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Treatment of radioactive liquid waste by the forced-air exhaust system - 15190

A Natural Evaporation Facility (NEF) was constructed to treat low level (LL) radioactive liquid waste arising from a Radioactive Waste Treatment Facility (RWTF) at the Korea Atomic Energy Research Institute (KAERI). In order to process LL radioactive liquid waste, we investigated the influence of climate condition, mainly atmospheric humidity and temperature, on the amount of evaporation in a forced-air exhaust system in KAERI. Evaporation media were made of cotton and polyester. Air exhaust in the facility was forced by exhaust fans. The evaporation rate and the decontamination factor were calculated and reported as shown in the results section. The evaporation and air supply flow increased at the same rate. In addition, the liquid waste feed rate and temperature of supplied air increased. Regarding the relative humidity of supplied air, the evaporation rate increased as the relative humidity decreased. As the result of this study, the operating conditions for the NEF were optimized in the following areas: The air temperature above 8 deg. C, the air relative humidity below 70%, the air flow rate 1.14-1.47 m/sec, and the liquid waste feed rate 4.6 l/hr.m{sup 2}. The decontamination factor was 1.1x10{sup 3}. (authors)

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Deployment of Low Temperature Aluminum Dissolution (LTAD) Technology to Retrieve H-Modified (HM) Sludge in SRS Tank 15 – 25672

Tank 15 is a 4,234,000-liter (1,118,500-gallon) Type 2 high-level waste storage tank located in H Tank Farm at the Savannah River Site. It was put into service in 1960 to receive high-activity, H-Modified (HM) waste from H Canyon. Between June 1964 and November 1972, the waste tank was filled six times, and supernate was decanted five times, leaving behind the sludge solids. Tank 15 also received a mixture of high-activity and low-activity HM waste from Tank 16. Tank 15 has more recently undergone several mixing campaigns to remove much of the sludge waste; however, the effectiveness of suspending the sludge heel via mechanical mixing has significantly diminished. Low Temperature Aluminum Dissolution (LTAD) is a process developed for the dissolution of suspended aluminum solids in a large waste storage tank. Originally intended for deployment during the preparation of sludge batches in H Tank Farm for the Defense Waste Processing Facility (DWPF), the process involves maintaining the waste storage tank at a slightly elevated temperature and highly alkaline chemistry to facilitate dissolution of aluminum solids. As mechanical heel removal efforts diminished in effectiveness in Tank 15, LTAD was selected to both reduce the volume of sludge solids remaining in the heel and to modify the sludge rheology to facilitate the suspension of additional solids using the installed mixing devices.

Campbell, Seth G.↗

Waste Treatment Plant - 12508

The Waste Treatment Plant (WTP) will immobilize millions of gallons of Hanford's tank waste into solid glass using a proven technology called vitrification. The vitrification process will turn the waste into a stable glass form that is safe for long-term storage. Our discussion of the WTP will include a description of the ongoing design and construction of this large, complex, first-of-a-kind project. The concept for the operation of the WTP is to separate high-level and low-activity waste fractions, and immobilize those fractions in glass using vitrification. The WTP includes four major nuclear facilities and various support facilities. Waste from the Tank Farms is first pumped to the Pretreatment Facility at the WTP through an underground pipe-in-pipe system. When construction is complete, the Pretreatment Facility will be 12 stories high, 540 feet long and 215 feet wide, making it the largest of the four major nuclear facilities that compose the WTP. The total size of this facility will be more than 490,000 square feet. More than 8.2 million craft hours are required to construct this facility. Currently, the Pretreatment Facility is 51 percent complete. At the Pretreatment Facility the waste is pumped to the interior waste feed receipt vessels. Each of these four vessels is 55-feet tall and has a 375,000 gallon capacity, which makes them the largest vessels inside the Pretreatment Facility. These vessels contain a series of internal pulse-jet mixers to keep incoming waste properly mixed. The vessels are inside the black-cell areas, completely enclosed behind thick steel-laced, high strength concrete walls. The black cells are designed to be maintenance free with no moving parts. Once hot operations commence the black-cell area will be inaccessible. Surrounded by black cells, is the 'hot cell canyon'. The hot cell contains all the moving and replaceable components to remove solids and extract liquids. In this area, there is ultrafiltration equipment, cesium-ion exchange columns, evaporator boilers and recirculation pumps, and various mechanical process pumps for transferring process fluids. During the first phase of pretreatment, the waste will be concentrated using an evaporation process. Solids will be filtered out, and the remaining soluble, highly radioactive isotopes will be removed using an ion-exchange process. The high-level solids will be sent to the High-Level Waste (HLW) Vitrification Facility, and the low activity liquids will be sent to the Low-Activity Waste (LAW) Vitrification Facility for further processing. The high-level waste will be transferred via underground pipes to the HLW Facility from the Pretreatment Facility. The waste first arrives at the wet cell, which rests inside a black-cell area. The pretreated waste is transferred through shielded pipes into a series of melter preparation and feed vessels before reaching the melters. Liquids from various facility processes also return to the wet cell for interim storage before recycling back to the Pretreatment Facility. (authors)

