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DOE OSTI2025

Processes in Salt Repositories for Radioactive Waste Disposal

This document summarizes the key processes (thermal, hydrological, mechanical, and chemical; THMC) impacting the features of a deep geological repository for radioactive waste in salt. Some processes are natural and on-going whether the repository is there or not, and other processes are driven by the perturbation associated with the repository. The features considered here include both engineered and natural components of the repository system. The engineered barrier system (EBS) in a salt repository is quite different from those implemented for a repository in clay or crystalline rocks, because it is comprised mostly of granular salt and salt-compatible cements, rather than bentonite. When compared to other rocks (i.e., silicates), salt has unique properties that make it an excellent potential host rock. Openings and fractures in salt creep closed readily. Salt has high thermal conductivity, which can reduce peak temperatures. Additionally, far away from the excavations the porosity of salt is unconnected, which leads to essentially zero advective or diffusive transport. The small amount of hypersaline brine occurring in salt minimizes microbial activity, reduces colloid-assisted transport, and eliminates in-package criticality (i.e., chloride is a neutron poison). At the end of the report, we present a brief outline for a potential salt repository, including considerations avoided in previous repository disposal concepts. We propose considering higher-temperature processes in future disposal concepts, rather than trying to minimize the thermal perturbation of the repository. Since hot salt is drier, a dry repository would limit corrosion, gas generation, and solute transport. Openings and fractures creep shut faster in hot salt. Therefore, higher temperatures could be seen as beneficial, rather than something to minimize, through increased spacing between waste packages (increasing repository costs).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Experimental study of the tritium distribution in the effluents resulting from the sodium hydrolysis

Within the framework of the dismantling of fast breeder reactors in France several processes are under investigation regarding sodium disposal. One of them, called ELA (radioactive sodium waste treatment process), is based on the implementation of the sodium-water reaction, in a controlled and progressive way, to remove residual sodium. This sodium contains impurities such as sodium hydride, sodium oxide and tritiated sodium hydride. The hydrolysis of these various chemical species leads to the production of a liquid effluent, mainly composed of an aqueous solution of sodium hydroxide, and a gaseous effluent, mainly composed of nitrogen (inert gas), hydrogen and steam. The tritium is distributed between these effluents, and, within the gaseous effluent, according to its forms HT and HTO (tritiated water). HTO being 10,000 times more radio-toxic than HT, a precise knowledge of the mechanisms governing the phase distribution of tritium is necessary. This paper presents the first experimental results from a parametric study on the tritium distribution between the various effluents generated during hydrolysis operations. A series of experiments have been performed in order to study the influence of water flow rate, argon flow rate, initial mass and specific activity of the hydrolyzed sodium sample. An important influence of the total tritium concentration in the hydrolyzed sample has been highlighted. As for the phenomena suspected to be responsible for the phase change of tritiated water, in the studied range of parameters, vaporization induced by the heat of reactions seems to be dominant over the evaporation induced by the inert gas flow rate.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Performance Evaluation of Decontamination Foam Containing Mesoporous silica NPs - 18453

It is necessary to secure decontamination technology that is easy to apply and minimizes the generation of secondary waste for large-scale equipment and large-area decontamination, which are difficult to be applied structurally when massive repairs or the decommissioning of a nuclear facility are required. In this respect, decontamination foam has a potentially wide application in the removal of contaminants from metallic walls, overhead surfaces, and elements of complex components and facilities. The decontamination efficiency can be enhanced by improving the contact time between the chemical decontamination agents and the contaminated surface through the introduction of stabilizers into the decontamination foam. In this study, the stability of a decontamination foam containing mesoporous silica NPs as a stabilizer and the decontamination performance of radioactive isotope contaminated specimens were evaluated. A process for waste liquid treatment after foam decontamination was also proposed. Decontamination performance tests using RI contaminated specimens were carried out in a decontamination foam prepared only with a solution containing only 1 wt% EM100 surfactant, and in a decontamination foam prepared with a solution containing 1 wt% EM100 surfactant and 1 wt% of various silica NPs. The stability of the decontamination foam was increased owing to the addition of various types of silica NPs. In particular, the liquid film thickness of the decontamination foam containing mesoporous silica NPs was the largest. As a result, the decontamination coefficient for RI contaminated specimens increased as the thickness of the foam liquid film increased owing to the longer contact time between the RI contaminated specimen and the chemical decontamination agent as the thickness of the foam liquid film increased. (authors)

