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At least 577 records · Page 32

IMPACT OF THE SMALL COLUMN ION EXCHANGE PROCESS ON THE DEFENSE WASTE PROCESSING FACILITY - 12112

The Savannah River Site (SRS) is investigating the deployment of a parallel technology to the Salt Waste Processing Facility (SWPF, presently under construction) to accelerate high activity salt waste processing. The proposed technology combines large waste tank strikes of monosodium titanate (MST) to sorb strontium and actinides with two ion exchange columns packed with crystalline silicotitanate (CST) resin to sorb cesium. The new process was designated Small Column Ion Exchange (SCIX), since the ion exchange columns were sized to fit within a waste storage tank riser. Loaded resins are to be combined with high activity sludge waste and fed to the Defense Waste Processing Facility (DWPF) for incorporation into the current glass waste form. Decontaminated salt solution produced by SCIX will be fed to the SRS Saltstone Facility for on-site immobilization as a grout waste form. Determining the potential impact of SCIX resins on DWPF processing was the basis for this study. Accelerated salt waste treatment is projected to produce a significant savings in the overall life cycle cost of waste treatment at SRS.

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Milestone Report - M3FT-15OR03120215 - Recommend HIP Conditions for AgZ

The purpose of this study was to continue research to determine if HIPing could directly convert I-Ag 0 Z into a suitable waste form. Fiscal year (FY) 2015 work completed studies of Phase IIA, IIB, and IIC samples. Product consistency testing (PCT) of Phase IIA samples resulted in iodine release below detection limit for six of twelve samples. This is promising and indicates that a durable waste form may be produced through HIPing even if transformation of the zeolite to a distinct mineral phase does not occur. From PCT results of Phase IIA samples, it was determined that future pressing should be conducted at a temperature of 900°C. Phase IIC testing continued production of samples to examine the effects of multiple source materials, compositional variations, and an expanded temperature range. The density of each sample was determined and x-ray diffraction (XRD) patterns were obtained. In all cases, there was nothing in the XRD analyses to indicate the creation of any AgI-containing silicon phase; the samples were found to be largely amorphous.

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Saltstone 3QCY12 TCLP Results

A Saltstone waste form was prepared in the Savannah River National Laboratory (SRNL) from a Tank 50H sample and Z-Area premix material for the third quarter of calendar year 2012 (3QCY12). After a 34 day cure, samples of the saltstone were collected, and the waste form was shown to meet the South Carolina Hazardous Waste Management Regulations (SCHWMR) R.61-79.261.24 and R.61-79.268.48(a) requirements for a nonhazardous waste form with respect to RCRA metals and underlying hazardous constituents. These analyses met all quality assurance specifications of USEPA SW-846.

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Saltstone 4QCY12 TCLP results

A Saltstone waste form was prepared in the Savannah River National Laboratory (SRNL) from a Tank 50H sample and Z-Area premix material for the fourth quarter of calendar year 2012 (4QCY12). After a 48 day cure, samples of the saltstone were collected, and the waste form was shown to meet the South Carolina Hazardous Waste Management Regulations (SCHWMR) R.61-79.261.24 and R.61-79.268.48(a) requirements for a nonhazardous waste form with respect to RCRA metals and underlying hazardous constituents. These analyses met all quality assurance specifications of USEPA SW-846.

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Saltstone 1QCY13 TCLP Results

A Saltstone waste form was prepared in the Savannah River National Laboratory (SRNL) from a Tank 50H sample and Z-Area premix material for the first quarter of calendar year 2013 (1QCY13). After a 49 day cure, samples of the saltstone were collected, and the waste form was shown to meet the South Carolina Hazardous Waste Management Regulations (SCHWMR) R.61-79.261.24 and R.61-79.268.48(a) requirements for a nonhazardous waste form with respect to RCRA metals and underlying hazardous constituents. These analyses met all quality assurance specifications of USEPA SW-846.

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Saltstone 2QCY13 TCLP Results

A Saltstone waste form was prepared in the Savannah River National Laboratory (SRNL) from a Tank 50H sample and Z-Area premix material for the second quarter of calendar year 2013 (2QCY13). After a 49 day cure, samples of the saltstone were collected, and the waste form was shown to meet the South Carolina Hazardous Waste Management Regulations (SCHWMR) R.61-79.261.24 and R.61-79.268.48(a) requirements for a nonhazardous waste form with respect to RCRA metals and underlying hazardous constituents. These analyses met all quality assurance specifications of USEPA SW-846.