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DEM and MELT In-Can Vitrification Process for Fukushima Daiichi Water Treatment Secondary Waste - 20034

Fukushima Daiichi Nuclear Power Station was severely damaged by a tsunami induced by the earthquake on March 11, 2011. A large amount of water treatment secondary waste has been generated from the purification work of contaminated water that consists of the water flowing into the reactor building. The Mid and Long-Term Road map specifies that solid waste at Fukushima Daiichi Nuclear Power Station should be stored safely and reasonably according to its characteristics. In order to meet this goal, a versatile treatment solution is needed, and the In-Can vitrification seems to be a suitable solution. The DEM and MELT In-Can vitrification process is being developed and qualified in France to treat waste arising from decommissioning and dismantling operations. This qualification is carried out through the consortium of the DEM and MELT project coordinated by CEA which gathers Orano, ECM technologies and Andra. The DEM and MELT In-Can vitrification process is a robust, simple and versatile in situ vitrification process. It is designed for high or intermediate level waste and is compact enough to be implemented in an existing facility or close to the waste to be treated. It is developed to treat liquid and solid waste, to produce a small amount of secondary waste and to minimize investment and operating costs. The DEM and MELT In-Can vitrification process is also developed with a modular design in order to be adapted to nuclear operators needs and requirements. The In-Can melting process has already been developed and fully qualified by the CEA for Alpha liquid waste and can be rated TRL 7 for this scope. The In-Can vitrification technology is considered as a qualified solution that could be applied for the treatment of waste resulting from the D and D work in Fukushima Daiichi Nuclear Power Station. The feasibility evaluation of the DEM and MELT In-Can vitrification process for Fukushima Daiichi water treatment secondary waste is a project led by Orano, CEA and ANADEC. This R and D project is performed by the funds from the Ministry of Economy, Trade and Industry as The Subsidy Program 'Project of Decommissioning and Contaminated Water Management'. This paper presents the first results of this feasibility evaluation, from the scenarios and glass formulation definition to the industrial implementation feasibility. (authors)

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New AREVA's Investment in Continuous Improvement to Increase Efficiency of La Hague Reprocessing Lines - 18588