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

Evaluating performance of high efficiency mist eliminators

Processing liquid wastes frequently generates off gas streams with high humidity and liquid aerosols. Droplet laden air streams can be produced from tank mixing or sparging and processes such as reforming or evaporative volume reduction. Unfortunately these wet air streams represent a genuine threat to HEPA filters. High efficiency mist eliminators (HEME) are one option for removal of liquid aerosols with high dissolved or suspended solids content. HEMEs have been used extensively in industrial applications, however they have not seen widespread use in the nuclear industry. Filtering efficiency data along with loading curves are not readily available for these units and data that exist are not easily translated to operational parameters in liquid waste treatment plants. A specialized test stand has been developed to evaluate the performance of HEME elements under use conditions of a US DOE facility. HEME elements were tested at three volumetric flow rates using aerosols produced from an iron-rich waste surrogate. The challenge aerosol included submicron particles produced from Laskin nozzles and super micron particles produced from a hollow cone spray nozzle. Test conditions included ambient temperature and relative humidities greater than 95%. Data collected during testing HEME elements from three different manufacturers included volumetric flow rate, differential temperature across the filter housing, downstream relative humidity, and differential pressure (dP) across the filter element. Filter challenge was discontinued at three intermediate dPs and the filter to allow determining filter efficiency using dioctyl phthalate and then with dry surrogate aerosols. Filtering efficiencies of the clean HEME, the clean HEME loaded with water, and the HEME at maximum dP were also collected using the two test aerosols. Results of the testing included differential pressure vs. time loading curves for the nine elements tested along with the mass of moisture and solid material on each element at final dP. Plots of overall filtering efficiencies for DOP (spherical aerosol) and dry surrogate (aspherical aerosols) at specified dPs were computed for each filter. Filtering efficiencies were determined as a function of particle size. Curves were also generated showing the most penetrating particle size as a function of dP. A preliminary set of tests was conducted to evaluate spray location, duration, pressure, and wash volume for in-place cleaning the interior surface (reducing dP) of the HEME element. A variety of nozzle designs were evaluated and test results demonstrated the potential to overload the HEME (saturate filter medium) resulting in very high dPs and extensive drain times. At least one combination of spray nozzle design, spray location on the surface of the element, and spray time/pressure was successful in achieving extension of operational life. (authors)

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

Strontium Speciation in Relevant Tank Waste Components Examined by Electrospray Ionization Mass Spectrometry

The identification of chemical species formed in complex nuclear waste is crucial for the development and employment of advanced separations technologies to remediate the Hanford site by processing tank waste. The current Tank Side Cesium Removal (TSCR) process deployed at Hanford utilizes crystalline silicotitanate (CST) ion exchange (IX) media to aid in the separation of low-activity waste for proper treatment and disposal. The inorganic IX media is highly selective for Cs but has been shown to also remove Sr from caustic simulants and small-scale IX processing of Hanford tank waste.(Fiskum, Rovira et al. 2019, Fiskum, Campbell et al. 2021, Westesen, Campbell et al. 2022) Quantitative Sr removal has not been observed in all tank waste supernates tested; thus, to better understand Sr removal and effectively predict processing behavior through TSCR, it is necessary to first investigate Sr speciation in tank waste. This work utilized electrospray ionization mass spectrometry (ESI-MS) to identify ionic Sr complexes that form in the presence of NO 3 –, NO 2 –, OH–, and Cl–. Although our results show that NO 3 –, NO 2 –, and OH– are competitive for Sr 2+ binding, previous data from IX studies indicate that [SrOH] + is not the dominant species of concern in tank waste processing schemes.(Fiskum, Campbell and Trang-Le 2020) Our results show that the [Sr(NO3)]+ species and the [Sr(NO2)] + species form in considerable abundances, which may affect the ability to separate Sr using CST in nuclear waste separation processes.