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Saltstone 3QCY13 TCLP Results

A Saltstone waste form was prepared in the Savannah River National Laboratory (SRNL) from a Tank 50H sample and Z-Area premix material for the third quarter of calendar year 2013 (3QCY13). After a 63 day cure, samples of the saltstone were collected, and the waste form was shown to meet the South Carolina Hazardous Waste Management Regulations (SCHWMR) R.61-79.261.24 and R.61-79.268.48(a) requirements for a nonhazardous waste form with respect to RCRA metals and underlying hazardous constituents. These analyses met all quality assurance specifications of USEPA SW-846.

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Saltstone 4QCY13 TCLP Results

A Saltstone waste form was prepared in the Savannah River National Laboratory (SRNL) from a Tank 50H sample and Z-Area premix material for the fourth quarter of calendar year 2013 (4QCY13). After a 62 day cure, samples of the saltstone were collected, and the waste form was shown to meet the South Carolina Hazardous Waste Management Regulations (SCHWMR) R.61-79.261.24 and R.61-79.268.48(a) requirements for a nonhazardous waste form with respect to RCRA metals and underlying hazardous constituents. These analyses met all quality assurance specifications of USEPA SW-846.

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Saltstone 1QCY14 TCLP Results

A Saltstone waste form was prepared in the Savannah River National Laboratory (SRNL) from a Tank 50H sample and Z-Area premix material for the first quarter of calendar year 2014 (1QCY14). After a 64 day cure, samples of the saltstone were collected, and the waste form was shown to meet the South Carolina Hazardous Waste Management Regulations (SCHWMR) R.61-79.261.24 and R.61-79.268.48(a) requirements for a nonhazardous waste form with respect to RCRA metals and underlying hazardous constituents. These analyses met all quality assurance specifications of USEPA SW-846.

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Saltstone 3QCY14 TCLP results

A saltstone waste form was prepared in the Savannah River National Laboratory (SRNL) from a Tank 50H sample and Z-Area premix material for the third quarter of calendar year 2014 (3QCY14). After a 78 day cure, samples of the saltstone were collected, and the waste form was shown to meet the South Carolina Hazardous Waste Management Regulations (SCHWMR) R.61-79.261.24 and R.61-79.268.48(a) requirements for a nonhazardous waste form with respect to RCRA metals and underlying hazardous constituents. These analyses met all quality assurance specifications of USEPA SW-846.

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Performance Assessment Results for the Hanford Integrated Disposal Facility - 19134

In 2014, the US DOE Office of River Protection and its contractors began to develop a performance assessment for the near-surface disposal of low-level and mixed low-level waste at the Hanford Integrated Disposal Facility. The Hanford Integrated Disposal Facility is a doubly-lined landfill that was constructed between 2004 and 2006 to be the disposal facility for the vitrified low-activity waste containers that will be produced at the Hanford Waste Treatment and Immobilization Plant. The disposal facility is also expected to receive solid secondary waste produced at the Hanford Waste Treatment and Immobilization Plant and other solid wastes from site activities. The constructed facility is awaiting authorization to receive waste. Although no waste destined for the facility is currently being generated, operations cannot begin until the facility has received a Disposal Authorization Statement from US DOE and a permit modification from the State of Washington Department of Ecology. Both the Disposal Authorization Statement and permit modification require a demonstration that the system of engineered and natural features of the disposal facility will limit releases from the facility and be protective of human health and the environment for the next 1,000 to 10,000 years. Based on the analyses presented in the Integrated Disposal Facility Performance Assessment, US DOE issued a conditional Disposal Authorization Statement for the facility in June 2018. Demonstrating the long-term performance of the facility to be protective of human health and the environment was evaluated under the requirements of US DOE Order 435.1, Radioactive Waste Management. Computer simulations were performed to evaluate whether or not the disposal facility would comply with the performance objectives specified in the order. The evaluation was performed using numerical models to simulate source term releases from vitrified and cement-based waste forms. Source term releases were subsequently transported through an air pathway and a groundwater pathway to a point of compliance 100 meters from the disposed of waste. In addition, an analysis evaluating the consequences of an inadvertent intrusion into the waste was also performed. In accordance with US DoE's technical standard for performance assessments, deterministic and probabilistic analyses were performed to demonstrate that the Hanford Integrated Disposal Facility, together with the engineered features of the waste forms and natural features of the selected site, provides a reasonable expectation that the low-level waste and mixed low-level waste disposal will meet the radiological performance objectives established by US DOE. Post-closure doses to future members of the public were evaluated after the time of closure and included a 100-year period of institutional controls that limited exposure immediately following closure of the facility. Calculated doses to a future member of the public from the air pathway were below the 100 μSv/yr dose performance objective. Calculated doses to a future member of the public from all exposure pathways, including the air pathway and the groundwater pathway, were below the 250 μSv/yr dose performance objective. Doses following an inadvertent intrusion into the facility after the institutional control period were below the 5 mSv acute dose performance measure and the 1 mSv/yr chronic dose performance measure. (authors)

