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All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

On-Line Monitoring and Real-Time Characterization of Low Level Waste and Off-Gas Condensate Samples using Raman Spectroscopy - 19274

The demand for Hanford waste property analysis is expected to significantly increase as the U.S. Department of Energy River Protection Project progresses toward transferring waste to the Hanford Waste Treatment and Immobilization Plant. The analysis of waste includes chemical, radionuclide, and physical property characterizations, with the goal of achieving safe/informed transfer and processing. Waste sampling followed by offline laboratory analysis is the current baseline analytical approach. The addition of field-deployed/in situ instruments has the potential to provide improvements and benefits over offline analysis in the areas of 1) Worker Safety: by reducing the need for sampling and thereby reducing the need for worker interaction with waste during sample handling; 2) Operational Efficiency: by reducing analysis turnaround times and offsetting demand on analytical laboratories; and 3) Accuracy: continuous analysis of a flowing waste stream allows cumulative results to be produced for the batch, which is more representative than discrete samples of batch properties. This paper describes the development and application of an online monitor to provide chemical characterizations; specifically, Raman spectroscopy is utilized to identify and quantify tank components. Accurate quantification is achieved through the use of chemometric analysis, which can be configured to provide real-time analysis in field deployment settings. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Operational Safety Analysis for Disposal Period of the Future Belgian Geological Disposal Facility - 19249

For more than 40 years, ONDRAF/NIRAS (the federal agency responsible for managing radioactive waste and enriched fissile materials in Belgium) has been studying the concept of a geological disposal facility (GDF) in poorly indurated clays as an option for the long-term management of both category B waste (low-level and intermediate-level long-lived waste - LILW-LL) and category C waste (high-level waste - HLW and spent fuel). In the absence of a policy decision, the proposed Belgian reference option for disposal of such waste is a GDF in poorly indurated clays, located at a depth between 200 m and 600 m. ONDRAF/NIRAS has developed a reference disposal concept for research, development and demonstration (RD and D) purposes. At the current stage of this program, the primary objective of ONDRAF/NIRAS is to establish the safety and feasibility of constructing, operating and closing a Geological Disposal Facility (GDF) at the conceptual design level, without any presumption about the exact location of the facility. This will be evaluated in a first version of the safety and feasibility case (SFC-1 V1) for the geological disposal of category B and C radioactive wastes. Analysis of the operational safety is an important part of the SFC-1 V1. ONDRAF/NIRAS, supported by BGE Technology GmbH, has prepared an operational safety analysis of the current GDF reference design and the processes occurring during the operation and closure period. A SWIFT methodology (Structured What If Technique) is used to identify risks and the measures required to mitigate their consequences. In parallel to this SWIFT analysis, specific attention is also given to fire risk, especially in the blind-ending disposal galleries (DG). Based on understanding of fire risk analysis, recommendations suggest using exclusively electrically driven equipment to reduce fire loads and minimize the fire hazard. In addition, strict working procedures could prevent the presence of workers between the DG dead end and equipment presenting significant fire load. The use of mobile refuge chambers should also be part of the safety concept. (authors)

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DOE OSTI2016

Optimized Qualification of Legacy Waste for the Repository Konrad - 16629

The repository Konrad in Germany is planned for the acceptance of low level waste and intermediate level waste. The Wiederaufarbeitungsanlage Karlsruhe (WAK GmbH) has the assignment to decommission and dispose of the nuclear facilities on the site of the Karlsruhe Institute of Technology (KIT). The entire inventory of these installations as well as the existing waste materials stored in the interim storage of the HDB are planned to be sent to the repository Konrad and are therefore treated according to the respective acceptance criteria. These acceptance criteria were defined on the basis of safety considerations and first published as a draft in 1995. After full legal recognition in 2007 the acceptance criteria were revised in 2010 and 2014. All legacy waste treated prior to 1995 has to be re-qualified according to the current state of acceptance. The repository will not open for acceptance of waste materials until 2022. Up to now only a small amount of acceptable waste packages exist. Therefore WAK GmbH established a scheme to improve and accelerate the processes and to have 2 annual delivery capacities (5000 m{sup 3} in total) ready by the end of the decade. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Collaboration for the Management of the Bruce Power Major Component Replacement (MCR) Nuclear Waste - 19630

