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At least 469 records · Page 26
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Evaluation of the Waste Streams from Management of Reprocessing Off-Gas Streams - 18483

Nuclear fuel reprocessing is being considered in the United States as part of an effort to help meet US energy needs and as a means to improve waste management. During the chemical processing of UNF to recover the valuable components for reuse, a number of volatile radionuclides will be released into the off-gas streams of the reprocessing plant. These volatile radionuclides include H-3, C-14, Kr-85, and I-129. Off-gas abatement systems for the volatile radionuclides will be required to ensure that air emissions from a UNF reprocessing facility can meet or exceed emission requirements set forth by the US Environmental Protection Agency and the US Nuclear Regulatory Commission. An initial engineering evaluation of offgas abatement systems was completed recently. The evaluation addressed the designs of the head-end offgas treatment system, which typically includes both the shear off-gas and the dissolver off-gas and vessel off-gas treatment system. Based on this study, the waste volumes for each of the four primary volatile radionuclides (H-3, C-14, Kr-85, and I-129) were estimated. Included in these estimates were the nonradioactive isotopes and the 'tramp' elements found in air (e.g., Kr and C-12 as CO{sub 2}) and process chemicals (e.g., I and Cl). Because the final waste form for these radionuclides has not been selected, estimated volumes were calculated for various alternative waste forms as appropriate. The resulting waste volume for I was found to be highly dependent on the final waste form selected and increased by a factor of 2 to 4 by tramp halogens that would be released into the off-gas stream from the process chemicals. The volume of tritiated water from the fuel was quite small, but the estimated volume of the final waste form depended on the tritium management approach employed within the reprocessing plant. The Kr waste volume was calculated assuming pure Kr (i.e., assuming complete separation of Xe and Ar). For the case of no Xe separation, the estimated waste volume increased by factor of ∼10. Argon from the air, if not removed, increased the estimated volume by a factor of 425. The estimated volume of C-14 waste was dominated by the CO{sub 2} from the air leaking into the processing cell. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2011

REPORTABLE RADIONUCLIDES IN DWPF SLUDGE BATCH 7A (MACROBATCH 8)

The Waste Acceptance Product Specifications (WAPS) 1.2 require that the waste producer 'shall report the curie inventory of radionuclides that have half-lives longer than 10 years and that are, or will be, present in concentrations greater than 0.05 percent of the total inventory for each waste type indexed to the years 2015 and 3115.' As part of the strategy to meet WAPS 1.2, the Defense Waste Processing Facility (DWPF) will report for each waste type all radionuclides that have half-lives longer than 10 years and contribute greater than 0.01 percent of the total curie inventory from the time of production through the 1100 year period from 2015 through 3115. The initial list of radionuclides to be reported is based on the design-basis glass identified in the Waste Form Compliance Plan (WCP) and Waste Form Qualification Report. However, it is required that the list be expanded if other radionuclides with half-lives greater than 10 years are identified that meet the 'greater than 0.01% of the curie inventory' criterion. Specification 1.6 of the WAPS, International Atomic Energy Agency Safeguards Reporting for High Level Waste (HLW), requires that the ratio by weights of the following uranium and plutonium isotopes be reported: U-233, U-234, U-235, U-236, and U-238; and Pu-238, Pu-239, Pu-240, Pu-241, and Pu-242. Therefore, the complete list of reportable radionuclides must also include these sets of U and Pu isotopes - and the U and Pu isotopic mass distributions must be identified. The DWPF receives HLW sludge slurry from Savannah River Site (SRS) Tank 40. For Sludge Batch 7a (SB7a), the waste in Tank 40 contained a blend of the heel from Sludge Batch 6 (SB6) and the Sludge Batch 7 (SB7) material transferred to Tank 40 from Tank 51. This sludge blend is also referred to as Macrobatch 8. Laboratory analyses of a Tank 40 sludge sample were performed to quantify the concentrations of pertinent radionuclides in the SB7a waste. Subsequently, radiological decay and in-growth were calculated over the time period from 2015 to 3115. This provided a basis for characterizing the radionuclide content of SB7a over time and for identifying the 'reportable radionuclides.' Details of the characterization methodology and the analytical results are the focus of this report. This work was performed at the request of the Waste Solidification Engineering Department of Savannah River Remediation, initiated via Technical Task Request (TTR) HLW-DWPF-TTR-2010-0031. A minor revision in the reporting requirements was requested via a subsequent email communication. The work was conducted in accordance with the protocols identified in Task Technical and Quality Assurance Plan SRNL-RP-2010-01218 and Analytical Study Plan SRNL-RP-2010-01219. All of the raw data related to this scope have been recorded in laboratory notebook SRNL-NB-2011-00061. The overall goal of this task was to characterize the radionuclide content of the SB7a waste sufficiently to meet the WAPS and DWPF reporting requirements. The specific objectives were: (1) Quantify the current concentrations of all radionuclides impacting (or potentially-impacting) the total curie content between calendar years 2011 and 3115. Also quantify the current concentrations of other radionuclides specifically requested in the TTR or required by the WAPS. (2) Calculate future concentrations of decayed and in-grown radionuclides impacting the total curie content between calendar years 2015 and 3115; (3) Identify as 'reportable' all radionuclides contributing {ge} 0.01% of the total curie content from 2015 to 3115 and having half-lives {ge} 10 years.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