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GLASSES CONTAINING IRON (II III) OXIDES FOR IMMOBILIZATION OF RADIOACTIVE TECHNETIUM

Technetium-99 (Tc-99) has posed serious environmental threats as US Department of Energy's high-level waste. This work reports the vitrification of Re, as surrogate for Tc-99, by iron-borosilicate and iron-phosphate glasses, respectively. Iron-phosphate glasses can dissolve Re as high as {approx} 1.2 wt. %, which can become candidate waste forms for Tc-99 disposal, while borosilicate glasses can retain less than 0.1 wt. % of Re due to high melting temperature and long melting duration. Vitrification of Re as Tc-99's mimic was investigated using iron-borosilicate and iron-phosphate glasses. The retention of Re in borosilicate glasses was less than 0.1 wt. % and more than 99 wt. % of Re were volatilized due to high melting temperature and long melting duration. Because the retention of Re in iron-phosphate glasses is as high as 1.2 wt. % and the volatilization is reduced down to {approx}50 wt. %, iron-phosphate glasses can be one of the glass waste form candidates for Tc (or Re) disposal. The investigations of chemical durability and leaching test of iron-phosphate glasses containing Re are now underway to test the performance of the waste form.

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Modeling and Analysis of the Transport and Disposal of Beryllium Moderator Blocks and Greater than Class C (GTCC) Waste

Radioactive waste in the United states is categorized based on its source, radioactivity, and security risk. The categorization method employed by the Nuclear Regulatory Committee (NRC) for low-level radioactive waste is not currently applied to any waste generated by the Department of Energy (DOE) or disposed of in DOE facilities. As a result, there is a large volume of DOE-generated waste with characteristics similar to the NRC’s “Greater than Class C” (GTCC) waste category which presently do not have a path for long-term disposal. Much of this waste cannot undergo the standard commercial disposition process due to it being categorized as Transuranic (TRU) waste under DOE guidelines, due to elevated concentrations of key fission products. Possible solutions to this dilemma are explored in the Environmental Impact Statement for the “Disposal of Greater-Than-Class-C Low Level Radioactive Wave” (EIS-0375). This paper analyzes several EIS possible paths for disposal for this “orphaned” waste. One example that highlights this categorization issue is the spent beryllium cladding that has been extracted from the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) throughout its operational lifetime. These beryllium blocks surround the reactor’s main chambers and components and serve as neutron moderators. This paper will use this particular waste form to analyze the economic and technological viability of transporting and storing this material under the storage and transportation criteria of the Waste isolation Pilot Plant (WIPP), which was deemed the most feasible disposition path according to the EIS. The example of this waste form used in this paper that highlights this categorization issue is the spent beryllium cladding extracted from Idaho National Laboratory’s (INL) Advanced Test Reactor (ATR). These beryllium blocks surround the reactor’s components and serve as neutron moderators, and thus have accumulated high concentration of high-activity transuranic isotopes. This paper analyzed the viability of transporting and storing this material under the storage and transportation criteria of the five primary disposition paths covered in EIS-0375. An array of calculations was carried out to assess the viability of the various disposition paths for the beryllium shipments as well as other waste shipments that fall within the GTCC category. Modeling with MCNP 6.2 was conducted to determine whether the beryllium blocks could be safely stored and transported within a 72-B cask; the standard shipping container for remote-handled, transuranic waste, while also meeting regulatory limits at various disposition paths. A cost analysis of the transportation and long-term disposal paths mentioned in the EIS was also carried out using available data and information from similar waste shipments. Lastly, a geochemical analysis of the various geological repository discussed in the EIS was also carried out using the Geochemists Workbench Release 14. Long-term disposal of the beryllium blocks at the Waste Isolation Pilot Plant (WIPP) proved to be the most cost-effective long-term disposition option of the ones considered in the EIS. A dose rate calculation at both contact and remote-handling distances indicate that the analyzed beryllium shipments should not exceed the exposure limits at any of the considered locations. Greater-than-class-C waste has been left in a regulatory state of limbo for years, resulting in backlogs of inventory across several research sites in the United States. Due to its high activity and presence of transuranic isotopes, it is imperative to ensure that it remains inaccessible and sequestered both in its short-term interim as well as a long-term geologic time scale. The information and assessments done in the paper could potentially serve as a reference for any future shipments of this waste form at the WIPP facility as well as other disposition paths that may be considered in the future.

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SNL Sigma Off-Gas Team Contribution to the FY15 DOE/NE-MRWFD Campaign Accomplishments Report

This program at Sandia is focused on Iodine waste form development for Fuel Cycle R&D needs. Our research has a general theme of “Capture and Storage of Iodine Fission Gas “ in which we are focused on silver loaded zeolite waste forms, evaluation of iodine loaded getter materials (eg., mordenite zeolite), and the development of low temperature glass waste forms that successfully incorporate iodine loaded getter materials from I 2 , organic iodide, etc. containing off-gas streams.

