this batch commences at DWPF. The primary objective of this task was to assess the impacts of the excess zeolite mound material in Tank 19F on the predicted glass and processing properties of interest when the zeolite becomes part of SB3. The two potential impacts of the Tank 19F zeolite mound on DWPF processing relates to (1) the samples taken for determination of the acceptability of a macrobatch of DWPF feed and (2) the achievable waste loading. The potential effects of the large size of the zeolite particles found in the Tank 19F solids, as reported in this study, are considered minimal for processing of SB3 in DWPF. Other findings about the zeolite conversion mechanism via a process of Ostwald ripening are discussed in the text and in the conclusions.
Date: January 7, 2004
Creator: Jantzen, C. M.
Partner: UNT Libraries Government Documents Department

Electron Equivalents REDOX Model for High Level Waste Vitrification

Description: Control of the REDuction/OXidation (REDOX) equilibrium in high level waste (HLW) glass melters is critical in order to eliminate the formation of metallic species from overly reduced melts while minimizing foaming from overly oxidized melts. To date, formates, nitrates, and manganic species in the melter feeds going to the Savannah River Site (SRS) Defense Waste Processing Facility (DWPF) have been the major parameters influencing melt REDOX. The sludge being processed for inclusion in the next DWPF Sludge Batch contains several organic components that are considered non-typical of DWPF sludge to date, e.g. oxalates and coal. A mechanistic REDOX model was developed to balance any reductants and any oxidants for any HLW melter feed. The model is represented by the number of electrons gained during reduction of an oxidant or lost during oxidation of a reductant. The overall relationship between the REDOX ratio of the final glass and the melter feed is given in terms of the transfer of molar Electron Equivalents.
Date: September 11, 2003
Creator: Jantzen, C.M.
Partner: UNT Libraries Government Documents Department

Vitrification of Simulated Fernald K-65 Silo Waste at Low Temperature

Description: Vitrification is the technology that has been chosen to solidify approximately 18,000 tons of geologic mill tailings at the Fernald Environmental Management Project (FEMP) in Fernald, Ohio. The geologic mill tailings are residues from the processing of pitchlende ore during 1949-1958. These waste residues are contained in silos in Operable Unit 4 (OU4) at the FEMP facility. Operable Unit 4 is one of five operable units at the FEMP. Operable Unit 4 is one of five operable units at the FEMP. Operating Unit 4 consists of four concrete storage silos and their contents. Silos 1 and 2 contain K-65 mill tailing residues and a bentonite cap, Silo 3 contains non-radioactive metal oxides, and Silo 4 is empty.
Date: March 15, 1999
Creator: Jantzen, C.M.
Partner: UNT Libraries Government Documents Department

Savannah River Site waste vitrification projects initiated throughout the United States: Disposal and recycle options

Description: A vitrification process was developed and successfully implemented by the US Department of Energy's (DOE) Savannah River Site (SRS) and at the West Valley Nuclear Services (WVNS) to convert high-level liquid nuclear wastes (HLLW) to a solid borosilicate glass for safe long term geologic disposal. Over the last decade, SRS has successfully completed two additional vitrification projects to safely dispose of mixed low level wastes (MLLW) (radioactive and hazardous) at the SRS and at the Oak Ridge Reservation (ORR). The SRS, in conjunction with other laboratories, has also demonstrated that vitrification can be used to dispose of a wide variety of MLLW and low-level wastes (LLW) at the SRS, at ORR, at the Los Alamos National Laboratory (LANL), at Rocky Flats (RF), at the Fernald Environmental Management Project (FEMP), and at the Hanford Waste Vitrification Project (HWVP). The SRS, in conjunction with the Electric Power Research Institute and the National Atomic Energy Commission of Argentina (CNEA), have demonstrated that vitrification can also be used to safely dispose of ion-exchange (IEX) resins and sludges from commercial nuclear reactors. In addition, the SRS has successfully demonstrated that numerous wastes declared hazardous by the US Environmental Protection Agency (EPA) can be vitrified, e.g. mining industry wastes, contaminated harbor sludges, asbestos containing material (ACM), Pb-paint on army tanks and bridges. Once these EPA hazardous wastes are vitrified, the waste glass is rendered non-hazardous allowing these materials to be recycled as glassphalt (glass impregnated asphalt for roads and runways), roofing shingles, glasscrete (glass used as aggregate in concrete), or other uses. Glass is also being used as a medium to transport SRS americium (Am) and curium (Cm) to the Oak Ridge Reservation (ORR) for recycle in the ORR medical source program and use in smoke detectors at an estimated value of $1.5 billion to the ...
Date: April 10, 2000
Creator: Jantzen, C.M.
Partner: UNT Libraries Government Documents Department

Thermodynamic Modeling of the SRS Evaporators: Part II. The 3H System

Description: Accumulations of two solid phases have formed scale deposits in the Savannah River Site 2H Evaporator system since late 1996. The aluminosilicate scale deposits caused the evaporator pot to become inoperable in October 1999. Accumulations of the diuranate phase have caused criticality concerns in the SRS 2H Evaporator. In order to ensure that similar deposits are not and will not form in the SRS 3H Evaporator, thermodynamically derived activity diagrams specific to the feeds processed from Tanks 30 and 32 are evaluated in this report.
Date: October 2, 2001
Creator: Jantzen, C.M.
Partner: UNT Libraries Government Documents Department