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

Overview of Upcoming Process Improvement Efforts at the Saltstone Facilities at the Savannah River Site - 20474

The Saltstone Facilities at the Savannah River Site (SRS) process and dispose of low activity salt solution waste generated by other liquid waste facilities. Today in the Liquid Waste System (LWS), approximately 125 million liters of salt waste remain to be processed. The Saltstone Production Facility (SPF) receives the decontaminated salt waste, which is mixed with dry feeds consisting of cement, fly ash and slag to produce the saltstone. The mixed slurry is pumped to the Saltstone Disposal Units (SDUs) for safe disposal in the Saltstone Disposal Facility (SDF). To support continuous Salt Waste Processing Facility (SWPF) operations at annual processing volumes many times higher than present, significant modifications are required to ensure that the SPF has the capacity to support the throughput of salt waste. These changes include several infrastructure upgrades that are being worked through the Enhanced Low Activity Waste Disposal (ELAWD) project. In addition to infrastructure upgrades, Savannah River Remediation (SRR), the Savannah River Site's Liquid Waste Contractor for the U.S. Department of Energy, has identified a flowsheet improvement opportunity of implementing a cement-free saltstone formulation to eliminate cement from the current three component formulation. This two-component formulation would simplify the procurement, transport, off-loading, and storage of the premix materials, and increase the storage capacity for the individual components. This flowsheet, when implemented, will reduce operational risk and improve throughput necessary to achieve disposal rates necessary for SWPF operations. Research has been completed to confirm the feasibility of this formulation, with the new cement-free flowsheet expected to be implemented in the next few years. Improvements have also been identified downstream of the SPF to increase SDU utilization. Specifically, a concentrated effort has taken place to raise the fill height in SDU 6 to the full disposal height, 13 meters, to maximize the usage of each SDU in support of the liquid waste disposal mission. This improvement has been made through additional thermal modeling and improved process understanding to create a SDU flammability model and control strategy more representative of facility operations. This work will allow for the completion of construction and operational readiness for SDU 7 while filling SDU 6 to ensure there is no lapse in available space once SWPF starts up. A summary of these improvements as they relate to increased throughput will be provided along with associated status. (authors)

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

Identification and Quantification of Technetium Species in Hanford Waste Tank AN-102

Technetium-99 (Tc) generated from the fission of 235U and 239Pu in high yields is one of the most difficult contaminants to be addressed at the U.S. Department of Energy Hanford Site. In strongly alkaline solutions typifying Hanford tank waste, Tc exists as pertechnetate (TcO4-) (oxidation state VII) as well as in reduced forms (oxidation state < VII) collectively known as non-pertechnetate species. Designing strategies for effective Tc management, including separation and immobilization, necessitates understanding the molecular structure of the non- pertechnetate species and their identification in the actual tank waste samples, which would facilitate development of new treatment technologies effective for dissimilar Tc species. Toward this objective, a spectroscopic library of the Tc(I) [fac-Tc(CO)3]+ and Tc(IV, VII) compounds was generated using a range of techniques and applied to the characterization of the actual tank waste supernatant collected from the tank 241-AN-102 at Hanford, WA. A sample of the 241-AN-102 tank waste supernatant was processed to adjust Na concentration to about 5.6 M and remove 137Cs by spherical resorcinol-formaldehyde (sRF) ion exchange resin. Cesium-loaded sRF column was eluted with 0.5 M HNO3. As received AN-102, Cs-depleted AN-102 effluent, and sRF eluate fractions were comprehensively characterized for chemical composition and speciation of Tc using 99Tc nuclear magnetic resonance spectroscopy and X-ray absorption spectroscopy. It was 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)3]+ and Tc(IV) compounds. This is the second experimental observation of the [fac-Tc(CO)3]+ species in the Hanford tank waste and the first demonstration of multiple forms of non-pertechnetate species existing simultaneously in the waste, cumulatively highlighting their importance for the waste processing.