Glass Ceramic Formulation Data Package

A glass ceramic waste form is being developed for treatment of secondary waste streams generated by aqueous reprocessing of commercial used nuclear fuel (Crum et al. 2012b). The waste stream contains a mixture of transition metals, alkali, alkaline earths, and lanthanides, several of which exceed the solubility limits of a single phase borosilicate glass (Crum et al. 2009; Caurant et al. 2007). A multi-phase glass ceramic waste form allows incorporation of insoluble components of the waste by designed crystallization into durable heat tolerant phases. The glass ceramic formulation and processing targets the formation of the following three stable crystalline phases: (1) powellite (XMoO 4 ) where X can be (Ca, Sr, Ba, and/or Ln), (2) oxyapatite Y x, Z (10-x) Si 6 O 26 where Y is alkaline earth, Z is Ln, and (3) lanthanide borosilicate (Ln 5 BSi 2 O 13 ). These three phases incorporate the waste components that are above the solubility limit of a single-phase borosilicate glass. The glass ceramic is designed to be a single phase melt, just like a borosilicate glass, and then crystallize upon slow cooling to form the targeted phases. The slow cooling schedule is based on the centerline cooling profile of a 2 foot diameter canister such as the Hanford High-Level Waste canister. Up to this point, crucible testing has been used for glass ceramic development, with cold crucible induction melter (CCIM) targeted as the ultimate processing technology for the waste form. Idaho National Laboratory (INL) will conduct a scaled CCIM test in FY2012 with a glass ceramic to demonstrate the processing behavior. This Data Package documents the laboratory studies of the glass ceramic composition to support the CCIM test. Pacific Northwest National Laboratory (PNNL) measured melt viscosity, electrical conductivity, and crystallization behavior upon cooling to identify a processing window (temperature range) for melter operation and cooling profiles necessary to crystallize the targeted phases in the waste form.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Organic Evaporation, Oxidation, and Hydrolysis Testing in Support of Hanford Sample-and-Send

The Hanford site has approximately 56 million gallons of radioactive mixed waste stored in 156 unretrieved underground storage tanks. The Hanford Waste Treatment and Immobilization Plant (WTP) is being built to treat and immobilize the tank waste. The baseline method for immobilization of Low Activity Waste (LAW) through the WTP is vitrification, but additional immobilization capacity is needed to supplement the initial LAW melters. An alternative cementitious waste form is being investigated for that future immobilization method. However, one impediment to a cementitious waste form is the presence of Land Disposal Restricted (LDR) organic chemicals in tank waste, which are regulated on a concentration-based standard in the final waste form. Hence, if the quantity of organics in LAW is high enough, they must be destroyed or removed to make a waste form compatible with disposal in a mixed low level waste landfill. This work evaluates potential avenues for treatment of LDR organics to eliminate the impediment and permit possible use of a cementitious waste form.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Cementation of Nuclear Wastes: From Past to Future in a Sustainable Approach - 18266

Cementation of nuclear wastes is a widespread technology due to several factors such as excellent feedback from past experience, simplicity of implementation and versatility. Indeed, as nuclear wastes come in many physical states (bulk wastes, powders, sludges), cementation is the immobilisation process that can easily accommodate this large range of wastes provided the appropriate mixing technology is used: heterogenous cementation (grouting of bulky wastes) or homogenous cementation (intermixing process between the powder or the sludge and the cementitious material). However, nuclear wastes equally present an important chemical variability. This is especially true as New Areva's range of nuclear activities covers the entire fuel cycle: mining, conversion, enrichment and recycling and dismantling and decommissioning. The chemical interactions between the waste and the cement matrix are carefully studied by New Areva's R and D teams in order to provide comprehensive solutions by developing cement formulations which are adapted to wastes characteristics. The aim of the studies is to obtain cement waste-forms that meet the waste acceptance criteria, the final product being thus in perfect adequacy with the disposal requirements. For over 30 years New Areva has been developing cement formulations adapted to the physical and chemical specificities of the wastes accounting for constrains imposed by waste characteristics. The overall approach includes the definition of the appropriate cement composition at laboratory scale, engineering design, equipment manufacturing, testing and commissioning of the cementation facilities. With this strong technical background, a specific methodology has been developed by New Areva teams in order to allow a rapid screening of possible solutions for a given waste (in terms of cement chemistry adapted to the waste) in a learn-fast approach that is based on theoretical analysis and small-scale tests. The methodology was applied for wastes presenting a complex nature (such as reactive metals, ion exchange resins, oils and solvents, mercury, etc), known to interact with Portland cement from a chemical point of view and leading to durability issues of the wasteform (swelling, strength reduction, leaching) and thus reducing the final waste loading. Once the screening process is over, the main drivers for choosing the most appropriate solution to be further developed at pilot scale are: reducing the overall volume and the impact of the final waste package, ensure the waste package performance as defined by the disposal requirements and control the long-term behaviour in final disposal. Moreover, a modelling approach is addressed, aiming to develop a thermo-chemo-mechanical (THM) concrete models that are able to predict the behaviour of the structure (waste form) in time. The aim of the present paper is to describe the main principles within the developed methodology and illustrate it by using a few specific examples. Several distinct cement formulations are developed for complex wastes, limiting the chemical interaction between the cement matrix and the given wastes, obtaining better waste loadings and better overall durability behaviour which facilitates long term behaviour demonstration. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Organic Evaporation, Oxidation, and Hydrolysis Testing in Support of Hanford Sample-and-Send

