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DOE OSTI2015

Glass Property Models, Constraints, and Formulation Approaches for Vitrification of High-Level Nuclear Wastes at the US Hanford Site

The legacy nuclear wastes stored in underground tanks at the US Department of Energy’s Hanford site is planned to be separated into high-level waste and low-activity waste fractions and vitrified separately. Formulating optimized glass compositions that maximize the waste loading in glass is critical for successful and economical treatment and immobilization of nuclear wastes. Glass property-composition models have been developed and applied to formulate glass compositions for various objectives for the past several decades. The property models with associated uncertainties and combined with composition and property constraints have been used to develop preliminary glass formulation algorithms designed for vitrification process control and waste form qualification at the planned waste vitrification plant. This paper provides an overview of current status of glass property-composition models, constraints applicable to Hanford waste vitrification, and glass formulation approaches that have been developed for vitrification of hazardous and highly radioactive wastes stored at the Hanford site.

composition constraint↗

DOE OSTI2020

Computationally-Aided Design of a Small-Scale Radioactive Waste Glass Melter

To provide mission support to the Hanford Waste Immobilization and Treatment Plant for the vitrification of legacy nuclear tank waste, a reduced-scale vitrification pilot system is being designed to process simulated and actual radioactive tank wastes. The tank waste will be separated into high-level waste (HLW) and low-activity waste (LAW) fractions, where the majority by mass (~90%) is LAW and by activity (>95%) is HLW. The first tank waste to be processed at the WTP will be LAW. Since Hanford tank waste and the resultant melter feed compositions are known to vary widely, pilot-scale operations are essential to identify potential problems, provide needed data, confirm assumptions, and determine impacts to full-scale melters and off-gas systems from these different feeds. In addition, the reduced-scale melter system must be capable of quickly providing results to operations without the typically high costs of radioactive operations and provide an engineering platform in close proximity to local operations staff. The objective is to produce similar process conditions to those encountered in the full-scale LAW melters and minimize the volume of radioactive waste necessary for the evaluations. To this end, melter design features are being explored to increase production without significantly increasing surface area, melter glass volumes or compromising data quality. In support of these objectives, a set of computational fluid dynamic models have been developed to provide insight into the flow patterns within these small melters and evaluate design features to increase production without significantly increasing glass inventory. Computed velocities were evaluated at the interface between the cold cap and the molten glass pool and within the bulk glass pool. A method was developed to correlate results to an estimated melt rate and down-select design features that produce the highest gains.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2020

Computationally-Aided Design of a Small-Scale Radioactive Waste Glass Melter

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2013

Waste Characterization Using Gamma Ray Spectrometry with Automated Efficiency Optimization - 13404

Gamma ray spectrometry using High Purity Germanium (HPGe) detectors is commonly employed in assaying radioactive waste streams from a variety of sources: nuclear power plants, Department of Energy (DOE) laboratories, medical facilities, decontamination and decommissioning activities etc. The radioactive material is typically packaged in boxes or drums (for e.g. B-25 boxes or 208 liter drums) and assayed to identify and quantify radionuclides. Depending on the origin of the waste stream, the radionuclides could be special nuclear materials (SNM), fission products, or activation products. Efficiency calibration of the measurement geometry is a critical step in the achieving accurate quantification of radionuclide content. Due to the large size of the waste items, it is impractical and expensive to manufacture gamma ray standard sources for performing a measurement based calibration. For well over a decade, mathematical efficiency methods such as those in Canberra's In Situ Object Counting System (ISOCS) have been successfully employed in the efficiency calibration of gamma based waste assay systems. In the traditional ISOCS based calibrations, the user provides input data such as the dimensions of the waste item, the average density and fill height of the matrix, and matrix composition. As in measurement based calibrations, the user typically defines a homogeneous matrix with a uniform distribution of radioactivity. Actual waste containers can be quite nonuniform, however. Such simplifying assumptions in the efficiency calibration could lead to a large Total Measurement Uncertainty (TMU), thus limiting the amount of waste that can be disposed of as intermediate or low activity level waste. To improve the accuracy of radionuclide quantification, and reduce the TMU, Canberra has developed the capability to optimize the efficiency calibration using the ISOCS method. The optimization is based on benchmarking the efficiency shape and magnitude to the data available in the analyzed gamma ray spectra. Data from measurements of a given item in multiple counting geometries are among the powerful benchmarks that could be used in the optimization. Also, while assaying a waste stream with fission products and activation products emitting gamma lines of multiple energies, optimizing the efficiency on the basis of line activity consistency is very effective. In the present paper, the ISOCS- based optimization methodology is applied to measurement scenarios involving multiple counting geometries, and multi-gamma-line radionuclides. Results will be presented along with accuracy and precision estimates for each measurement. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2013

