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Performance Assessment Results for the Hanford Integrated Disposal Facility - 19134

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

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Grout testing to support tank closure at the Savannah River Site - 15370

Final operational closure of the Savannah River Site (SRS) waste tanks is performed by filling the voids inside the tanks with a series of tailored grout formulas thereby stabilizing the remaining residuals. These tailored grout formulas were each developed to meet the demands of the closure process. The needs identified for the closure process include (1) a flowable, self-consolidating, reducing grout for tank fill, (2) a highly flowable grout capable of filling the 7-1/4 kilometers (4-1/2 miles) of 5-centimeter (2-inch) diameter cooling coils inside the tank, and (3) a highly flowable grout capable of filling equipment remaining in the tank. The specific characteristics of each of the grout formulations were tested to ensure they would function as required during implementation. Performance testing was also done on some of the grout formulations to confirm mix effectiveness and improve worker handling proficiency and conducted additional testing on the bulk fill grout to expand the placement capabilities of the grout. This paper focuses on the evolution of the closure process' identified needs, formulation development, and the subsequent testing of the tailored grout formulations utilized in the past tank closure processes at SRS. (authors)

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Development of an Integrated Leachate Treatment Solution for the Port Granby Waste Management Facility - 12429

The Port Granby Project (the Project) is located near the north shore of Lake Ontario in the Municipality of Clarington, Ontario, Canada. The Project consists of relocating approximately 450,000 m{sup 3} of historic Low-Level Radioactive Waste (LLRW) and contaminated soil from the existing Port Granby Waste Management Facility (WMF) to a proposed Long-Term Waste Management Facility (LTWMF) located adjacent to the WMF. The LTWMF will include an engineered waste containment facility, a Wastewater Treatment Plant (WTP), and other ancillary facilities. A series of bench- and pilot-scale test programs have been conducted to identify preferred treatment processes to be incorporated into the WTP to treat wastewater generated during the construction, closure and post-closure periods at the WMF/LTWMF. (authors)

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Continuous high-solids anaerobic co-digestion of organic solid wastes under mesophilic conditions

Highlights: > High-solids (dry) anaerobic digestion is attracting a lot of attention these days. > One reactor was fed with food waste (FW) and paper waste. > Maximum biogas production rate of 5.0 m{sup 3}/m{sup 3}/d was achieved at HRT 40 d and 40% TS. > The other reactor was fed with FW and livestock waste (LW). > Until a 40% LW content increase, the reactor exhibited a stable performance. - Abstract: With increasing concerns over the limited capacity of landfills, conservation of resources, and reduction of CO{sub 2} emissions, high-solids (dry) anaerobic digestion of organic solid waste (OSW) is attracting a great deal of attention these days. In the present work, two dry anaerobic co-digestion systems fed with different mixtures of OSW were continuously operated under mesophilic conditions. Dewatered sludge cake was used as a main seeding source. In reactor (I), which was fed with food waste (FW) and paper waste (PW), hydraulic retention time (HRT) and solid content were controlled to find the maximum treatability. At a fixed solid content of 30% total solids (TS), stable performance was maintained up to an HRT decrease to 40 d. However, the stable performance was not sustained at 30 d HRT, and hence, HRT was increased to 40 d again. In further operation, instead of decreasing HRT, solid content was increased to 40% TS, which was found to be a better option to increase the treatability. The biogas production rate (BPR), CH{sub 4} production yield (MPY) and VS reduction achieved in this condition were 5.0 m{sup 3}/m{sup 3}/d, 0.25 m{sup 3} CH{sub 4}/g COD{sub added}, and 80%, respectively. Reactor (II) was fed with FW and livestock waste (LW), and LW content was increased during the operation. Until a 40% LW content increase, reactor (II) exhibited a stable performance. A BPR of 1.7 m{sup 3}/m{sup 3}/d, MPY of 0.26 m{sup 3} CH{sub 4}/g COD{sub added}, and VS reduction of 72% was achieved at 40% LW content. However, when the LW content was increased to 60%, there was a significant performance drop, which was attributed to free ammonia inhibition. The performances in these two reactors were comparable to the ones achieved in the conventional wet digestion and thermophilic dry digestion processes.