Low Activity Waste (LAW), High Level Waste, nuclea↗
DOE OSTI2019

Idaho-Developed Technologies, Processes Offer Unique Solutions to Waste Management Challenges - 19281

INL Site cleanup contractor Fluor Idaho, LLC (Fluor Idaho) and its predecessors have developed a host of state-of-the-art technologies and specialized processes for nuclear waste handling, treatment, storage, and shipping. The cleanup program has also repurposed and upgraded previously idled Cold War-era buildings to treat and repackage both on- and off-site generated radioactive and hazardous wastes. And lastly, Fluor Idaho is using one-of-a-kind equipment to protect waste management personnel as they undertake their work to meet cleanup commitments that the Department of Energy (DOE) made to the state of Idaho. Such applications, although developed for Idaho's wastes, have the potential to benefit other facilities within the DOE Complex. Engineers and fabricators have teamed up to develop truly unique devices made of specialized metals, like stainless steel alloys, and more readily available materials such as Lexan. All waste types that Fluor Idaho manages daily - spent nuclear fuel, both high- and low-level wastes, and both remote-handled (RH) and CH-TRU - benefit from these innovations. Without question, the greatest number of specialized equipment, processes, and worker-protection technologies developed are for the retrieving, characterizing, repackaging, and shipping of TRU wastes, which makes sense considering the INL Site stores the most TRU waste in the DOE Complex. On the CH-TRU retrieval front, crews are using specially modified Gradall excavators to protect workers while they exhume Rocky Flats Plant wastes that were sent to Idaho in the 1950's and 1960's and buried in a shallow landfill. The excavators are supplied with extra Lexan windows to protect operators from the potential for flying debris during exhumation. On-board supplied air, pumped into the cab, and individual air supplies protect workers from inhalation hazards. Also on-board infrared cameras allow operators to detect heat generating from reactive metals at the dig face. At the Advanced Mixed Waste Treatment Facility (AMWTP), Brokk robotic arm operators use a 'super clam' claw, designed and built at the INL Site, on the remotely operated heavy equipment to size-reduce waste boxes, drums, concrete, and metal containing TRU constituents. The super clam is so powerful and dexterous that operators can crush a 55-gallon drum into the size of a basketball, but also delicately turn the pages of a magazine. Boxes too large to be sized at the AMWTP's box-lines are sent to the Accelerated Retrieval Project VII facility to be sized by Waste Management personnel. Such waste boxes, which sometimes contain 'flighty' plutonium-238 material, have forced the project to employ a heavily ventilated and filtered interior tent. Not all technologies developed for the Waste Management program are high-technology; in fact, one piece of equipment is so simple it's literally ingenious. Operators who inspect and repackage RH-TRU inside a hot cell at the Idaho Nuclear Technology and Engineering Center (INTEC) needed a backup plan in case the hot cell port couldn't be closed in the event of a power outage. Engineers and fabricators collaborated to develop a port cover actuator mechanism, which allows operators to open and close the cell's hatch cover with nothing more than a cordless drill. Also at INTEC, Maintenance Shop fabricators developed a Lexan container or 'pig' suitable for shielding neutron-emitting TRU waste sources. Lexan was chosen as the preferred shielded container material for this special type of waste because neutrons are best shielded with materials containing lots of hydrogen, such as plastics, which are composed of carbon and hydrogen molecules. Once approved for shipment, the high-activity waste will be loaded into specialty casks and shipped to WIPP for permanent disposal. Although all these technologies and processes were initially developed for resolving challenges at the INL Site, their applications could be far-reaching. Like the INL Site, many other facilities in the DOE Complex manage, treat, store, and ship TRU, low-and high-level wastes, and spent nuclear fuel. The technologies and processes developed in Idaho protect Waste Management personnel as they undertake their work, but these benefits could be utilized by other facilities and personnel doing similar work. (authors)

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

Process Control Approach to Implement High Waste Loading Glass Formulations for Hanford Low Activity Waste Vitrification - 18435