Recent Developments in the Management of Cameco Corporation's Fuel Services Division Waste - 13144

Cameco Corporation is a world leader in uranium production. Headquartered in Saskatoon, Saskatchewan our operations provide 16% of the world uranium mine production and we have approximately 435 million pounds of proven and probable uranium reserves. Cameco mining operations are located in Saskatchewan, Wyoming, Nebraska and Kazakhstan. Cameco is also a major supplier of uranium processing services required to produce fuel for the generation of clean energy. These operations are based in Blind River, Cobourg and Port Hope, Ontario and are collectively referred to as the Fuel Services Division. The Fuel Services Division produces uranium trioxide from uranium ore concentrate at the Blind River Refinery. Cameco produces uranium hexafluoride and uranium dioxide at the Port Hope Conversion Facility. Cameco operates a fuel manufacturing facility in Port Hope, Ontario and a metal fabrication facility located in Cobourg, Ontario. The company manufactures fuel bundles utilized in the Candu reactors. Cameco's Fuel Services Division produces several types of low-level radioactively contaminated wastes. Internal processing capabilities at both the Blind River Refinery and Port Hope Conversion Facility are extensive and allow for the recycling of several types of waste. Notwithstanding these capabilities there are certain wastes that are not amenable to the internal processing capabilities and must be disposed of appropriately. Disposal options for low-level radioactively contaminated wastes in Canada are limited primarily due to cost considerations. In recent years, Cameco has started to ship marginally contaminated wastes (<500 ppm uranium) to the United States for disposal in an appropriate landfill. The landfill is owned by US Ecology Incorporated and is located near Grand View, Idaho 70 miles southeast of Boise in the Owyhee Desert. The facility treats and disposes hazardous waste, non-hazardous industrial waste and low-activity radioactive material. The site's arid climate, deep groundwater and favourable geology help ensure permanent waste isolation. Combined with a state of the art multi-layer landfill liner system, the Grand View facility represents an ideal choice to minimize environmental liability. Marginally contaminated wastes from operations within the Fuel Services Division are typically loaded into PacTec IP-2 rated Intermediary Bulk Containers and then transported by road to a nearby rail siding. The Intermediary Bulk Containers are then loaded in US Ecology owned gondola rail-cars. The gondolas are then transported via Canadian Pacific and Union Pacific railroads to the US Ecology Rail Transfer facility located in Mayfield, Idaho. The Intermediary Bulk Containers are unloaded into trucks for transport to the disposal facility located approximately 32 miles away. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

Vitrification of Sellafield Wastes with High Metallic Magnesium Alloy Content - 17495