The Hanford site has approximately 54 to 56 million gallons of radioactive mixed waste stored in 156 unretrieved underground storage tanks. The Hanford Waste Treatment and Immobilization Plant (WTP) is being built to treat and immobilize the tank waste. The baseline method for immobilization of Low Activity Waste (LAW) through the WTP is vitrification, but additional immobilization capacity is needed to supplement the initial LAW melters. An alternative cementitious waste form is being investigated for that future immobilization method. However, one impediment to a cementitious waste form is the presence of Land Disposal Restricted (LDR) organic chemicals in tank waste, which are regulated on a concentration based standard in the final waste form. Hence, if the quantity of organics in LAW is high enough, they must be destroyed or removed to make a waste form compatible with disposal in a mixed low level waste landfill. This work evaluates potential avenues for treatment of LDR organics to eliminate the impediment and permit possible use of a cementitious waste form. Vacuum evaporation testing to remove LDR organics consisted of preparing a non-radioactive LAW simulant, spiking that simulant with organic chemicals, and evaporating the mixture via differential distillation. The apparatus was a laboratory-scale vacuum evaporator operated at 60 ±5 torr absolute (vacuum evaporation). The LAW simulant represented the liquid expected to be retrieved from the Hanford tank farms at approximately 4.0 M [Na+] total sodium ion concentration. The concentration of the organic chemicals added was significantly higher than typically found in the tank waste samples since the higher levels were necessary to assist in analytical measurement and tracking of the spiked species.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

FY25 Organic Evaporation Testing in Support of Hanford Sample-and-Send

The Hanford Waste Treatment and Immobilization Plant (WTP) is being built to treat and immobilize the approximately 56 million gallons of radioactive mixed waste stored in 156 underground storage tanks. The baseline method for immobilization of Low-Activity Waste (LAW) is vitrification, but additional immobilization capacity is needed to supplement the initial LAW melters. Additionally, efforts are being undertaken to accelerate the disposition of Pre-Treated Waste (PTW) in Hanford’s West Area. An alternative cementitious waste form is being investigated as an alternative immobilization method. However, one impediment to a cementitious waste form is the presence of Land Disposal Restricted (LDR) organic chemicals in tank waste, which are regulated on a concentration-based standard in the final waste form. Hence, if the quantity of organics in PTW and LAW is high enough, treatment may be needed to destroy or remove said organics to make a waste form compatible with disposal in a mixed low level waste landfill. This work evaluates evaporation as a potential avenue for treatment of LDR organics to eliminate the impediment and permit possible use of a cementitious waste form.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Salt-stone Oxidation Study: Leaching Method - 13092