New AREVA currently operates 2 commercial reprocessing lines at their La Hague facility in France. Each line is capable of reprocessing up to 800 tHM/year of commercial spent light water reactor fuel. The site and process is self-contained with all waste arising from the reprocessing operations segregated into various waste streams and stabilized at the site. New AREVA operates a culture of continuous improvement and continues to invest in new technology, processes, waste management and procedures that have increased plant throughput while reducing waste volumes, operating cost while at the same time maintaining our strong safety culture. New investments have also been driven by changing customer needs and new markets. A good example of this has emerged over the last several years as New AREVA has been transporting a large amount of Research Reactor (RR) Uranium-Aluminum (UAl) type spent fuel (SF) on behalf of international RR operators to the La Hague site. Last year more than 100 RR fuel elements shipments were carried out by New AREVA (for fresh and spent fuel). In our current commercial lines the reprocessing of RRSF has been limited to approximately 1.4 ton a year. To meet increasing demand for reprocessing of RRSF, the range of RRSF that can now be reprocessed at La Hague has been extended to include Uranium-Aluminium-Silicium fuel (USi also called Silicide fuel) and also non-intact RRSF. In addition, New AREVA supported large scale research, development and engineering programs to identify processes/technology to allow the La Hague Facilities to reprocess a targeted average of additional 25 tons/year of non-commercial fuel types (RRSF and also fast reactor fuel, specific MOX fuel... ). A new facility head end called Traitement de Combustibles Particuliers (TCP) facility is currently planned to be implemented in the La Hague plant allowing reprocessed special fuels to be introduced into the existing commercial processing lines. The purpose of the paper is to demonstrate New AREVA's culture of continuous improvement that has been implemented at our commercial reprocessing facility highlighting how we have invested in processes and procedures for improving services to the nuclear operators. A case study will be presented related to our latest development of the TCP head end that will allow New AREVA to reprocess a targeted average of additional 25 tons/year of specialized fuels. (authors)

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Machine Learning Surrogates of a Fuel Matrix Degradation Process Model for Performance Assessment of a Nuclear Waste Repository

Spent nuclear fuel repository simulations are currently not able to incorporate detailed fuel matrix degradation (FMD) process models due to their computational cost, especially when large numbers of waste packages breach. The current paper uses machine learning to develop artificial neural network and k-nearest neighbor regression surrogate models that approximate the detailed FMD process model while being computationally much faster to evaluate. Further, using fuel cask temperature, dose rate, and the environmental concentrations of CO 3 2- , O 2 , Fe 2+ , and H 2 as inputs, these surrogates show good agreement with the FMD process model predictions of the UO 2 degradation rate for conditions within the range of the training data. A demonstration in a full-scale shale repository reference case simulation shows that the incorporation of the surrogate models captures local and temporal environmental effects on fuel degradation rates while retaining good computational efficiency.

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Modeling and Analysis of Coupled Flow Through Packed Beds as a Diagnostic Tool for Packed Bed Reactor Health - 17172