The Sellafield nuclear complex, located in northwest England, is the largest such site in the United Kingdom. Sellafield Limited, which is the operating contractor for the Nuclear Decommissioning Authority, is evaluating thermal treatment technologies for a potential future facility to process a variety of Intermediate Level Waste sludges and slurries accumulated from historical site operations. In support of that effort, Atkins and the Vitreous State Laboratory (VSL) of The Catholic University of America conducted studies to evaluate the feasibility and implications of installing a joule-heated ceramic melter (JHCM) vitrification system to process a variety of sludge and slurry radioactive wastes at the Sellafield site. These sludges have the potential to contain a significant proportion of reactive metals in localized regions. The metallic fraction derives from unreacted Magnox spent fuel cladding. Thus, an important objective of this work was to investigate the performance of JHCM technology for the treatment of streams with localized regions of high reactive metal contents. The present work was focused on Site Ion Exchange Effluent Plant (SIXEP) Magnox sludges, investigated previously. Suitable waste simulants were developed for the SIXEP Magnox sludge waste stream, which included surrogates for all important radionuclides. High waste-loading, fully compliant glass formulations were designed using a combination of previous experience, extensive waste glass data bases, and small-scale crucible melts. Glasses from the test melts were characterized for properties that are important for processability and product quality. The most viable compositions were then selected for small-scale melter trials to collect information on processing rates and behavior, off-gas and secondary waste compositions, and overall system mass balance. These data were used to support the technology evaluation, which considered such factors as minimization of inactive additions, waste volume reduction, retention of radionuclides, minimization of secondary wastes, and overall process efficiency and effectiveness. The data were also used to develop sizing estimates for the full-scale facility and to estimate treatment durations under various treatment scenarios. A series of tests was then performed with partial replacement of the magnesium hydroxide with a metallic magnesium alloy that is close in composition to Magnox alloy. A commercially available magnesium alloy, Elecktron AZ31B, which has a nominal magnesium content of 95.7%, was used for this purpose. The testing was performed on a DuraMelter 10 (DM10) JHCM system to characterize the melter off-gas, glass product, and the effects of magnesium alloy particle size on feed processing. Data on processing rate and any processing issues, product glass composition, off-gas composition including hydrogen, and system operating parameters, including flows, temperatures, pressures, etc., were collected and analyzed to evaluate the effects of magnesium alloy on the vitrification process. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2018

Development of a Methodology to Assess the Impact of Non-Conventional Waste Streams Following a Radiological Incident - 18271

The Environmental Protection Agency's (EPA's) Waste Estimation Support Tool (WEST) is a Geographic Information System (GIS)-based decision support tool for estimating waste generated from remediation and cleanup activities following a radiological incident. WEST uses the Federal Emergency Management Agency's (FEMA's) Hazus-MH software's infrastructure information combined with image classification capabilities to estimate the quantity of waste and debris that may result from a radiological incident. WEST can be used to create customized decontamination strategies that allow end-users to evaluate different means of decontamination and how those strategies may impact the resulting waste stream in each geographical locality. This capability can potentially reduce the response/recovery time frame and cost. One of the key observations from past applications of WEST is that planning scenarios are site-specific, especially when dealing with chemical, biological, radiological, or nuclear (CBRN) waste management issues. Furthermore, site-specific issues are not limited to resource limitations or response bottlenecks but should also account for differences in an area's geography and urban footprint, a phenomenon observed during the Fukushima response. For instance, a recent WEST case study found that the amount of waste generated per unit of affected area, for a given decontamination approach, is largely a function of the building surface area in the affected area (relative to the ground surfaces). However, differences in the geometry and density of contaminated objects in the landscape may impact the amount and characteristics of the waste stream. Although buildings typically constitute a large portion of the surface area in urban environments, other objects such as vehicles and trees can be just as prominent. For example, seasonal changes in the amount of foliage present can alter the non-structural surface area calculated for an area like New York City by 20-30%. Very little is known about the impact that these non-conventional waste streams have on overall remediation strategies, or to what extent differences in characteristics of urban versus rural environments, traffic activities, or seasonal variations may have on these factors. This paper will present the proposed methodology to incorporate into WEST the impacts that niche, nonconventional waste streams might have from rural and urban environments following a nuclear power plant incident or detonation of a radiological dispersal device or improvised nuclear device. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2018