Cementitious waste forms can be designed to chemically stabilize selected contaminants, such as Tc{sup +7} and Cr{sup +6}, by chemically reduction to lower valance states, Tc{sup +4} and Cr{sup +3}, respectively, and precipitation of these species in alkaline media as low solubility solid phases. Data for oxidation of this type of cementitious waste form cured under field conditions as a function of time is required for predicting the performance of the waste form and disposal facility. The rate of oxidation (oxidation front advancement) is an important parameter for predicting performance because the solubilities of some radionuclide contaminants, e.g., technetium, are a function of the oxidation state. A non-radioactive experiment was designed for quantifying the oxidation front advancement using chromium, as an approximate redox-sensitive surrogate (Cr{sup +6} / Cr{sup +3}) for technetium (Tc{sup +7} / Tc{sup +4}). Nonradioactive cementitious waste forms were prepared in the laboratory and cured under both laboratory and 'field conditions'. Laboratory conditions were ambient temperature and sealed sample containers. Field conditions were approximated by curing samples in open containers which were placed inside a plastic container stored outdoors at SRS. The container had a lid and was instrumented with temperature and humidity probes. Sub-samples as thin as 0.2 mm were taken as a function of distance from the exposed surface of the as-cast sample. The sub-samples were leached and the leachates were analyzed for chromium, nitrate, nitrite and sodium. Nitrate, nitrite, and sodium concentrations were used to provide baseline data because these species are not chemically retained in the waste form matrix to any significant extent and are not redox sensitive. 'Effective' oxidation fronts for Cr were measured for samples containing 1000, 500 and 20 mg/kg Cr added as soluble sodium chromate, Na{sub 2}CrO{sub 4}. For a sample cured for 129 days under field conditions, leachable Cr (assumed to be the oxidized form, i.e., Cr{sup +6} was detected in the top 15 to 20 mm of the sample spiked with 1000 mg/kg Cr. Below about 20 mm, the Cr concentrations in leachates were below the detection limit (< 0.010 mg/L) which indicates that the oxidation as the result of exposure to air was limited to the top 20 mm of the sample after exposure for 129 days and that the bulk of the waste form was not affected, i.e., the Cr was stabilized and insoluble. For samples cured in the laboratory, leachable Cr was detected in the top 8 mm of the Cr 1000 sample cured in the laboratory for 37 days. Between 8 and 14 mm, the concentration Cr in the leachate dropped by a factor of about 20 to just above the detection limit. These depth of penetration results indicate that the rate of advancement of the oxidation front for samples spiked with 1000 mg/kg Cr cured under 'field conditions' for 129 days is less than that for the sample cured in the laboratory for 37 days, i.e., 0.156 and 0.216 mm/day, respectively. Additional data are presented for samples spiked with 500 and 20 mg/kg Cr. In summary, cementitious waste forms are porous solids with a network of interconnected pores ranging in diameter from 10 E-10 m to greater than a few mm. The oxidation process is assumed to occur as the result of oxygen transport through the interconnected porosity which may be filled with air, aqueous salt solution, or both. Upon oxidation, the Cr becomes soluble and can be transported in solution through aqueous pore fluid or leachate. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Modeling Interfaces to Support Low-Level Waste Disposal System Performance Assessments - 20366

In low-level waste (LLW) disposal sites, interfaces between cementitious materials, used as waste forms and/or engineered barriers, and the surrounding soil or backfill material are often encountered. Reactions across these interfaces may lead to chemical and structural alteration of the cementitious and backfill materials that may ultimately affect long-term performance. For example, the ingress of carbon dioxide from soil gas into a waste tank concrete shell at the Hanford Site or Savannah River Site (SRS) or a low-level waste disposal vault at SRS leads to carbonation of cement hydration products (e.g. portlandite, calcium-silicate hydrate or C-S-H, and ettringite). The result of carbonation is a decrease in pH in the cement paste portion of the tank wall or vault concrete structure that could ultimately lead to de-passivation of the embedded structural steel and, eventually, to cracking. Cracking can lead to increased ingress of water into the structure, corresponding increased release of constituents of concern, and to subsequent increased transport of these constituents to the surrounding environment. The mobility of trace constituents (e.g., radionuclides of concern) in waste forms may increase in response to changes in pH, pore structure, and mineralogical gradients. Depletion of portlandite results in subsequent decalcification of C-S-H that leads to changes in the cement strength and may lead to structural failure. At the boundary between waste forms and concrete vaults, the migration of sulfate ions from the salt waste form (e.g., the saltstone waste form used for LLW disposal at the Savannah River Site in Saltstone Disposal Units, or SDUs) into the barrier or vault concrete has been predicted to result in expansive mineral phase formation (or 'sulfate attack'). Expansive mineral formation could result in cracking and potential loss of structural integrity in the vault concrete. In the performance assessment (PA) for the SRS SDUs, prediction of both the carbonation and sulfate attack fronts are critical to assessing the long-term performance of these LLW disposal vaults. A general purpose geochemical reactive transport model has been developed in LeachXS/ORCHESTRA to evaluate typical interfaces for LLW disposal environments, including SDUs at SRS. The model may be used to predict the long-term performance of interfaces between cementitious materials and saltstone or backfill materials with respect to primary phases by considering mass transport, cement chemistry, geochemical speciation, and multi-ionic diffusion over interfaces between different materials. The present study compares the performance and interactions between vault concrete and saltstone waste form for a representative SDU scenario. The materials were characterized using pH-dependence leaching test (EPA 1313) and semi-dynamic transport tests (EPA 1315) with test results used to parameterize the interface model. The development of a credible inter model provides an important basis for supporting future PAs where carbonation and sulfate attack are important aging and degradation mechanisms. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Solid Waste Release Calculations for the Composite Analysis Baseline Assessment

The purpose of the Hanford Site Composite Analysis (CA) is to provide an estimate of the cumulative radiological impacts from active and planned low-level radioactive waste disposals and other potentially interacting radioactive waste disposal sources that will remain following Hanford Site closure. The need for a waste form release model addressing multiple solid waste forms is documented in CP-60410, Hanford Site Composite Analysis Technical Approach Description: Waste Form Release.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Organic Acid-Assisted Thermal Dehalogenation of Halide Salt Nuclear Wastes: From Waste Salts to Borosilicate Glass