Bruce Power's Major Component Replacement (MCR) Program will generate large volume waste via steam generators (SGs) and intermediate-level nuclear waste, which will require specialized packaging and handling to safely receive and store it at Ontario Power Generation's (OPG) Western Waste Management Facility (WWMF). The refurbishment of a Candu reactor involves, in part, the removal and replacement of all of a reactor core's fuel channel assemblies. Bruce Power and OPG collaborated to develop nuclear waste management solutions to optimize the use of OPG's existing Steam Generator Storage Building (SGSB) and Retube Component Storage Building (RCSB). By designing specialized Retube Waste Containers (RWCs) and optimizing configurations for the waste to be stored in the buildings, we will reduce the interim nuclear waste storage needs at the OPG WWMF, while improving Bruce Power MCR process efficiencies and reducing the overall costs of the MCR Program (Units 3 through 8). (authors)

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

Estimating Residual Solids Volume In Underground Storage Tanks

The Savannah River Site liquid waste system consists of multiple facilities to safely receive and store legacy radioactive waste, treat, and permanently dispose waste. The large underground storage tanks and associated equipment, known as the 'tank farms', include a complex interconnected transfer system which includes underground transfer pipelines and ancillary equipment to direct the flow of waste. The waste in the tanks is present in three forms: supernatant, sludge, and salt. The supernatant is a multi-component aqueous mixture, while sludge is a gel-like substance which consists of insoluble solids and entrapped supernatant. The waste from these tanks is retrieved and treated as sludge or salt. The high level (radioactive) fraction of the waste is vitrified into a glass waste form, while the low-level waste is immobilized in a cementitious grout waste form called saltstone. Once the waste is retrieved and processed, the tanks are closed via removing the bulk of the waste, chemical cleaning, heel removal, stabilizing remaining residuals with tailored grout formulations and severing/sealing external penetrations. The comprehensive liquid waste disposition system, currently managed by Savannah River Remediation, consists of 1) safe storage and retrieval of the waste as it is prepared for permanent disposition; (2) definition of the waste processing techniques utilized to separate the high-level waste fraction/low-level waste fraction; (3) disposition of LLW in saltstone; (4) disposition of the HLW in glass; and (5) closure state of the facilities, including tanks. This paper focuses on determining the effectiveness of waste removal campaigns through monitoring the volume of residual solids in the waste tanks. Volume estimates of the residual solids are performed by creating a map of the residual solids on the waste tank bottom using video and still digital images. The map is then used to calculate the volume of solids remaining in the waste tank. The ability to accurately determine a volume is a function of the quantity and quality of the waste tank images. Currently, mapping is performed remotely with closed circuit video cameras and still photograph cameras due to the hazardous environment. There are two methods that can be used to create a solids volume map. These methods are: liquid transfer mapping / post transfer mapping and final residual solids mapping. The task is performed during a transfer because the liquid level (which is a known value determined by a level measurement device) is used as a landmark to indicate solids accumulation heights. The post transfer method is primarily utilized after the majority of waste has been removed. This method relies on video and still digital images of the waste tank after the liquid transfer is complete to obtain the relative height of solids across a waste tank in relation to known and usable landmarks within the waste tank (cooling coils, column base plates, etc.). In order to accurately monitor solids over time across various cleaning campaigns, and provide a technical basis to support final waste tank closure, a consistent methodology for volume determination has been developed and implemented at SRS.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

Report for Treating Hanford LAW and WTP SW Simulants: Pilot Plant Mineralizing Flowsheet