Basis of Knowledge for Oxidizing Chemicals in TRU Waste - 18654

A new Waste Isolation Pilot Plant (WIPP) document for implementing new requirements in the WIPP Waste Acceptance Criteria (WAC) titled Basis of Knowledge for Evaluating Oxidizing Chemicals in the TRU waste was issued in July, 2017 and has now been implemented on several waste streams. The February 14, 2014, airborne radiation release in Room 7 of the WIPP underground disposal unit 7 was caused by an exothermic chemical reaction of nitrate oxidizing chemicals and organic materials in a single 55-gallon container. That event exposed hazards not previously evaluated in the WIPP safety basis. A new WIPP Documented Safety Analysis (DSA), Revision 5b, was developed that addressed potential hazards related to the airborne radiation release and evaluated and addressed new fire hazard scenarios. The revised hazards analysis within the DSA specifically included potential chemical exothermic reactions in waste containers and propagating fires. Compliance with WAC was credited as initial conditions to the hazards analysis, and Chapter 18 was added making the WIPP WAC compliance program a safety management program within the DSA. The WIPP DSA, Revision 5b, Chapter 18, and the WIPP WAC require the use of this document when performing the enhanced acceptable knowledge (AK) process on transuranic (TRU) waste determined to contain oxidizing chemicals. The WIPP WAC was revised to address the conditions and requirements of DSA Revision 5b. The use of this Basis of Knowledge (BoK) by WIPP Certified Programs is specified in the WIPP WAC, Appendix H, Enhanced Acceptable Knowledge. This document establishes criteria that must be used by the WIPP Certified Programs to evaluate TRU waste containing one or more oxidizing chemicals to determine acceptability at the WIPP as-is and to identify when additional evaluation or treatment is required. This BoK also includes options and requirements to be used by TRU waste sites when further evaluation, testing, and/or treatment are required. The U.S. Department of Energy (DOE) Carlsbad Field Office (CBFO) will review requests, plans, and technical justifications and provide written decisions prior to implementation by the TRU waste sites. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2019

Characterization of Tc-99 Speciation and Chemical Composition of 241-AN-102 Tank Waste Supernatant - 19201

Technetium (Tc) in its isotopic form {sup 99}Tc is generated in high yields from the fission of {sup 235}U and {sup 239}Pu. A significant inventory of {sup 99}Tc remains at the U.S. Department of Energy's Hanford Site, and it is one of the most difficult contaminants to address. In the strongly alkaline solutions that are typical of Hanford tank waste, Tc exists as both pertechnetate (TcO{sub 4}{sup -} in oxidation state VII) and in reduced forms (oxidation state <VII) collectively known as non-pertechnetate species. Pertechnetate is a well-characterized anionic Tc species that can be removed from low-activity waste by anion exchange or other methods. There is no definitive information on the origin or a comprehensive description of the non-pertechnetate species in Hanford tanks. Designing strategies for effective management of Tc, including separation and immobilization, requires an understanding of the molecular structure of the non-pertechnetate species and their identification in actual tank waste samples. Developing this advanced understanding would facilitate development of new technologies for treating dissimilar Tc species. Toward this objective, a spectroscopic library of the Tc(I) [fac-Tc(CO){sub 3}]{sup +} and Tc(IV, VI, VII) compounds was generated using a range of techniques and applied to the characterization of actual 241-AN-102 tank waste supernatant. A sample of the 241-AN-102 tank waste (collected in 2013) was processed to remove {sup 137}Cs by spherical resorcinol-formaldehyde (sRF) ion exchange resin after the sodium concentration was adjusted to 5.6 M. The Cs-loaded sRF column was eluted with 0.45 M HNO{sub 3}. As-received AN-102, Cs-depleted AN-102 effluent, and sRF eluate fractions were comprehensively characterized for chemical composition and speciation of Tc using {sup 99}Tc nuclear magnetic resonance (NMR) spectroscopy and X-ray absorption spectroscopy. The results of these analyses demonstrated for the first time that non-pertechnetate Tc present in the 241-AN-102 tank waste is composed of several low-valent Tc species, including the Tc(I) [fac-Tc(CO){sub 3}L{sub 3}]{sup +}, Tc(IV), and possibly Tc(VI) compounds. This is the second experimental observation of the [fac-Tc(CO){sub 3}L{sub 3}]{sup +} species in the Hanford tank waste that highlights their importance for waste processing. In addition, the chemical composition of the 241-AN-102 tank waste was characterized with the focus on small chelators. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2019