Only a handful of high-halide salt waste forms have been demonstrated for vitrification-based immobilization strategies for halide-salt nuclear waste streams (e.g., pyroprocessing wastes, molten salt reactor wastes) and they all have low waste loading potential and most have low chemical durabilities for high-alkali streams. An alternative approach to direct salt immobilization is salt partitioning prior to waste form fabrication and one option for partitioning is halide removal (called dehalogenation). Removing the halogen fraction through dehalogenation can significantly reduce the waste volume required for disposal in the primary waste form. Furthermore, when dehalogenation is performed using organic acids, the dehalogenation reagent can decompose during high-temperature vitrification, reducing waste loading limitations in the waste form. In the current work, different organic acids (i.e., oxalic, formic, acetic, oxamic, and citric) were evaluated for dehalogenation efficiency of a simple chloride salt simulant (7.19% LaCl 3 , 53.77% LiCl, and 39.04% KCl, by mole) and a more complex chloride salt simulant called ERV3 (electrorefiner version 3) at 150 °C–300 °C and using H + /Cl – molar ratios of 1:1, 2:1, and 3:1. Additionally, a borosilicate glass waste form called TARS (or the average of refined specifications) was formulated, produced, and characterized for dehalogenated ERV3.

Amorphous materials↗
DOE OSTI2024

Nitrate and Nitrite at Hanford – From Tanks to Natural Attenuation

Washington River Protection Solutions, LLC (WRPS) contracted Pacific Northwest National Laboratory (PNNL) in support of their maturation testing and analysis to develop and deploy an ambient temperature solidification/stabilization process for low-activity waste (LAW) (referred to as “grout” in this report). The intent of this effort is to support a sample-and-send regulatory and processing strategy in the Hanford 200 West Area, specifically considering grout waste forms disposal at Hanford in the Integrated Disposal Facility (IDF). Currently the disposal of grouted LAW is not an option because the IDF is not permitted to receive this waste form. The material presented in this report is intended to help inform the U.S. Department of Energy (DOE), the site operating contractors, regulatory agencies, and stakeholders of the future implications of IDF disposal of a grouted LAW waste form, specifically for the behavior of nitrate (NO 3 - ) and nitrite (NO 2 - ) at the Hanford. The only calculation of nitrate release from a hypothetical grouted LAW inventory in the IDF showed that the overall release was slightly above the compliance limit of 45 mg/L nitrate and 3.32 mg/L nitrate after 1000 years (Asmussen et al. 2019). As such, uncertainty around the impact of nitrate and nitrite release from the IDF was identified in the recent National Academies of Sciences study of Hanford supplemental LAW by both the national laboratory team and Hanford stakeholders (Bates et al. 2023). However, these calculations (and others for the IDF involving secondary waste grout) may be conservative as no nitrate and nitrite retention in the grout waste form was assumed, nor any attenuation in the subsurface. There is evidence from the literature of processes that can attenuate nitrate migration both in the waste form and in the Hanford subsurface. If found to be technically defensible and likely to occur in the IDF and associated subsurface, then these processes should be included in modeling of the IDF to represent the behavior of nitrate and nitrite more accurately. Doing so would remove unnecessary conservatism in the modeling projections and reduce the uncertainty in assessments of facility compliance of a grouted LAW waste form in the IDF.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Saltstone Oxidation Study: Leaching Method