Acceptable glass formulations for vitrification of Hanford low activity waste (LAW) at The Hanford Tank Waste Treatment and Immobilization Plant (WTP) must meet a variety of product quality, processability, and waste loading requirements. The glass formulation development work underpinning the WTP baseline LAW glasses was captured in a vitrification process control approach developed by the Vitreous State Laboratory (VSL) and Atkins Energy Federal EPC, Inc. that permits the calculation of the glass formulation required for each particular batch of LAW delivered to the LAW vitrification facility. Since that time, we have developed a wide range of glass formulations that achieve considerably higher waste loadings and which therefore are more economically attractive in terms of reduced amounts of glass produced and decreased time for treatment. That work was done in support of the Department of Energy's Office of River Protection program that is examining options to optimize the LAW facility and LAW glass waste form. Implementation of these high waste loading formulations at the WTP requires the development of a new LAW vitrification process control algorithm, which is the subject of this paper. Waste loadings in LAW glasses are typically limited by either the alkali loading for LAW with low sulfate content or the SO{sub 3} loading for LAW with high sulfate content. In the former case, waste loadings are limited by refractory corrosion and glass corrosion on the vapor hydration test (VHT) and product consistency test (PCT), all of which tend to increase with higher alkali concentrations. In the latter case, waste loadings are limited by the formation of a molten sulfate phase in the melter, which has deleterious effects. The LAW vitrification process control strategy must mitigate these effects while optimizing waste loadings. Given the composition of any LAW stream, the new LAW vitrification process control algorithm permits the calculation of a suitable high waste loading glass composition and the required glass forming additives such that all processing, waste loading, and product quality requirements are met. This algorithm was developed from data for a large number of glasses that were formulated, and characterized with respect to a variety of properties affecting processability and product quality; many of these formulations were also subjected to melter testing to verify acceptable processing characteristics, processing rates, and the absence of secondary sulfate phase formation. This paper describes the development of the present algorithm, comparisons to the present lower waste loading baseline WTP process control approach, and plans for further development and testing. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2014

Yttrium recovery from primary and secondary sources: A review of main hydrometallurgical processes

Highlights: • Review of the main hydrometallurgical processes to recover yttrium. • Recovery of yttrium from primary sources. • Recovery of yttrium from e-waste and other types of waste. - Abstract: Yttrium is important rare earths (REs) used in numerous fields, mainly in the phosphor powders for low-energy lighting. The uses of these elements, especially for high-tech products are increased in recent years and combined with the scarcity of the resources and the environmental impact of the technologies to extract them from ores make the recycling waste, that contain Y and other RE, a priority. The present review summarized the main hydrometallurgical technologies to extract Y from ores, contaminated solutions, WEEE and generic wastes. Before to discuss the works about the treatment of wastes, the processes to retrieval Y from ores are discussed, since the processes are similar and derived from those already developed for the extraction from primary sources. Particular attention was given to the recovery of Y from WEEE because the recycle of them is important not only for economical point of view, considering its value, but also for environmental impact that this could be generated if not properly disposal.

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

Glass formulation and lab-scale testing of glasses designed for in-can Melter and in-container Vitrification of high-assay low-enriched uranium aqueous polishing Raffinate waste

Glasses were designed for processing a nuclear waste from aqueous polishing of high-assay low-enriched uranium using either In-Can Melter (ICM) or GeoMelt® In-Container Vitrification™ (ICV) technologies, which operate at temperatures of T p ≤ 1100 °C and T p ≤ 1450 °C, respectively. Due to the different operating conditions, the melt and glass properties were optimized differently for each technology. Each glass was designed to optimize for maximum waste loading while simultaneously satisfying processing (e.g., crystallization, viscosity, and conductivity) and product quality (e.g., durability, hazard characteristic, and crystal content) constraints. Here, the raffinate waste contains high nitric acid (4 M) and low total solids (9 g·L -1 ) concentrations. Feed preparation processes were tested to facilitate concentration and nitrate destruction/removal while controlling redox of the melter feed. A successful feed process including sugar addition and spray-drying was performed to generate an adequate melter feed for ICV processing.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