The US Department of Energy, Office of River Protection, is building the Waste Treatment and Immobilization Plant (WTP) in Richland, Washington, to immobilize the hundreds of millions of liters of radioactive waste stored at the Hanford Site in Washington state. The offgas generated during vitrification of these wastes requires treatment to remove and/or destroy chemical and radioactive hazardous components before release to the atmosphere. In particular, volatile and semi-volatile hydrocarbon components are to be catalytically oxidized to carbon dioxide and water, whereas nitrogen oxides (NOx) are to be catalytically reduced to nitrogen and water. These tasks are accomplished in a series of packed bed reactors. Achieving the requisite Destruction/Removal Efficiency (DRE) for each of the tracked components is highly dependent on the flow characteristics of the offgas through the packed beds. The flow kinetics may be monitored and characterized during the initial warm-up of the system. The gases, initially at 76 deg. C (169 deg. F), (a) flow through one side of a countercurrent heat exchanger to recover heat from the exiting gases (which have already passed through the catalytic beds), (b) are then further heated electrically to approximately 400 deg. C (750 deg. F; the preferred operating temperature of the Thermo Catalytic Oxidizer (TCO)), (c) pass through the TCO where the reduced components of the offgas are oxidized over a supported Pt catalyst bed, (d) are partially cooled to approximately 315 deg. C (600 deg. F) by addition of approximately 25% (by weight) diluted ammonia/air mixture, (e) pass through the Selective Catalytic Reducer (SCR) where the NOx components are converted to N{sub 2} and H{sub 2}O over a supported V2O{sub 5}/TiO{sub 2} catalyst bed, and (f) exit through the hot side of the countercurrent heat exchanger. The electrical heater is not powerful enough to raise the temperature of the offgas to the desired 400 deg. C without the contribution of the heat recovered in the heat exchanger. Moreover, the thermal mass of the beds and housings of the TCO and SCR require a significant input of thermal energy to reach the desired operating temperatures; the better part of a day is required to bring the unit from ambient temperatures to operating temperatures. Therefore, the temperature of the stream during the initial warm-up of the system at each point in the system is dependent upon the thermal mass, flow dispersion, and permeability of the components, quantified as characteristic/representative volumes and residence times. By monitoring the temperature of the offgas as it enters and exits each portion of the system during this warm-up period one may model the system as a series of small Continuously Stirred Tank Reactors (CSTR)s (referred to as 'CSTRlets') to explain the flow kinetics through the catalyst beds. These component CSTRlets are adjustable in size and number, that one may model the flow profile of the gases through the catalyst beds. For example, a single large CSTR will present an effluent temperature profile with a simple exponential decay shape. Concatenating more CSTRlets generate more complex sigmoidal responses. Concatenating one CSTRlet significantly larger than the others in the train also alters the shape of the effluent temperature profile, in a fashion not replicable by a train of equal-sized CSTRlets. The ultimate goal is to accurately emulate the entire system as a representative CSTRlet train. Once accomplished, this characterization may be used in subsequent warm-ups (following shutdown for catalyst maintenance, for example) or by an intentional step change in inlet offgas temperature to detect/diagnose anomalies in flow through the beds (e.g., channeling or partial blockage.) This work presents development of this model through analytical, calculational, and empirical means. - The analytic model is accomplished by Laplace transform analysis of each component to recover component process transfer functions. - The calculational portion is performed via brute force finite element calculation. - Fitting empirical data from the warm-up of the TCO skid allows one to select the hypothetical CSTRlet trains which best fit the TCO and SCR beds for the analytical and finite element calculations. The first portion of this paper presents the development of the analytical (Laplace space) model of the coupled system. Estimated values for the initial temperature, mass, specific heat, electrical heat, and relative quantities of the offgas, catalyst beds, stainless steel housings, and ammonia dilution stream are used to generate expected temperature profiles at several selected junctures of the system. The number and size of the component hypothetical CSTRlets are calculated from the analytical model and are used to predict temperature profiles at the exit of the system and at intermediate junctions between system components (where thermocouples are present in the actual unit.) The finite element calculations employ a series of CSTRlet volumes to produce a similar suite of temperature profiles, for comparison with the analytical model results. Finally, empirical data from the Factory Acceptance Test system warm-up are used to select and compare the results from each modeling strategy to select the CSTRlet train(s) which best model the behavior of the actual system. This work presents a novel method of system component characterization, of value to the WTP plant and similar vitrification facilities as well as chemical process industries unrelated to waste management. (authors)

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Effect of Silica Particle Size of Nuclear Waste-to-Glass Conversion - 17319

The process for converting nuclear waste-to-glass in an electric melter occurs in the cold cap, a crust of reacting solids floating on the glass pool. As the melter feed (a mixture of the nuclear waste and glass forming and modifying additives) heats up in the cold cap, glass-forming reactions ensue, causing the feed matrix to connect, trapping reaction gases to create a foam layer. The foam layer reduces the rate of melting by separating the reacting feed from the melt pool. The size of the silica particle additives in the melter feed affects melt viscosity and, hence, foam stability. To investigate this effect, seven nuclear waste simulant feeds of a high-level waste were batched as slurries and prepared with dissimilar ranges of silica particle size. Each slurry feed was charged into a laboratory-scale melter (LSM) to produce a cold cap and the propensity of feeds to foam was determined by pressing dried feeds into pellets and monitoring the change of pellet volume in response to heating. Two of these slurries were designed to have dissimilar glass viscosities at 1150°C. In the low temperature region of the cold cap, before the melter feed connects, the feeds without fine silica particles behaved similar to the high viscosity feed as their volume contracted while the feed with silica particles no larger than 5 µm reacted like the low viscosity feed. However, the feed volume similarities reversed as the feed connected and expanded through the foam region of the cold cap.