Cementitious waste forms can be designed to chemically stabilize selected contaminants, such as Tc{sup +7} and Cr{sup +6}, by chemically reduction to lower valance states, Tc{sup +4} and Cr{sup +3}, respectively, and precipitation of these species in alkaline media as low solubility solid phases. Data for oxidation of this type of cementitious waste form cured under field conditions as a function of time is required for predicting the performance of the waste form and disposal facility. The rate of oxidation (oxidation front advancement) is an important parameter for predicting performance because the solubilities of some radionuclide contaminants, e.g., technetium, are a function of the oxidation state. A non-radioactive experiment was designed for quantifying the oxidation front advancement using chromium, as an approximate redox-sensitive surrogate (Cr{sup +6} / Cr{sup +3}) for technetium (Tc{sup +7} / Tc{sup +4}). Nonradioactive cementitious waste forms were prepared in the laboratory and cured under both laboratory and ?field conditions.? Laboratory conditions were ambient temperature and sealed sample containers. Field conditions were approximated by curing samples in open containers which were placed inside a plastic container stored outdoors at SRS. The container had a lid and was instrumented with temperature and humidity probes. Subsamples as thin as 0.2 mm were taken as a function of distance from the exposed surface of the as-cast sample. The subsamples were leached and the leachates were analyzed for chromium, nitrate, nitrite and sodium. Nitrate, nitrite, and sodium concentrations were used to provide baseline data because these species are not chemically retained in the waste form matrix to any significant extent and are not redox sensitive. ?Effective? oxidation fronts for Cr were measured for samples containing 1000, 500 and 20 mg/kg Cr added as soluble sodium chromate, Na{sub 2}CrO{sub 4}. For a sample cured for 129 days under field conditions, leachable Cr (assumed to be the oxidized form, i.e., Cr{sup +6} was detected in the top 15 to 20 mm of the sample spiked with 1000 mg/kg Cr. Below about 20 mm, the Cr concentrations in leachates were below the detection limit (< 0.010 mg/L) which indicates that the oxidation as the result of exposure to air was limited to the top 20 mm of the sample after exposure for 129 days and that the bulk of the waste form was not affected, i.e., the Cr was stabilized and insoluble. For samples cured in the laboratory, leachable Cr was detected in the top 8 mm of the Cr 1000 sample cured in the laboratory for 37 days. Between 8 and 14 mm, the concentration Cr in the leachate dropped by a factor of about 20 to just above the detection limit. These depth of penetration results indicate that the rate of advancement of the oxidation front for samples spiked with 1000 mg/kg Cr cured under ?field conditions? for 129 days is less than that for the sample cured in the laboratory for 37 days, i.e., 0.156 and 0.216 mm/day, respectively. Additional data are presented for samples spiked with 500 and 20 mg/kg Cr. In summary, cementitious waste forms are porous solids with a network of interconnected pores ranging in diameter from 10E-10 m to greater than a few mm. The oxidation process is assumed to occur as the result of oxygen transport through the interconnected porosity which may be filled with air, aqueous salt solution, or both. Upon oxidation, the Cr becomes soluble and can be transported in solution through aqueous pore fluid or leachate.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Performance Assessment of a Generic Nuclear Waste Repository in Shale - 18392

A performance assessment (PA) is an important component of a comprehensive safety analysis for a nuclear waste repository. In a PA, probabilistic simulations of the total repository system are performed, and results are evaluated against performance metrics. Uncertainty and sensitivity analyses help prioritize additional research and model development. The United States Department of Energy (DOE) has been developing a state of the art PA simulation software tool-kit, Geologic Disposal Safety Assessment (GDSA) Framework, that couples increasingly higher fidelity models of subsystem processes into total system PA simulations. Over the past several years, PAs of generic repositories in three geologic media (salt, shale, and crystalline rock) have demonstrated ongoing developments in capability. The current PA of a nuclear waste repository in a generic shale formation showcases GDSA Framework, including capabilities in domain discretization (Cubit), multi-physics simulations (PFLOTRAN), uncertainty and sensitivity analysis (Dakota), and visualization (Paraview). The generic shale reference case considers the disposal of 22,000 metric tons heavy metal of commercial spent nuclear fuel (SNF) in a generic shale formation. PA simulations account for the thermal load and radionuclide inventory of the waste form, components of the engineered barrier system including waste package and buffer, and components of the natural barrier system including the host rock shale and underlying and overlying stratigraphic units. Two repository layouts are considered, one for emplacement of waste packages containing 12 pressurized water reactor (PWR) assemblies, and one for 4-PWR waste package emplacement. Model domains are half-symmetry and contain between 7 and 22 million grid cells. Grid refinement captures the detail of individual waste packages, emplacement drifts, access drifts, and shafts. Simulations are run in a high-performance computing (HPC) environment on 512 or 2048 processes. The governing equations describing coupled heat and fluid flow and reactive transport are solved with PFLOTRAN, an open-source, massively parallel multiphase flow and reactive transport code. Additional simulated processes include waste package degradation; waste form dissolution; radioactive decay and ingrowth in aqueous, solid, and adsorbed phases; and a simple biosphere based on a pumping well. Simulations are run to 106 y; dose and radionuclide concentrations are observed within aquifers at a point approximately 5 km downgradient of the repository. Dakota is used to sample likely ranges of input parameters including waste form and waste package degradation rates and properties of engineered and natural materials to quantify uncertainty in predicted concentrations and sensitivity to input parameters. Given the assumptions of the reference case, results of 12-PWR and 4-PWR simulations are very similar. Dose at the pumping well 5-km downgradient of the repository does not exceed 10-9 Sv/y. I-129 is the largest contributor to dose. The dose due to Cl-36, the next largest contributor, is orders of magnitude smaller. At the pumping well, the concentration of I-129 5-km downgradient does not exceed 10-14 mol/L. Rank correlation coefficients indicate that the maximum concentration of I-129 within the aquifer is sensitive to shale (repository host rock) porosity and aquifer permeability. These results affirm that HPC-capable codes can be used to simulate important multi-physics couplings directly in a total system performance assessment of a geologic repository, and can be used in prioritization of future research and model development. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Structure and thermodynamics of uranium-containing iron garnets