A case-study of landfill minimization and material recovery via waste co-gasification in a new waste management scheme

Highlights: • A new waste management scheme and the effects of co-gasification of MSW were assessed. • A co-gasification system was compared with other conventional systems. • The co-gasification system can produce slag and metal with high-quality. • The co-gasification system showed an economic advantage when bottom ash is landfilled. • The sensitive analyses indicate an economic advantage when the landfill cost is high. - Abstract: This study evaluates municipal solid waste co-gasification technology and a new solid waste management scheme, which can minimize final landfill amounts and maximize material recycled from waste. This new scheme is considered for a region where bottom ash and incombustibles are landfilled or not allowed to be recycled due to their toxic heavy metal concentration. Waste is processed with incombustible residues and an incineration bottom ash discharged from existent conventional incinerators, using a gasification and melting technology (the Direct Melting System). The inert materials, contained in municipal solid waste, incombustibles and bottom ash, are recycled as slag and metal in this process as well as energy recovery. Based on this new waste management scheme with a co-gasification system, a case study of municipal solid waste co-gasification was evaluated and compared with other technical solutions, such as conventional incineration, incineration with an ash melting facility under certain boundary conditions. From a technical point of view, co-gasification produced high quality slag with few harmful heavy metals, which was recycled completely without requiring any further post-treatment such as aging. As a consequence, the co-gasification system had an economical advantage over other systems because of its material recovery and minimization of the final landfill amount. Sensitivity analyses of landfill cost, power price and inert materials in waste were also conducted. The higher the landfill costs, the greater the advantage of the co-gasification system has. The co-gasification was beneficial for landfill cost in the range of 80 Euro per ton or more. Higher power prices led to lower operation cost in each case. The inert contents in processed waste had a significant influence on the operating cost. These results indicate that co-gasification of bottom ash and incombustibles with municipal solid waste contributes to minimizing the final landfill amount and has great possibilities maximizing material recovery and energy recovery from waste.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Integrity Assessment of Hanford 242-A Evaporator Facility E-A-1 Re-boiler Vessel via Fluorescein Liquid Tracer Test Methods - 17357

The 242-A Evaporator Facility is critical to the United States Department of Energy's (DOE) Hanford Site clean-up mission due to limited space within the existing Double-Shell Tanks (DSTs) used to store nuclear and chemical waste until future processing. The 242-A Evaporator Facility is currently the sole waste volume reduction process used at Hanford, and the E-A-1 Re-boiler is essential to the Evaporator process. Due to the potential for cross-contamination between the Re-boiler tubes through which waste passes and the tube sheets of uncontaminated steam, an assessment of the Re-boiler vessel's integrity is required for continuing operations of the facility. In order to meet these requirements, an independent integrity assessment was designed to use fluorescein liquid tracer (LT) to verify the leak-tightness of the Re-boiler vessel in accordance with Washington River Protection Solutions, LLC (WRPS) tightness test standards, as well as, the ASME Boiler and Pressure Vessel (B and PV) Code, Section III, Division 1, NB-6300. LT test methodologies were evaluated and determined to be appropriate for a vessel with restricted access, such as the E-A-1 Re-boiler, as it provided the needed sensitivity, reliability, and cost; as well as, exceeded the leak-tightness specifications required by both WRPS and ASME B and PV Code standards. The integrity test requires water laced with fluorescein be circulated through the waste processing loop of the 242-A Evaporator and samples of steam condensate be collected and analyzed for traces of fluorescein to determine leak-tightness. Liquid condensate sampling was conducted prior to and during testing and the samples were analyzed for fluorescein tracer to determine the amount of fluorescein that flowed out of the tubes of the Re-boiler into the steam side of the vessel. Given the established criteria for 'leak-tight,' test results determined that the Re-boiler passed the integrity assessment. After further scrutiny of the 2014 test and data results, it was determined that although the Re-boiler was 'leak-tight,' as established by the test criteria, there was concern of future leakage. The DOE and the DOE Office of River Protection (ORP) requested that WRPS repeat the integrity assessment in two and a half years to determine if the potential leak was growing, which could threaten continuing operations of the 242-A Evaporator. The 2017 test has been redesigned using lessons learned from the 2014 test and was modified to account for ergonomic and safety considerations, ease of repeatability, and accuracy of results. The liquid tracer methodology and test plan developed for the Re-boiler will continue to be implemented in 5 year increments to monitor current and future integrity. The method may also prove useful for other aging components across the Hanford Site, where appropriate. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Chapter 6: Recycling Plastic Waste to Produce Chemicals - A Techno-Economic Analysis and Life-Cycle Assessment