Development of Waste Inventory for Probabilistic Performance Assessment Modeling - 19361

Estimation of radioactive waste inventory is a key initial step in the development of an environmental fate and transport model for performance assessment (PA) of radioactive waste sites. PA model output is often sensitive to parameters that represent inventory as a source term, yet estimation is a difficult problem. Inventory estimation for a probabilistic performance assessment (PPA) is related to, but distinct from, that performed for waste excavation, characterization, and offsite disposal. The latter appropriately focuses on obtaining the most detailed information possible on specific hazards and locations; however, these questions cannot be fully answered without intrusive site characterization efforts. This type of characterization is often not feasible when assembling information regarding a source term in a PPA model. Development of waste inventory for a PPA is one piece of a larger hierarchical model. The physical process model represented by Goldsim computer code represents one level of the hierarchy, and the second level of the hierarchy involves estimation of inputs such as inventory, transport, and risk parameters. The estimation of inputs is not a straightforward problem. The parameters of a computer model often do not have an equivalence to the physical quantities they describe, and estimation of the physical quantities themselves can be difficult due to limitations of available information. Inventory information for the West Valley Site in upstate New York is based on data from historical shipping and disposal forms. In the 1990's, the West Valley Demonstration Project (WVDP) compiled information from existing databases and historical waste records into one complete modern database. A conceptual model was then applied and used with recorded information to characterize the waste mass and radionuclide composition for each waste burial. Information was ultimately aggregated over all waste records to estimate the mass of each radionuclide within each trench, hole, or caisson. Three decades later, we revisit the master database, make updates with newly-identified information, and incorporate the conceptual model previously used for inventory estimation in the hierarchical modeling framework. We characterize uncertainty using probability distributions for waste mass and radionuclide composition at the individual shipment or individual burial level. Uncertainty incorporated at the lowest level of the hierarchy is ultimately carried through to uncertainty in PPA model endpoints. Sensitivity analysis of early model results is then applied to identify influential inputs and guide subsequent efforts towards parameters that drive risk. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2018

D and D of a Large Waste Treatment Plant in Germany - 18458

With its 40 years of nuclear experience, GNS, the world leading supplier of casks for spent fuel, ILW and HLW, also offers services for conditioning and management of all types of radioactive waste, especially for the German NPP. Since 1985, GNS treats and packages ILW and LLW from the operations and phase-out of German NPP in its Duisburg plant. To enable its permanent safe final disposal, solid wastes are compacted and dried before being shipped to an interim storage facility or - later - the Konrad repository. The total premises cover 20.000 m{sup 2} (215.000 sq ft), of which 9000 m{sup 2} (97.000 sq ft) is controlled area. The Duisburg Plant has three large industrial warehouses (each 25 m x 136 m or 82 ft x 446 ft) with a total capacity of 10.200 m{sup 2} or 110.000 sqft. For the treatment of open contaminated waste, six enclosures were built. These enclosures have a special ventilation system including a specific filtration plant. In one enclosure a high compaction plant was used, in another enclosures a drying facility for drums was in operation. During peak periods, more than 900 tonne of waste were treated annually. One enclosure was mainly used for maintenance, restoration and extension of mobile facilities, which were normally used for campaigns in NPP. Due to the decreasing amount of operational waste during the nuclear phase out in Germany, GNS has decided to close the Duisburg Plant. In the first half-year of 2017, the last waste packages were treated but around 80 20'-containers are still located on site. The D and D started in January 2017. The presentation will show the first experience of decommission processes of the first three enclosures. The following steps including the techniques used will be explained in detail: - Dismantling and removal of the high compaction facility; - Dismantling and removal of the drying facility; - Cleaning and wipe tests of the empty enclosures; - Removal of the stainless steel plates of the floor; - Investigations of the lower floor parts for contamination; - Removal of contaminated underground drainpipes; - Decontamination of the concrete floor; - Dismantling of the ventilation system in three enclosures; - Dismantling of the steel constructions. An important aim during the dismantling is to reduce the amount of nuclear waste and maximize the fraction material for free release. Next to the first three enclosures, in one of the enclosure IV - VI, a 150 tons scrap shear had been installed before the construction of the enclosure building. The scrap shear has to be dismantled and handled by heavy duty lifting equipment. Only afterwards is it possible to start the decommissioning of the last three enclosures. Our goal is to transfer this very large controlled area and the whole plant to the site owner, the Duisburg Port company, after 2 years and 5 months and doing so in a way so that the industrial warehouses can be used again for all possible types of industries. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2013