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Performance Test of Removal of Gaseous Radioactive Methyl Iodide by TEDA-Impregnated Activated Carbon - 18049

During the operation of nuclear facility as well as decommissioning stage, various radioactive gaseous wastes are generated. It is generally recognized that radioiodine is potentially one of the most hazardous gas on the human body and environment. Iodine-131 release in the gaseous effluents at the nuclear facilities had been a critical environmental concern. Reducing iodine release is a critical issue in radioactive waste system evaluation, especially in their adherence to the 'as low as reasonably achievable' (ALARA) criteria because radioiodine among these gaseous wastes has radiological harm. Therefore, numerous studies for the removal of radioiodine from the off-gas stream of nuclear facilities have been performed with formation processes and chemical structure. This paper aims to evaluate the performance of removal efficiency of radioiodine gas by TEDA-impregnated activated carbon, which is used for radioiodine retention in nuclear facilities. The removal efficiencies of methyl iodide were evaluated by using a radioactive methyl iodide tracer at various temperatures up to 250 deg. C. (authors)

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HEPA Filter Waste Management at Select Department of Energy National Laboratories - 17022

High Efficiency Particulate Air (HEPA) filters are routinely used to filter exhaust air at Department of Energy (DOE) National Laboratory facilities that work with radioactive materials. While the HEPA filter configurations are fairly standardized (such as metal framed with glass micro-fiber media), the range and concentrations of radioactive and potentially other hazardous materials in the used filters presents for a variety of characterization approaches as well as challenges. This paper will summarize, compare and contrast the current HEPA filter waste management experience at a selection of DOE National Laboratories. Topic areas to be explored include process knowledge collection, characterization methods, characterization results, waste packaging and disposition pathways, and problems and success stories. (authors)

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CORROSION TESTING OF CARBON STEEL IN OXALIC ACID CHEMICAL CLEANING SOLUTIONS

Radioactive liquid waste has been stored in underground carbon steel tanks for nearly 60 years at the Savannah River Site. The site is currently in the process of removing the waste from these tanks in order to place it into vitrified, stable state for longer term storage. The last stage in the removal sequence is a chemical cleaning step that breaks up and dissolves metal oxide solids that cannot be easily pumped out of the tank. Oxalic acid has been selected for this purpose because it is an effective chelating agent for the solids and is not as corrosive as other acids. Electrochemical and immersion studies were conducted to investigate the corrosion behavior of carbon steel in simulated chemical cleaning environments. The effects of temperature, agitation, and the presence of sludge solids in the oxalic acid on the corrosion rate and the likelihood of hydrogen evolution were determined. The testing showed that the corrosion rates decreased significantly in the presence of the sludge solids. Corrosion rates increased with agitation, however, the changes were less noticeable.

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Freeze-casting as a Novel Manufacturing Process for Fast Reactor Fuels. Final Report

Advanced burner reactors are designed to reduce the amount of long-lived radioactive isotopes that need to be disposed of as waste. The input feedstock for creating advanced fuel forms comes from either recycle of used light water reactor fuel or recycle of fuel from a fast burner reactor. Fuel for burner reactors requires novel fuel types based on new materials and designs that can achieve higher performance requirements (higher burn up, higher power, and greater margins to fuel melting) then yet achieved. One promising strategy to improved fuel performance is the manufacture of metal or ceramic scaffolds which are designed to allow for a well-defined placement of the fuel into the host, and this in a manner that permits greater control than that possible in the production of typical CERMET fuels.

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Radioactive waste management treatments: A selection for the Italian scenario

The increased attention for radioactive waste management is one of the most peculiar aspects of the nuclear sector considering both reactors and not power sources. The aim of this paper is to present the state-of-art of treatments for radioactive waste management all over the world in order to derive guidelines for the radioactive waste management in the Italian scenario. Starting with an overview on the international situation, it analyses the different sources, amounts, treatments, social and economic impacts looking at countries with different industrial backgrounds, energetic policies, geography and population. It lists all these treatments and selects the most reasonable according to technical, economic and social criteria. In particular, a double scenario is discussed (to be considered in case of few quantities of nuclear waste): the use of regional, centralized, off site processing facilities, which accept waste from many nuclear plants, and the use of mobile systems, which can be transported among multiple nuclear sites for processing campaigns. At the end the treatments suitable for the Italian scenario are presented providing simplified work-flows and guidelines. (authors)

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Management of Legacy Enriched Uranium and Transuranic Materials at ORNL - 19062