This study presents the results of a preliminary assessment on the technology, economics, and sustainability of using waste plastics as a feedstock to produce low molecular weight olefins via pyrolysis and gasification pathways. We also studied the formation of formaldehyde - an important polymer precursor - via the gasification of waste plastics. Process pathways for recovery of ethylene and propylene and production of formaldehyde from a simulated mixed plastic waste were constructed based on literature data. They included direct and indirect pathways based on either gasification or pyrolysis as the primary conversion step. Results of these studies have shown that significant economic challenges exist of producing olefins via pyrolysis and gasification, with the latter pathway especially difficult. Base costs of pyrolysis naphtha from waste plastics are higher when compared to the same material produced from fossil feedstocks. Base olefin costs for one direct route are two times higher than comparable costs from steam cracking of fossil naphtha. These costs are driven primarily by feedstock costs; some pyrolysis scenarios become more economically feasible when very low feedstock costs are used. Similarly, for gasification, the cost of methanol - the central intermediate - was found to be noticeable higher when produced by gasification of waste plastics compared to current selling prices, which negatively impacts all pathways that go through methanol as the central intermediate. Life-cycle assessment indicates that the production of pyrolysis naphtha from plastics is a carbon intensive operation; no major advantages were found in terms of greenhouse gas emissions for any of the pathways producing C2 and C3 olefins. The results of this study can serve as the baseline for future comparison to other plastic waste valorization processes.

BIOMASS FUELS,ENERGY PLANNING, POLICY, AND ECONOMY↗
DOE OSTI2025

Process Feasibility Analysis of Waste Biomass Valorization to Biochar and Bio-Oil via Slow and Fast Pyrolysis

The United States has abundant biomass and waste feedstock to support the nation's energy addition and affordability targets. Pyrolysis, a thermochemical conversion process, decomposes lignocellulosic feedstocks into liquid, solid, and gaseous fuels that can contribute to the domestic production of biofuels, biopower, and bioproducts. Growing private sector interest in this technology is a key motivation for this comprehensive techno-economic process modeling analysis of a respective biorefinery that includes feedstock preprocessing, slow and fast pyrolysis, and product separation to bio-oil, biochar, and syngas hydrocarbons. Results show that biochar from slow pyrolysis could achieve minimum selling prices (MSPs) of $\$$188-$\$$260/t, competitive with reported market values, while bio-oil from fast pyrolysis is estimated to yield MSPs of $\$$6.49-$\$$9.68/GGE, approximately twice conventional fuel benchmarks. Sensitivity analysis identifies feedstock cost, product yield, and scale as primary cost drivers, while scenarios involving biochar carbon credits and high value applications may substantially improve economics. Overall, these results suggest that continued innovation in feedstock logistics, process integration, and market development will be critical to achieving economically viable and scalable bioproducts.