Use of crystalline garnet as a waste form phase appears to be advantageous for accommodating actinides from nuclear waste. Previous studies show that large amounts of uranium (U) and its analogues such as cerium (Ce) and thorium (Th) can be incorporated into the garnet structure. In this study, we synthesized U loaded garnet phases, Ca 3 U x Zr 2–x Fe 3 O 12 (x = 0.5–0.7), along with the endmember phase, Ca 3 (Zr 2 )SiFe 3+ 2 O 12 , for comparison. The oxidation states of U were determined by X-ray photoelectron and absorption spectroscopies, revealing the presence of mixed pentavalent and hexavalent uranium in the phases with x = 0.6 and 0.7. The oxidation states and coordination environments of Fe were measured using transmission 57 Fe-Mössbauer spectroscopy, which shows that all iron is tetrahedrally coordinated Fe 3+ . U substitution had a significant effect on local environments, the extent of U substitution within this range had a minimal effect on the structure, and unlike in the x = 0 sample, Fe exists in two different environments in the substituted garnets. The enthalpies of formation of garnet phases from constituent oxides and elements were first time determined by high temperature oxide melt solution calorimetry. The results indicate that these substituted garnets are thermodynamically stable under reducing conditions. Furthermore, our structural and thermodynamic analysis further provides explanation for the formation of natural uranium garnet, elbrusite-(Zr), and supports the potential use of Ca 3 U x Zr 2–x Fe 3 O 12 as viable waste form phases for U and other actinides.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Cast Stone Oxidation Front Evaluation: Preliminary Results For Samples Exposed To Moist Air

The rate of oxidation is important to the long-term performance of reducing salt waste forms because the solubility of some contaminants, e.g., technetium, is a function of oxidation state. TcO 4 - in the salt solution is reduced to Tc(IV) and has been shown to react with ingredients in the waste form to precipitate low solubility sulfide and/or oxide phases. Upon exposure to oxygen, the compounds containing Tc(IV) oxidize to the pertechnetate ion, Tc(VII)O 4 - , which is very soluble. Consequently the rate of technetium oxidation front advancement into a monolith and the technetium leaching profile as a function of depth from an exposed surface are important to waste form performance and ground water concentration predictions. An approach for measuring contaminant oxidation rate (effective contaminant specific oxidation rate) based on leaching of select contaminants of concern is described in this report. In addition, the relationship between reduction capacity and contaminant oxidation is addressed. Chromate (Cr(VI) was used as a non-radioactive surrogate for pertechnetate, Tc(VII), in Cast Stone samples prepared with 5 M Simulant. Cast Stone spiked with pertechnetate was also prepared and tested. Depth discrete subsamples spiked with Cr were cut from Cast Stone exposed to Savannah River Site (SRS) outdoor ambient temperature fluctuations and moist air. Depth discrete subsamples spiked with Tc-99 were cut from Cast Stone exposed to laboratory ambient temperature fluctuations and moist air. Similar conditions are expected to be encountered in the Cast Stone curing container. The leachability of Cr and Tc-99 and the reduction capacities, measured by the Angus-Glasser method, were determined for each subsample as a function of depth from the exposed surface. The results obtained to date were focused on continued method development and are preliminary and apply to the sample composition and curing / exposure conditions described in this report. The Cr oxidation front (depth to which soluble Cr was detected) for the Cast Stone sample exposed for 68 days to ambient outdoor temperatures and humid air (total age of sample was 131 days) was determined to be about 35 mm below the top sample surface exposed. The Tc oxidation front, depth at which Tc was insoluble, was not determined. Interpretation of the results indicates that the oxidation front is at least 38 mm below the exposed surface. The sample used for this measurement was exposed to ambient laboratory conditions and humid air for 50 days. The total age of the sample was 98 days. Technetium appears to be more easily oxidized than Cr in the Cast Stone matrix. The oxidized forms of Tc and Cr are soluble and therefore leachable. Longer exposure times are required for both the Cr and Tc spiked samples to better interpret the rate of oxidation. Tc spiked subsamples need to be taken further from the exposed surface to better define and interpret the leachable Tc profile. Finally Tc(VII) reduction to Tc(IV) appears to occur relatively fast. Results demonstrated that about 95 percent of the Tc(VII) was reduced to Tc(IV) during the setting and very early stage setting for a Cast Stone sample cured 10 days. Additional testing at longer curing times is required to determine whether additional time is required to reduce 100 % of the Tc(VII) in Cast Stone or whether the Tc loading exceeded the ability of the waste form to reduce 100 % of the Tc(VII). Additional testing is required for samples cured for longer times. Depth discrete subsampling in a nitrogen glove box is also required to determine whether the 5 percent Tc extracted from the subsamples was the result of the sampling process which took place in air. Reduction capacity measurements (per the Angus-Glasser method) performed on depth discrete samples could not be correlated with the amount of chromium or technetium leached from the depth discrete subsamples or with the oxidation front inferred from soluble chromium and technetium (i.e., effective Cr and Tc oxidation fronts). Residual reduction capacity in the oxidized region of the test samples indicates that the remaining reduction capacity is not effective in re-reducing Cr(VI) or Tc(VII) in the presence of oxygen. Depth discrete sampling and leaching is a useful for evaluating Cast Stone and other chemically reducing waste forms containing ground granulated blast furnace slag (GGBFS) or other reduction / sequestration reagents to control redox sensitive contaminant chemistry and leachability in the near surface disposal environment. Based on results presented in this report, reduction capacity measured by the Angus-Glasser Ce(IV) method is not an appropriate or meaningful parameter for determining or predicting Tc and Cr oxidation / retentions, speciation, or solubilities in cementitious materials such as Cast Stone. A model for predicting Tc(IV) oxidation to soluble Tc(VII) should consider the waste form porosity (pathway for oxygen ingress), oxygen source, and the contaminant specific oxidation rates and oxidation fronts. Depth discrete sampling of materials exposed to realistic conditions in combination with short term leaching of crushed samples has potential for advancing the understanding of factors influencing performance. This information can be used to support conceptual model development.