Emission of volatile sulfur compounds during composting of municipal solid waste (MSW)

Highlights: ► We compare the volatile sulfur compounds (VSCs) emissions during three types of municipal solid wastes (MSWs) composting. ► The VSCs released from the kitchen waste composting was significantly higher than that from 15–80 mm fraction of MSW. ► Among the five VSCs, H{sub 2}S was the most abundant compound with 39.0–43.0% of total VSCs released. ► Addition of 20% cornstalks could significantly reduce the VSCs emissions during kitchen waste composting. - Abstract: Volatile sulfur compounds (VSCs) are the main source for malodor from composting plants. In this study, the VSCs generated from composting of 15–80 mm municipal solid waste (T0), kitchen waste (T1) and kitchen waste mixed dry cornstalks (T2) were measured in 60 L reactors with forced aeration for a period of 30 days. The VSCs detected in all treatments were hydrogen sulfide (H{sub 2}S), methyl mercaptan (MM), dimethyl sulfide (DMS), carbon bisulfide (CS{sub 2}) and dimethyl disulfide (DMDS). Over 90% of the VSCs emissions occurred during the first 15 days, and reached their peak values at days 4–7. The emission profiles of five VSCs species were significantly correlated with internal materials temperature and outlet O{sub 2} concentration (p < 0.05). Total emissions of the VSCs were 216.1, 379.3 and 126.0 mg kg{sup −1} (dry matter) for T0, T1 and T2, respectively. Among the five VSCs, H{sub 2}S was the most abundant compound with 39.0–43.0% of total VSCs released. Composting of kitchen waste from separate collection posed a negative influence on the VSC and leachate production because of its high moisture content. An addition of dry cornstalks at a mixing ratio of 4:1 (wet weight) could significantly reduce the VSCs emissions and avoid leachate. Compared to pure kitchen waste, VSCs were reduced 66.8%.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2013

Behavior of an MBT waste in monotonic triaxial shear tests

Highlights: ► We studied the stress–strain–strength characteristics of an MBT waste. ► Rate of mobilization of strength with strain depends on initial density. ► Image analysis technique was used to determine whole-specimen displacement fields. ► Initial mode of deformation of a loose specimen is one-dimensional compression. ► Reinforcing elements enhance the resistance to lateral and volumetric deformation. - Abstract: Legislation in some parts of the world now requires municipal solid waste (MSW) to be processed prior to landfilling to reduce its biodegradability and hence its polluting potential through leachate and fugitive emission of greenhouse gases. This pre-processing may be achieved through what is generically termed mechanical–biological-treatment (MBT). One of the major concerns relating to MBT wastes is that the strength of the material may be less than for raw MSW, owing to the removal of sheet, stick and string-like reinforcing elements during processing. Also, the gradual increase in mobilized strength over strains of 30% or so commonly associated with unprocessed municipal solid waste may not occur with treated wastes. This paper describes a series of triaxial tests carried out to investigate the stress–strain–strength characteristics of an MBT waste, using a novel digital image analysis technique for the determination of detailed displacement fields over the whole specimen. New insights gained into the mechanical behavior of MBT waste include the effect of density on the stress–strain response, the initial 1-D compression of lightly consolidated specimens, and the likely reinforcing effect of small sheet like particles remaining in the waste.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2017

Assessing the costs of municipal solid waste treatment technologies in developing Asian countries