The DOE manages an inventory of materials that contains a range of long-lived radioactive isotopes that were produced from the 1960's through the 1980's. These materials were made by irradiating targets in production reactors to produce special heavy isotopes for DOE programmatic use, scientific research, and industrial and medical applications. ORNL uses these materials in DoE's center for production, storage, and distribution of transuranium isotopes (plutonium through californium) for the heavy-element research program and has used them as feedstock in the Calutron Electromagnetic Isotope Enrichment Facility, one of only two facilities in the world with capabilities to enrich radioisotopes in multigram quantities. Both the production reactors and enrichment facilities have been shut down, and many of these unique materials will not be produced again in the foreseeable future [1, 2]. As a result, ORNL has an inventory of radioisotopes that are being held for reuse because their potential intrinsic value to DOE, but many of these materials have no currently defined use. ORNL is undertaking an initiative to more actively manage these materials by reviewing the existing inventory to determine if the contents have programmatic use. Steps are being taken to process, repackage, and stage the materials for distribution to programmatic users or for waste disposal, primarily at WIPP. This paper describes the approach that ORNL has taken to determine whether the inventory of materials should be kept or disposed and the processes that are being used to manage or disposition the materials. (authors)

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Sludge Batch 10 Flowsheet Testing with Non-radioactive Simulants

Fourteen Chemical Processing Cell (CPC) simulations were performed with nonradioactive sludge simulants at the Aiken County Technology Laboratory in Aiken, SC. Four of these experiments were performed with Tank 51 sludge simulant. The remaining ten were performed with Tank 40 sludge simulant. The purpose of these experiments was to elucidate the chemistry and characteristics of Sludge Batch (SB) 10 as anticipated in the Defense Waste Processing Facility (DWPF). Experiments were performed at acid stoichiometries between 76% and 138% of the Koopman Minimum Acid requirement (85% - 144% of the Hsu acid requirement) and at REDuction/OXidation (REDOX) targets between 0.1 and 0.3. Testing examined the impact of coupled operations and sludge-only operations during Sludge Receipt and Adjustment Tank (SRAT) and Slurry Mix Evaporator (SME) processing at both design basis and nominal boilup rates. This report shares conclusions made as a result of this testing.

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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.

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The S.An.A.® concept: Semi-aerobic, Anaerobic, Aerated bioreactor landfill

Highlights: • Sustainable Hybrid bioreactor landfill lab-scale application. • Semi-aerobic, Anaerobic, Aerated (S.An.A.®) concept. • Enhancement of methane production and biochemical processes. • Aeration and flushing for reaching Final Storage Quality. - Abstract: Hybrid Bioreactor Landfills are designed to enhance and speed up biological processes, aiming at reducing the duration of post operational phase until landfill completion. S.An.A.®(Semi-aerobic, Anaerobic, Aerated) concept consists in a Hybrid Bioreactor featuring a first semi-aerobic phase to enhance the methane production occurring in the following anaerobic step and a forced aeration for the abatement of the residual emissions. At the end of the last step, semi-aerobic conditions are restored and flushing applied for leaching residual non-biodegradable compounds. Results of the application of S.An.A.® concept to a lab scale bioreactor system showed that pre-aeration was effective in controlling the concentration of VFA, increasing pH and stimulating methane production during anaerobic phase; in particular with intermittent airflow the methane potential was 50% higher respect to control reactors. Forced aeration reduced organic compounds and nitrogen concentration in leachate of an order of magnitude, better performing in low airflow reactors. S.An.A.® Hybrid bioreactors proved to be an efficient system both for increasing methane production and reaching landfill completion in shorter time, suggesting that with proper landfill management, the duration of post-closure care might be reduced by 25–35%.

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Considerations for Involving Citizen Science in an Integrated Waste Management System - 17506

The U.S. Department of Energy's (DOE's) goal is to develop solutions for the long-term, sustainable management of the nation's spent nuclear fuel (SNF) and high-level radioactive waste (HLW). The DOE has been exploring options for an integrated waste management system to transport, store, and dispose of SNF and HLW from commercial electricity generation, as well national defense activities. To support its efforts, DOE has been exploring ways to seek broad input from the public. The role of citizen science in a wide range of projects is increasing. In the U.S. 2013 and 2015 Open Government National Action Plans, the U.S. government encouraged all Federal agencies to pursue public participation through citizen science, and a large number of agencies and organizations have responded. This paper explores the potential for considering citizen science projects as part of an integrated waste management system, including the siting of facilities. A literature review was performed to gain insights into how citizen science projects can be effectively developed and deployed, including success factors and lessons learned. Results highlight the experiences shared by many people who have developed and implemented citizen science projects, which provide valuable context for potential application to a facility siting process. (authors)