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DOE OSTI2011

SALTSTONE VAULT CLASSIFICATION SAMPLES MODULAR CAUSTIC SIDE SOLVENT EXTRACTION UNIT/ACTINIDE REMOVAL PROCESS WASTE STREAM APRIL 2011

Savannah River National Laboratory (SRNL) was asked to prepare saltstone from samples of Tank 50H obtained by SRNL on April 5, 2011 (Tank 50H sampling occurred on April 4, 2011) during 2QCY11 to determine the non-hazardous nature of the grout and for additional vault classification analyses. The samples were cured and shipped to Babcock & Wilcox Technical Services Group-Radioisotope and Analytical Chemistry Laboratory (B&W TSG-RACL) to perform the Toxic Characteristic Leaching Procedure (TCLP) and subsequent extract analysis on saltstone samples for the analytes required for the quarterly analysis saltstone sample. In addition to the eight toxic metals - arsenic, barium, cadmium, chromium, mercury, lead, selenium and silver - analytes included the underlying hazardous constituents (UHC) antimony, beryllium, nickel, and thallium which could not be eliminated from analysis by process knowledge. Additional inorganic species determined by B&W TSG-RACL include aluminum, boron, chloride, cobalt, copper, fluoride, iron, lithium, manganese, molybdenum, nitrate/nitrite as Nitrogen, strontium, sulfate, uranium, and zinc and the following radionuclides: gross alpha, gross beta/gamma, 3H, 60Co, 90Sr, 99Tc, 106Ru, 106Rh, 125Sb, 137Cs, 137mBa, 154Eu, 238Pu, 239/240Pu, 241Pu, 241Am, 242Cm, and 243/244Cm. B&W TSG-RACL provided subsamples to GEL Laboratories, LLC for analysis for the VOCs benzene, toluene, and 1-butanol. GEL also determines phenol (total) and the following radionuclides: 147Pm, 226Ra and 228Ra. Preparation of the 2QCY11 saltstone samples for the quarterly analysis and for vault classification purposes and the subsequent TCLP analyses of these samples showed that: (1) The saltstone waste form disposed of in the Saltstone Disposal Facility in 2QCY11 was not characteristically hazardous for toxicity. (2) The concentrations of the eight RCRA metals and UHCs identified as possible in the saltstone waste form were present at levels below the UTS. (3) Most of the inorganic species measured in the leachate do not exceed the MCL, SMCL or TW limits. (4) The inorganic waste species that exceeded the MCL by more than a factor of 10 were nitrate, nitrite and the sum of nitrate and nitrite. (5) Analyses met all quality assurance specifications of US EPA SW-846. (6) The organic species (benzene, toluene, 1-butanol, phenol) were either not detected or were less than reportable for the vault classification samples. (7) The gross alpha and radium isotopes could not be determined to the MCL because of the elevated background which raised the detection limits. (8) Most of the beta/gamma activity was from 137Cs and its daughter 137mBa. (9) The concentration of 137Cs and 90Sr were present in the leachate at concentrations 1/40th and 1/8th respectively than in the 2003 vault classification samples. The saltstone waste form placed in the Saltstone Disposal Facility in 2QCY11 met the SCHWMR R.61-79.261.24(b) RCRA metals requirements for a nonhazardous waste form. The TCLP leachate concentrations for nitrate, nitrite and the sum of nitrate and nitrite were greater than 10x the MCLs in SCDHEC Regulations R.61-107.19, Part I A, which confirms the Saltstone Disposal Facility classification as a Class 3 Landfill. The saltstone waste form placed in the Saltstone Disposal Facility in 2QCY11 met the R.61-79.268.48(a) non wastewater treatment standards.

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