Cold Cap↗

Effect of Silica Particle Size on Nuclear Waste-to-Glass Conversion - 17319

The process for converting nuclear waste-to-glass in an electric melter occurs in the cold cap, a crust of reacting solids floating on the glass pool. As the melter feed (a mixture of the nuclear waste and glass forming and modifying additives) heats up in the cold cap, glass-forming reactions ensue, causing the feed matrix to connect, trapping reaction gases to create a foam layer. The foam layer reduces the rate of melting by separating the reacting feed from the melt pool. The size of the silica particle additives in the melter feed affects melt viscosity and, hence, foam stability. To investigate this effect, seven nuclear waste simulant feeds of a high-level waste were batched as slurries and prepared with dissimilar ranges of silica particle size. Each slurry feed was charged into a laboratory-scale melter (LSM) to produce a cold cap and the propensity of feeds to foam was determined by pressing dried feeds into pellets and monitoring the change of pellet volume in response to heating. Two of these slurries were designed to have dissimilar glass viscosities at 1150 deg. C. In the low temperature region of the cold cap, before the melter feed connects, the feeds without fine silica particles behaved similar to the high viscosity feed as their volume contracted while the feed with silica particles no larger than 5 μm reacted like the low viscosity feed. However, the feed volume similarities reversed as the feed connected and expanded through the foam region of the cold cap. (authors)

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Phyto-screening for Volatile Organic Compounds in a Clay-based Glacial Till at the Niagara Falls Storage Site, New York - 17372

The Niagara Falls Storage Site (NFSS) and associated vicinity properties (VPs) occupy approximately 607 hectares (1,500 acres) of the original 3,035-hectare (7,500-acre) Lake Ontario Ordnance Works (LOOW). The NFSS was used to store radioactive materials, residues and waste from uranium ore processing between 1944 and 1952. The waste-handling practices produced an array of groundwater impacts throughout the site. Although the groundwater at the NFSS is naturally high in total dissolved solids that preclude untreated consumption, the groundwater contaminant levels in some areas of the site are very high and may pose risk to a Construction Worker receptor, as well as limit the options for future site re-use. Volatile organic compound (VOC) impacted soil in the northern portion of the site has produced a groundwater plume with significant concentrations of tetrachloroethene (PCE) and associated daughter compounds. Active groundwater remediation is impractical due to the clayey soils, as evident from adjacent property owners who have unsuccessfully attempted active remediation of their VOC plumes. Consequently, the Buffalo District may evaluate excavation options, with secondary phyto-remediative approaches as passive alternatives to lessen the groundwater impacts during the project's life cycle. A preliminary field-study in the VOC plume area sampled the interior of a Black Locust (Robinia pseudoacacia) tree using a Haglof tree corer. Detectable levels of tetrachloroethene (PCE), trichloroethene (TCE), and cis-1,2,-dichloroethene (cis-DCE) were found in the woody interior of the tree. Subsequent sampling and analyses were performed on proximal trees of varying species in and around the currently delineated VOC plume. The results indicate that poplar and native cottonwood variants preferentially contain the VOCs, whereas collocated hardwood trees produced non-detectable results. Additional sampling of Cottonwoods in the VOC-impacted area promoted further delineations of the impacts. The tree-sampling data can be considered a broad surrogate for soil sampling to optimally delineate the VOC plume without performing additional invasive sampling of the subsurface (e.g., drilling). The results also will be used to determine which native tree species preferentially uptake the VOCs from soil and groundwater. These species, and possibly other hybrid species known for VOC uptake and transpiration, would be propagated to promote interim phyto-extraction of the VOCs during the maintenance period between decision-document completion and likely remedial action under the Formerly Utilized Sites Remedial Action Program (FUSRAP). (authors)

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Progress and Technology Assessment of Melt-Cast Ceramic Waste Forms

The Savannah River National Laboratory (SRNL) conducted research from FY2011 – FY2020 to develop ceramic waste forms to treat High-Level Waste (HLW) resulting from reprocessing of commercial Used Nuclear Fuel (UNF) as part of the Department of Energy (DOE) office of Nuclear Energy (NE) Fuel Cycle Technologies (FCT) research program. The objective of the research was to develop a reference ceramic waste form and process technology needed to immobilize the combined HLW raffinate stream, including undissolved solids and soluble technetium, and potentially the TRU waste stream resulting from reprocessing processes. This report compiles the relevant literature produced under, and pertinent to, the program objectives and provides a perspective on the technology development as well as prospects for future.