The US Department of Energy is responsible for managing the disposal of radioactive liquid waste in underground storage tanks at the Hanford site in Washington State. The Hanford waste treatment and immobilization plant (WPT) will separate the waste into a small volume of high level waste (HLW), containing most of the radioactive constituents, and a larger volume of low activity waste (LAW), containing most of the non-radioactive chemical and hazardous constituents. The HLW and LAW will be converted into immobilized waste forms for disposal. Currently there is inadequate LAW vitrification capacity planned at the WTP to complete the mission within the required timeframe. Therefore additional LAW capacity is required. One candidate supplemental treatment technology is the fluidized bed steam reformer process (FBSR). This report describes the demonstration testing of the FBSR process using a mineralizing flowsheet for treating simulated Hanford LAW and secondary waste from the WTP (WTP SW). The FBSR testing project produced leach-resistant solid products and environmentally compliant gaseous effluents. The solid products incorporated normally soluble ions into an alkali alumino-silicate (NaS) mineral matrix. Gaseous emissions were found to be within regulatory limits. Cesium and rhenium were captured in the mineralized products with system removal efficiencies of 99.999% and 99.998 respectively. The durability and leach performance of the FBSR granular solid were superior to the low activity reference material (LMR) glass standards. Normalized product consistency test (PCT) release rates for constituents of concern were approximately 2 orders of magnitude less than that of sodium in the Hanford glass [standard].

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

Process for Obtaining TI Waiver for Recalcitrant Radionuclides at DOE Sites - 18225

The current plan for treating radioactive iodine-129 ({sup 129}I) at the 200-UP-1 operable unit (OU) within the Central Plateau at the Hanford Site is to hydraulically contain the groundwater plume using injection wells placed at the leading edge of the plume. In the previous feasibility study for the OU, no treatment technologies were identified that could achieve the federal drinking water standard of 1 pCi/L. In addition to further study to identify viable remediation technology options, efforts were initiated to assess the potential human health and environmental risks associated with using hydraulic containment of {sup 129}I, and to refine transport estimates based on evidence of attenuation. An initial step was to develop and apply a waste-site scale integrated vadose zone-aquifer model. Because viable {sup 129}I remediation technologies for the OU may not be identified in current efforts, a process to facilitate examining the suitability of a technical impracticability waiver for the OU conditions was initiated by defining the monitoring data analysis, documentation, and simulation support that would be needed for obtaining a waiver. (authors)

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

Comprehensive process for the recovery of value and critical materials from electronic waste

The development of technologies that contribute to the proper disposal and treatment of electronic waste is not just an environmental need, but an opportunity for the recovery and recycle of valuable metals and critical materials. Value elements in electronic waste include gold, palladium, silver, copper, nickel, and rare earth elements (RE). Here, we present the development of a process that enables efficient recycling of metals from scrap mobile electronics. An electro recycling (ER) process, based on the regeneration of Fe 3+ as a weak oxidizer, is studied for the selective recovery of base metals while leaving precious metals for separate extraction at reduced chemical demand. A separate process recovers rare earth oxides from magnets in electronics. Furthermore, recovery and extraction efficiencies ca. 90 % were obtained for the extraction of base metals from the non-ferromagnetic fraction in the two different solution matrices tested (H 2 SO 4 , and HCl). The effect of the pre-extraction of base metals in the increase of precious metals extraction efficiency was verified. On the other hand, the extraction of rare earths from the ferromagnetic fraction, performed by means of anaerobic extraction in acid media, was assessed for the selective recovery of rare earths. We developed a comprehensive flow sheet to process electronic waste to value products.

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

Optimal planning for the sustainable utilization of municipal solid waste

Highlights: • An optimization approach for the sustainable management of municipal solid waste is proposed. • The proposed model optimizes the entire supply chain network of a distributed system. • A case study for the sustainable waste management in the central-west part of Mexico is presented. • Results shows different interesting solutions for the case study presented. - Abstract: The increasing generation of municipal solid waste (MSW) is a major problem particularly for large urban areas with insufficient landfill capacities and inefficient waste management systems. Several options associated to the supply chain for implementing a MSW management system are available, however to determine the optimal solution several technical, economic, environmental and social aspects must be considered. Therefore, this paper proposes a mathematical programming model for the optimal planning of the supply chain associated to the MSW management system to maximize the economic benefit while accounting for technical and environmental issues. The optimization model simultaneously selects the processing technologies and their location, the distribution of wastes from cities as well as the distribution of products to markets. The problem was formulated as a multi-objective mixed-integer linear programing problem to maximize the profit of the supply chain and the amount of recycled wastes, where the results are showed through Pareto curves that tradeoff economic and environmental aspects. The proposed approach is applied to a case study for the west-central part of Mexico to consider the integration of MSW from several cities to yield useful products. The results show that an integrated utilization of MSW can provide economic, environmental and social benefits.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