Highlights: • Municipal waste treatment costs are assessed for the first time in developing Asia. • A database with primary and secondary data from existing plants was prepared. • Capital and operational expenditures patterns were analyzed and discussed. • Incineration identified by large margin as the costliest approach for treating waste. • Context specific circumstances are key to explain deviations from the norm. - Abstract: The management of municipal solid waste (MSW) is one of the main costs incurred by local authorities in developing countries. According to some estimates, these costs can account for up to 50% of city government budgets. It is therefore of importance that policymakers, urban planners and practitioners have an adequate understanding of what these costs consist of, from collection to final waste disposal. This article focuses on a specific stage of the MSW value chain, the treatment of waste, and it aims to identify cost patterns associated with the implementation and operation of waste treatment approaches in developing Asian countries. An analysis of the capital (CAPEX) and operational expenditures (OPEX) of a number of facilities located in countries of the region was conducted based on a database gathering nearly 100 projects and which served as basis for assessing four technology categories: composting, anaerobic digestion (AD), thermal treatment, and the production of refuse-derived fuel (RDF). Among these, it was found that the least costly to invest, as a function of the capacity to process waste, are composting facilities, with an average CAPEX per ton of 21,493 USD{sub 2015}/ton. Conversely, at the upper end featured incineration plants, with an average CAPEX of 81,880 USD{sub 2015}/ton, with this treatment approach ranking by and large as the most capital intensive of the four categories assessed. OPEX figures of the plants, normalized and analyzed in the form of OPEX/ton, were also found to be higher for incineration than for biological treatment methods, although on this component differences amongst the technology groups were less pronounced than those observed for CAPEX. While the results indicated the existence of distinct cost implications for available treatment approaches in the developing Asian context, the analysis also underscored the importance of understanding the local context as a means to properly identify the cost structure of each specific plant. Moreover, even though CAPEX and OPEX figures are important elements to assess the costs of a waste treatment system, these should not be considered on a standalone basis for decision making purposes. In complement to this internal cost dimension, the broader impacts – to the economy, society and the environment – resulting from the adoption of a certain treatment approach should be properly understood and, ideally, measured and expressed in monetary terms.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2017

Waste Management Strategies for Production of Mo-99

Production of Mo-99 for medical isotope use is being investigated using dissolved low enriched uranium (LEU) fissioned using an accelerator driven process. With the production and separation of Mo-99, a low level waste stream will be generated. Since the production facility is a commercial endeavor, waste disposition paths normally available for federally generated radioactive waste may not be available. Disposal sites for commercially generated low level waste are available, and consideration to the waste acceptance criteria (WAC) of the disposal site should be integral in flowsheet development for the Mo-99 production. Pending implementation of the “Uranium Lease and Take-Back Program for Irradiation for Production of Molybdenum-99 for Medical Use” as directed by the American Medical Isotopes Production Act of 2012, there are limited options for disposing of the waste generated by the production of Mo-99 using an accelerator. The commission of a trade study to assist in the determination of the most favorable balance of production throughput and waste management should be undertaken. The use of a waste broker during initial operations of a facility has several benefits that can offset the cost associated with using a subcontractor. As the facility matures, the development of in-house capabilities can be expanded to incrementally reduce the dependence on a subcontractor.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2017

Nuclear Liquid Wastes Calcination: The High-Level French Experience - 17184

Calcination is an efficient process for volume reduction and stabilization of nuclear liquid wastes. It is often the first step of waste containment processes such as vitrification or grouting. The CEA (French Alternative Energies and Atomic Energy Commission) and AREVA have acquired a high-level experience in the field of calcination through more than forty years of R and D and industrial operation. During this period, a broad range of liquid wastes have been studied and treated. R and D work and associated engineering studies allowed defining the design and sizing of the equipment and the process control parameters to be applied for the calcination of all kinds of solutions. The calcination process has been qualified for the following effluents amongst others: - Liquid waste from UOx fuel reprocessing; - Liquid waste from UMo fuel reprocessing (high molybdenum and phosphorus contents); - Hanford HLW; - Liquid waste from D and D operation (sodium bearing waste). This paper presents several studies led by the Joint Vitrification Laboratory (L.C.V), a common research laboratory between CEA and AREVA in charge of qualifying new processes and matrices for waste containment. These studies allowed defining the calcination rules, in particular for very hard to process solutions such as effluents with high boron, molybdenum and sodium contents. The feedback from industrial operation of the calcination process with fission products and effluents from D and D operations, highly loaded with sodium, is also presented. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2017