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Secondary Waste Cast Stone Waste Form Qualification Testing Plan (Rev.1)

The Hanford Tank Waste Treatment and Immobilization Plant (WTP) is being constructed to treat the 56 million gallons of radioactive waste stored in 177 underground tanks at the Hanford Site. The WTP includes a pretreatment facility to separate the wastes into high-level waste (HLW) and low-activity waste (LAW) fractions for vitrification and disposal. The LAW will be converted to glass for final disposal at the Integrated Disposal Facility (IDF). Cast Stone – a cementitious waste form, has been selected for solidification of this secondary waste stream after treatment in the ETF. The secondary-waste Cast Stone waste form must be acceptable for disposal in the IDF. This secondary waste Cast Stone waste form qualification testing plan outlines the testing of the waste form and immobilization process to demonstrate that the Cast Stone waste form can comply with the disposal requirements. Specifications for the secondary-waste Cast Stone waste form have not been established. For this testing plan, Cast Stone specifications are derived from specifications for the immobilized LAW glass in the WTP contract, the waste acceptance criteria for the IDF, and the waste acceptance criteria in the IDF Permit issued by the State of Washington. This testing plan outlines the testing needed to demonstrate that the waste form can comply with these waste form specifications and acceptance criteria. The testing program must also demonstrate that the immobilization process can be controlled to consistently provide an acceptable waste form product. This testing plan also outlines the testing needed to provide the technical basis for understanding the long-term performance of the waste form in the disposal environment. These waste form performance data are needed to support performance assessment analyses of the long-term environmental impact of the secondary-waste Cast Stone waste form in the IDF

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Challenges with vitrification of Hanford High-Level Waste (HLW) to borosilicate glass – An overview

The Hanford site in Washington State houses ~56 million gallons of radioactive wastes stored in 177 underground tanks. The waste must be immobilized and permanently stored, and the plan is to separate the tank wastes into low activity waste (LAW) and high-level waste (HLW) streams. The U.S. Department of Energy is building a Waste Treatment and Immobilization Plant at Hanford site to separately vitrify these two waste streams in borosilicate glass using Joule-heated ceramic melters (JHCM). Although the process of nuclear waste vitrification seems to be well established, in practicality, it is faced with complex problems starting from the design of glass compositions, to processing in melters and long-term performance of the final vitrified waste forms. The article presents an overview of our current understanding of critical challenges related to the development and performance of HLW glasses.

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Nucleation and crystal growth behavior of nepheline in simulated high-level waste glasses

The Savannah River National Laboratory (SRNL) has been tasked with supporting glass formulation development and process control strategies in key technical areas, relevant to the Department of Energy’s Office of River Protection (DOE-ORP) and related to high-level waste (HLW) vitrification at the Waste Treatment and Immobilization Plant (WTP). Of specific interest is the development of predictive models for crystallization of nepheline (NaAlSiO4) in HLW glasses formulated at high alumina concentrations. This report summarizes recent progress by researchers at SRNL towards developing a predicative tool for quantifying nepheline crystallization in HLW glass canisters using laboratory experiments. In this work, differential scanning calorimetry (DSC) was used to obtain the temperature regions over which nucleation and growth of nepheline occur in three simulated HLW glasses - two glasses representative of WTP projections and one glass representative of the Defense Waste Processing Facility (DWPF) product. The DWPF glass, which has been studied previously, was chosen as a reference composition and for comparison purposes. Complementary quantitative X-ray diffraction (XRD) and optical microscopy confirmed the validity of the methodology to determine nucleation and growth behavior as a function of temperature. The nepheline crystallization growth region was determined to generally extend from ~ 500 to >850 °C, with the maximum growth rates occurring between 600 and 700 °C. For select WTP glass compositions (high Al2O3 and B2O3), the nucleation range extended from ~ 450 to 600 °C, with the maximum nucleation rates occurring at ~ 530 °C. For the DWPF glass composition, the nucleation range extended from ~ 450 to 750 °C with the maximum nucleation rate occurring at ~ 640 °C. The nepheline growth at the peak temperature, as determined by XRD, was between 35 - 75 wt.% /hour. A maximum nepheline growth rate of ~ 0.1 mm/hour at 700 °C was measured for the DWPF composition using optical microscopy. This research establishes a viable alternative to more traditional techniques for evaluating nepheline crystallization in large numbers of glasses, which are prohibitively time consuming or otherwise impractical. The ultimate objective is to combine the nucleation and growth information obtained from DSC, like that presented in this report, with computer simulations of glass cooling within the canister to accurately predict nepheline crystallization in HLW during processing through WTP.