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Hydraulic Properties of Grout Monoliths Containing Organic Ion Exchange Resin - 18369

The Waste Treatment and Immobilization Plant (WTP) at Hanford is being constructed to treat 56 million gallons of radioactive waste currently stored in underground tanks at the Hanford site. Operation of the WTP will generate several solid secondary waste (SSW) streams including spent spherical resorcinol formaldehyde cation exchange resin (IXr). The current plan for safe disposal of IXr is solidification in cementitious material (CM) typically comprised of cement, fly ash and blast furnace slag. Three CM formulations were tested to stabilize IXr for long-term land disposal. Two IXr loading rates, 0.1 and 0.3 v/v IXr:grout were used with the three CM formulations to prepare simulated waste form monoliths for hydraulic property testing. Following curing, the monoliths were removed from molds and tested to determine saturated hydraulic conductivity (K{sub sat}) and water retention properties. K{sub sat} was measured using a standard flexible permeameter method ASTM D-5084-10. A constant vapor pressure (CVP) method was used to determine van Genuchten parameters for desorption moisture retention curves for each of the SSW grout formulations. For the CVP method, vacuum saturated, intact wafers approximately 1.25 cm thick were placed above a saturated salt solution inside a sealed container for equilibration. Each saturated salt solution has an associated vapor pressure related to relative humidity in the container. Visual examination and three-dimensional x-ray micro computed tomography (μCT) indicated that the IXr was uniformly distributed in the cured waste form. Saturated hydraulic conductivity was within the range previously reported by Flach et al. in a study of the hydraulic properties of 98 pastes and grouts. The solidification of IXr in CM resulted in a waste form with water retention properties similar to cementitious pastes. (authors)

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Separated waste stream immobilization of iodine and off-gas caustic scrubber solution (Final Report)

In this work, the approach was to immobilize the CS solution in a glass-bonded composite of cancrinite/sodalite. The synthesis of ceramic waste forms was accomplished by low-temperature (~90°C) processes developed by the team in the framework of a previous NEUP project. By building on our earlier efforts for iodine waste forms, we minimized the schedule and technical risk while still providing novel process demonstrations for the off-gas waste forms. In the case of AgI-based sorbents, a proof-of-concept stripping process was shown to remove iodine from AgI by converting it to NaI(aq) followed by immobilization of these ions in iodosodalite (ISOD), which could be glass-bonded in the same way as for the CS solution cancrinite/sodalite powders.

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Microstructure and Cs Behavior of Ba-Doped Aluminosilicate Pollucite Irradiated with F+ Ions

Radionuclide 137Cs is one of the major fission products that dominate heat generation in spent fuels over the first 300 hundred years. A durable waste form for 137Cs that decays to 137Ba is needed to minimize its environmental impact. Aluminosilicate pollucite CsAlSi2O6 is selected as a model waste form to study the decay-induced structural effects. While Ba-containing precipitates are not present in charge-balanced Cs0.9Ba0.05AlSi2O6, they are found in Cs0.9Ba0.1AlSi2O6 and identified as monoclinic Ba2Si3O8. Pollucite is susceptible to electron irradiation induced amorphization. The threshold density of the electronic energy deposition for amorphization is determined to be ~235 keV/nm3. Pollucite can be readily amorphized under F+ ion irradiation at 673 K. A significant amount of Cs diffusion and release from the amorphized pollucite is observed during the irradiation. However, cesium is immobile in the crystalline structure under He+ ion irradiation at room temperature. The critical temperature for amorphization is not higher than 873 K under F+ ion irradiation. If kept at or above 873 K all the time, the pollucite structure is unlikely to be amorphized; Cs diffusion and release are improbable. A general discussion regarding pollucite as a potential waste form is provided in this report.

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Iodine solid sorbent design: a literature review of the critical criteria for consideration

Designing sorbents for iodine capture in different conditions requires selection and optimization of a large and diverse range of variables. These variables fall into general categories (or features) of sorbent activity, sorbent stability, and the fate of the loaded material in terms of the disposal (waste form) options available. To illustrate, silver-loaded, high-porosity sorbents make for maximized iodine capture and less pressure drop in a column-based sorption system approach, however, this high porosity can lead to less mechanically stable sorbents. Additionally, waste forms containing silver must also be compliant with additional criteria for hazardous waste disposal. Thus, all these aspects must be considered simultaneously when selecting a sorbent for utilization under specific conditions. Information is given for different types of sorbent design considerations for different operating conditions and some emphasis is also given on promising alternatives for silver as the active (chemisorption-based) getter metal. Discussion is given around demonstrated options for waste forms for different metal-iodide compounds.

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