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Historic and modern nuclear graphite impurities: Pathways to improved waste strategies

Graphite has been used in large volumes as a structural material and neutron moderator since the earliest days of nuclear fission. However, no international consensus exists on the disposal of irradiated graphite, leaving much of the historic radioactive graphite inventory in interim vault or silo storage. With several new graphite-moderated reactors planned or under construction, the issue of graphite waste management is becoming increasingly urgent. This paper reviews and quantifies impurities in both historic and modern nuclear graphite, with emphasis on nitrogen—responsible for much of the 14 C inventory—and chlorine, which plays a critical role in repository performance and design. Modern graphites, benefitting from stringent quality-control measures developed for non-nuclear industries, meet or exceed the ASTM Ultra-High Purity nuclear standards, even without halide purification. Both chlorine and nitrogen concentrations have declined over time. For chlorine, identified as a key impurity influencing U.S. waste repository design, we propose a target of 0.1 appm in as-fabricated billets as a reasonable benchmark. Nitrogen sources are traced throughout the graphite production process, with surface and bulk concentrations characterized for all materials studied. Modern graphites commonly exhibit nitrogen levels below 5 appm, with values approaching 1 appm achievable. Using such reduced-nitrogen grades is critical to keeping graphite-induced radioactivity below the greater-than-Class-C waste threshold, thereby avoiding disposal cost penalties of nearly an order of magnitude.

Chlorine↗

Geological criteria for site selection of an LILW radioactive waste repository in the Philippines

In the selection of sites for disposal facilities involving low- and intermediate-level radioactive waste (LILW), International Atomic Energy Agency (IAEA) recommendations require that 'the region in which the site is located shall be such that significant tectonic and surface processes are not expected to occur with an intensity that would compromise the required isolation capability of the repository'. Evaluating the appropriateness of a site therefore requires a deep understanding of the geological and tectonic setting of the area. The Philippines sits in a tectonically active region frequented by earthquakes and volcanic activity. Its highly variable morphology coupled with its location along the typhoon corridor in the west Pacific region subjects the country to surface processes often manifested in the form of landslides. The Philippine LILW near surface repository project site is located on the north eastern sector of the Island of Luzon in northern Philippines. This island is surrounded by active subduction trenches; to the east by the East Luzon Trough and to the west by the Manila Trench. The island is also traversed by several branches of the Philippine Fault System. The Philippine LILW repository project is located more than 100 km away from any of these major active fault systems. In the near field, the project site is located less than 10 km from a minor fault (Dummon River Fault) and more than 40 km away from a volcanic edifice (Mt. Caguas). This paper presents an analysis of the potential hazards that these active tectonic features may pose to the project site. The assessment of such geologic hazards is imperative in the characterization of the site and a crucial input in the design and safety assessment of the repository. (authors)

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Modeling glass degradation and release of radionuclides from vitrified waste for performance assessment simulations

The release of radionuclides initially encapsulated in a slowly degrading solid waste form and contained in an eventually corroding canister defines the source term for numerical simulations for the assessment of a geologic repository for high-level radioactive waste. While the details of waste degradation, canister corrosion, and dissolution and mobilization of the radionuclides in pore water include complex chemical reaction and transport processes that are coupled to the thermal, hydrological, microbiological, and mechanical conditions in the repository, the source-term model suitable for use in a numerical performance assessment model should be a defensible abstraction of these mechanisms. We developed a radiological source-term model and implemented it into a non-isothermal flow and transport simulator. While the proposed source-term model is applicable to various waste forms, canister systems, and disposal concepts, we specifically considered radionuclide releases from vitrified high-level waste placed in a cylindrical canister disposed in a deep vertical borehole repository. In this model, waste degradation is a function of temperature, and it can be adjusted to evaluate the influence of and propagate uncertainties in pH, passivation reactions, and chemical conditions as well as geometrical factors. The time-dependent, congruent release of safety-relevant radionuclides present in the decaying inventory is then calculated. Finally, the radionuclides are mobilized by diffusive and advective transport according to the thermo-hydraulic conditions prevailing in the near field of the repository, from where they migrate through the geosphere to the accessible environment. We examine the influence of the source-term model’s parameters on performance assessment calculations through sensitivity and uncertainty propagation analyses, identifying influential factors and confirming the upper bound of their impact. These considerations align with the overarching goal of repository design, which is to demonstrate that engineered and natural barriers can collectively delay radionuclide migration for timescales far exceeding human planning, thereby providing multiple, redundant barriers against environmental contamination.

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