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Vitrification of Hanford Tank 241-AP-101 Waste and Simulant

Hanford tank 241-AP-101 (referred to herein as AP-101) is the second Hanford radioactive tank waste planned to be processed and vitrified. A simulant version of AP 101 waste was formulated from the best-basis inventory (BBI) for the Hanford Tank 241-AP-101 liquid with an assumed target dilution of the waste from the BBI sodium molarity of 8.61 M to the desired 5.5 M Na. After the addition of glass-forming chemicals (GFCs), the simulant melter feed was processed in a non-radioactive, continuous laboratory-scale melter (CLSM) system. The AP-101 simulant melter feed was charged into the CLSM for 6.11 h of processing, which produced 6.55 kg of glass, for an average glass production rate of 2275 kg m 2 d -1 . Since there were no processing issues with the AP-101 simulant melter feed, AP-101 melter feed made with actual waste was then processed in a CLSM system built into a contamination area in a radioactive environment. The melting behavior characteristics appeared similar for both the simulant and waste melter feeds. The AP-101 waste melter feed was charged into the CLSM for 12.14 h of processing, which produced 8.75 kg of glass, for an average glass production rate of 1530 kg m 2 d -1 . During the AP-101 waste melter feed charging, the pump used to move the feed reached a maximum and it is believed that if the pump had a greater capacity, a greater average glass production rate could have been achieved. A constituent of interest present in low quantities in the AP-101 waste is 99 Tc or its non-radioactive surrogate, Re, added to the AP-101 simulant. Analysis for the quantities of 99 Tc and Re in the AP-101 glass product resulted in an average single-pass retention from the melter feed during relative chemical steady state of 55 ± 2 % for 99 Tc and 45 ± 2 % for Re. Compared to the processing of other melter feeds, the retention of 99 Tc in the AP-101 glass was greater than in both AP-107 and AP-105 glass, while the retention of Re in the AP-101 was less than in the AP-107 glass, but greater than in the AP-105 glass. A spike of I was added into the AP-101 melter feed that could be detected above the analysis detection limits. However, the iodine was only detectable above the ~6 ppm limit in one glass pour: the pour immediately following the burn off of the cold cap, where the I level reached ~30 ppm. This event was significant because the glass was poured immediately after burn off and thus it is presumed that the iodine had yet to volatilize from the glass melt while idling. It is recommended to perform future tests with I spikes at greater levels so that it can be detected in additional glass pours to determine if the expected 50 % retention of I used in the Kim et al. glass models can be confirmed. Offgas liquid samples were analyzed for acetonitrile, which was present at greater concentrations in CLSM liquids than in other scaled melter systems. This result was expected based on unique conditions with the CLSM system including a small plenum space leading to low residence time for offgas and the rapidity of offgas cooling upon exiting the CLSM vessel due to the location and environment. About 90 % of the total acetonitrile captured during both the AP-101 simulant and waste CLSM runs was found in the offgas condensate and demister liquids, thus it is recommended that only those liquids be sent for analysis if future testing to study the presence of acetonitrile in offgas products is desired.