09 BIOMASS FUELS↗
DOE OSTI2013

The effects of the mechanical–chemical stabilization process for municipal solid waste incinerator fly ash on the chemical reactions in cement paste

Highlights: ► Milling extracted MSWI fly ash. ► Increasing specific surface area, destruction of the crystalline texture, and increasing the amount of amorphous materials. ► Increasing heavy metal stability. ► Inducing pozzolanic reactions and increasing the early and later strength of the cement paste. - Abstract: A water extraction process can remove the soluble salts present in municipal solid waste incinerator (MSWI) fly ash, which will help to increase the stability of the synthetic materials produced from the MSWI fly ash. A milling process can be used to stabilize the heavy metals found in the extracted MSWI fly ash (EA) leading to the formation of a non-hazardous material. This milled extracted MSWI fly ash (MEA) was added to an ordinary Portland cement (OPC) paste to induce pozzolanic reactions. The experimental parameters included the milling time (96 h), water to binder ratios (0.38, 0.45, and 0.55), and curing time (1, 3, 7 and 28 days). The analysis procedures included inductively coupled plasma atomic emission spectroscopy (ICP/AES), BET, mercury intrusion porosimetry (MIP), X-ray diffraction (XRD), and nuclear magnetic resonance (NMR) imaging. The results of the analyses indicate that the milling process helped to stabilize the heavy metals in the MEA, with an increase in the specific surface area of about 50 times over that of OPC. The addition of the MEA to the OPC paste decreased the amount of Ca(OH){sub 2} and led to the generation of calcium–silicate–hydrates (C–S–H) which in turned increased the amount of gel pores and middle sized pores in the cement. Furthermore, a comparison shows an increase in the early and later strength over that of OPC paste without the addition of the milled extracted ash. In other words, the milling process could stabilize the heavy metals in the MEA and had an activating effect on the MEA, allowing it to partly substitute OPC in OPC paste.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2015

Characterisation of metals in the electronic waste of complex mixtures of end-of-life ICT products for development of cleaner recovery technology

Highlights: • New characterisation methodology has been established to understand an industrially processed ICT waste. • Particle size distribution, composition, thermal–chemical behaviour and occurrence of metals were considered. • The characterisation provides direct guidelines for values recovery from the waste. - Abstract: Recycling of valuable metals from electronic waste, especially complex mixtures of end-of-life information and communication technology (ICT) products, is of great difficulty due to their complexity and heterogeneity. One of the important reasons is the lack of comprehensive characterisation on such materials, i.e. accurate compositions, physical/chemical properties. In the present research, we focus on developing methodologies for the characterisation of metals in an industrially processed ICT waste. The morphology, particle size distribution, compositional distribution, occurrence, liberation as well as the thermo-chemical properties of the ICT waste were investigated with various characterisation techniques, including X-ray Fluorescence Spectrometry (XRF), differential scanning calorimetry (DSC) and scanning electron microscopy (SEM) with energy dispersed spectroscopy (EDS). Due to the high heterogeneity of the material, special sample preparation procedures were introduced to minimise the discrepancies during compositional analyses. As a result, a clearer overview of the ICT waste has been reached. This research provides better understanding of the extractability of each metal and improves the awareness of potential obstacles for extraction. It will lead to smarter decisions during further development of a clean and effective recovery process.

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

Experience with the Quality Control of Radioactive Waste Packages for the German Repository Konrad - 16493

The German repository Konrad for radioactive waste with negligible heat generation is designed to accept 303,000 m{sup 3} of waste packages in total. Based on the Konrad licensing conditions, any treatment and conditioning of radioactive waste should be planned and carried out in accordance with the waste acceptance requirements of the repository Konrad in order to produce waste packages suitable for disposal. The qualification of the waste treatment processes and the resulting products are achieved by interaction of different parties involved. On behalf of the Federal Office for Radiation Protection - the competent German authority for disposal of radioactive waste - TUV NORD EnSys Hannover GmbH and Co. KG acts as one of the independent expert organizations for more than twenty years in the field of qualification of waste treatment campaigns and review of waste packages documentations. The permitted amount of nuclide activity in the radioactive waste packages depends on the characteristics of the radioactive waste products. Projecting a suitable strategy for loading the waste containers contributes to determine in advance, what requirements the radioactive waste products have to fulfill as well as what properties of the waste products need to be verified and documented during conditioning or treatment process. On the basis of various examples, it is being demonstrated how appropriate advance process planning can ensure that radioactive waste packages are generated to comply with the waste acceptance requirements. In consequence, technical measures due to requalification can be avoided even in case of unpredictable difficulties. (authors)

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