Evaluating the Impacts of Reactive Gas (Carbon Dioxide and Oxygen) Ingress on Leaching from Cementitious Waste Forms - 17482

Cementitious materials are currently being used as the waste forms for low-activity waste disposal (Saltstone) at the Savannah River Site. In addition, cementitious materials have been selected as the waste form (Cast Stone) for secondary wastes and are being investigated for use for supplemental low-activity waste treatment (Cast Stone) at the Hanford Site. These materials are subjected to numerous aging and degradation phenomena in the disposal environment that potentially impact their long-term durability and performance. One important phenomenon is the ingress of reactive gases (i.e., carbon dioxide and oxygen) that can react with and change the chemistry and pore structure and thus mass transport properties of the waste forms. The reaction of carbon dioxide (i.e., carbonation) reduces the pH of the waste form pore solution impacting the solubility of constituents of potential concern (COPCs). In Cast Stone and Saltstone, the reaction of oxygen (i.e., oxidation) impacts the solubility of redox-sensitive COPCs, notably the technetium-99 (Tc-99) and chromium. The objective of this research is to experimentally characterize the carbonation and oxidation impacts on chemistry and leaching behavior in cementitious waste forms using experimental techniques including United States Environmental Protection Agency (U.S. EPA) Method 1313 and U.S. EPA Method 1315. Results of experimental characterization will be used to parameterize reactive mass transport models to project leaching under a range of field scenarios in future applications. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2016

Evaluation and Impacts of Mercury in the SRS Liquid Waste System - 16121

The Savannah River Site (SRS) Liquid Waste System (LWS) stores radioactive waste in 43 underground tanks. The radionuclides in the waste are removed through a series of separation processes and the low-level fraction is immobilized in a grout waste form while the high level fraction is disposed in a glass waste form. Mercury originated from decades of canyon processing (used as a catalyst for dissolving the aluminum cladding of reactor fuel) and is present throughout the LWS (∼60 metric tons). Mercury has long been a consideration in the LWS, from both hazard and processing perspectives. Mercury is removed from the LWS in many ways including: 1) the Tank Farm evaporator condensates; 2) at the Defense Waste Processing Facility (DWPF) where it is steam-stripped and removed during the feed preparation processes; 3) it is removed at the Effluent Treatment Plant (ETP); and it is immobilized in the low-level, grout waste form. An integrated, system-wide evaluation and assessment of mercury behavior in the LWS is being performed. The recently completed Phase I evaluation focused on mercury inventory and speciation in the LWS; mercury holdup and chemical processing behavior; mercury impact Identification, including worker safety and equipment degradation; and mercury removal and disposal options. The mercury removal system at the DWPF has been unable to remove mercury successfully due to equipment and chemistry challenges since the 2008 timeframe. Phase I evaluations indicate that approximately 43 percent of the mercury in waste sent to the DWPF is now returning to the Tank Farm in the recycle stream and is concentrating to higher than expected levels (∼500 mg/l) around the 2H evaporator system. The chemical behavior of mercury is further complicated by the formation of higher than expected quantities of organomercury species, specifically methyl mercury, now present in the DWPF recycle evaporator system (2H Tank farm evaporator), which is used to volume-reduce the recycle stream. Phase II activities are building on the Phase I mercury activities, including additional sampling and characterization activities, a re-assessment of overall system knowledge, ranking and prioritizing critical gaps/information in mercury behavior across the flowsheet, assessing impacts of mercury removal and disposal options, and documenting an action plan for overall mercury management in the LWS. Phase II activities include two System Engineering Evaluations to 1) 'Reestablish Mercury removal Capability within DWPF', and 2) 'Determine the Best Alternative Mercury Removal Location with the LWS'. This paper provides the results of the Phase I and II mercury evaluations for the SRS LWS. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