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Arvia Organics Destruction: An Innovative Treatment Approach for Dioxins, Furans and Other Problematic (Orphan) Organic Wastes - 16262

The safe and efficient disposal of radionuclide-contaminated organic wastes is a key component in the lifecycle of a nuclear power plant. Technologies are available for the processing of some of these wastes although there are a number of waste streams for which the disposition path is not clear or does not exist at all. For example, waste streams such as dioxin, furan and PCB-laden oils and aqueous organic matrices pose an issue for the US Dept. of Energy complex. These wastes present a challenge as they cannot be readily treated using existing treatment technologies because of legislative and material constraints such as gaseous phase radioactivity discharge limits, waste elemental content, by-product generation and waste resistance to thermal decomposition. Alternate methods of organics destruction can circumvent some of these issues by using different mechanisms of organics oxidation. One such innovative method, known as the Arvia organics destruction process, combines adsorption with electrochemical combustion and offers such an alternative. This technology, based on a patented material, Nyex{sup TM} that is able to act as both an adsorbent and as an electrode, allows the electrochemical combustion of adsorbed organics to occur on the material's surface. The result is the destruction of organics whilst the bulk of inorganic and radionuclide components remain in the aqueous phase. These components can subsequently be treated using existing waste processing technologies. This paper outlines recent work conducted on demonstrating the Arvia process against the destruction of dioxin, furan and PCB-laden wastes that demonstrates that the technology successfully removes these target species to below discharge standards in the aqueous phase. Based on these successful trials a test unit has been designed to demonstrate the process on actual radioactive wastes at the DOE Oak Ridge facility. During commissioning at Arvia's facility the unit achieved a destruction rate of over 1.5 ml of 100% organic waste per hour. The unit is currently at a partner laboratory facility based at Oak Ridge National Laboratories and has recently completed a further suite of simulant trials which are reported here. The above work showcases the development, testing and initial application of an innovative technology, first presented to the industry at the Waste Management 2012, as another option for the treatment of challenging wastes with expensive, complex or no current disposal routes. Over the years the technology has gone through a number of proving stages culminating in the testing on real orphan wastes based at the Oak Ridge Facility. The Arvia Organics Destruction Cell technology adds to the portfolio of available treatment technologies and will help facilitate the decommissioning and waste disposal responsibilities of the nuclear industry. (authors)

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Enhanced prediction of Cs removal by CST from Hanford tank waste with K accountability

Abstract The treatment of Hanford tank waste is one of the most challenging environmental cleanup activities to date. To expedite the processing of liquid waste stored in underground tanks in Washington State it is necessary to remove the significant dose contributor, 137 Cs. Crystalline silicotitanate ion exchanger is currently used to remove 137 Cs from the aqueous phase of Hanford tank wastes in preparation for vitrification at the Waste Treatment and Immobilization Plant (WTP). Improving the understanding of potassium impacts on ion exchange behavior of Cs will help in the operation of a critical component of one of the most complex treatment processes in the world today. Optimization of this process can result in significant cost savings and less waste production. Toward this effort, a series of batch contact tests varied in potassium concentration were conducted to look at the impact of potassium concentration on Cs distribution. Experimental distribution ratios ( K d ) were compared to the distribution ratios predicted using the ZAM model. A significant underprediction of Cs capacity in the presence of potassium was seen with the existing model. A revision of the equilibrium constants was determined and provided a statistically better fit for determining the Cs K d values in tank waste matrices.

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Treatment of Uranium Catalyst Waste for High Volume Reduction of the Final Waste to be Disposed - 19023

A large volume of problematic radioactive waste of spent uranium catalyst, approximately 7,000 drums was generated by a chemical company in Korea. The spent uranium catalyst has since been stored onsite awaiting a management strategy that is to minimize the volume/cost requirements for final disposal of the uranium-bearing waste. This work developed for the first time a process to greatly reduce the waste volume to be disposed while the secondary wastes generated from the process being released to the environment are controlled to meet the discharge criteria in Korea. (authors)

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