International Low Level Waste Management Practices and Technical Bases for Very Low Level Waste Classification - 19415

Different countries around the globe use varying schemes for the classification of radioactive waste. Many countries use adaptations of the International Atomic Energy Agency (IAEA) waste classification scheme. The United States uses the Class A, B, and C waste classification scheme that is loosely based on the IAEA concepts. Classification may be based on concentration limits or dose rate limits or a combination of concentration and dose rate limits. In some countries the existence of very low-level waste categories allows nuclear power plant waste to be disposed of in hazardous waste or shallow land burial sites separate from low and intermediate level waste disposal facilities. In the United States, the very low-level waste (VLLW) classification does not exist. Nuclear power plants use alternative disposal procedures under 10CFR20.2002 and other licensed processes in agreement states in order to dispose of very low activity waste in hazardous waste disposal sites. [1] Disposal of VLLW in hazardous waste or separate VLLW facilities is generally more cost effective than disposing of waste in radioactive waste sites and is protective of the health and safety of the public. This can become an important factor when large volumes of very low activity waste are generated during decommissioning of nuclear power plants. The Electric Power Research Institute (EPRI) Radiation Safety (RS) Program has developed technical bases for the establishment of a very low-level waste category. EPRI studied how VLLW waste is defined, managed, and disposed of in various countries. This included the study of disposal site design, characteristics, and how these sites protect the health and safety of the environment and the public from low levels of radioactivity. EPRI conducted a cost benefit analysis to understand the economic impact of disposing VLLW in hazardous waste disposal sites versus low-level waste disposal sites. Finally, EPRI calculated potential radionuclide limits that may be used to define very low-level waste. The purpose of this work was to provide an example of how one might approach defining limits for VLLW. The same or similar approaches could be applied to derive radionuclide limits for different dose limit objectives. This paper presents the results of research that explored the waste management and disposal practices at six countries around the world with additional focus and details on the implementation of the very low-level waste category. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Applying Technology and Experience to Ensure Safe In Situ Decommissioning - 19587

Permanent entombment of a radioactively contaminated facility as a decommissioning option (In Situ Decommissioning, ISD) has been completed for DOE facilities at the Savannah River Site (P and Reactors, HWCTR reactor), Hanford (U Canyon) and Idaho (CP-601/604),and is currently being considered for implementation at a limited number of selected DOE facilities as well as for the NPD and WR-1 reactors in Canada and various facilities in the United Kingdom. A general description of an ISD project encompasses an entombed facility which in some cases is limited to just the below-grade portion of a facility with the above-ground portion being decommissioned, packaged and disposed in a conventional manner. The entombed portions of the facility are of robust construction, generally of reinforced concrete exterior, that provides a migration barrier between internal contamination and the environment with significant internal void spaces backfilled or grouted. The scope of entombment may include ancillary equipment and structures and may contain radioactive and hazardous materials and contamination within the facility and waste imported from outside the facility. ISD is most often a permanent decommissioning end-state although in some instances (e.g. the U.K.) there may be a need to revisit the ISD facility after 300 years to effect traditional decommissioning. The defined completion (the end-state) of the decommissioned facility is therefore project-specific and in conformance with environmental approval processes and planned future activities. In all cases however, the final condition is passive, meaning that there are no requirements for ongoing operational systems or equipment within the decommissioned facility. ISD projects in the USA are presumed to be under indefinite institutional control of the U.S. Government. The regulatory framework in the U.S. is currently in place to provide assurance that the risk posed by an ISD facility is within regulatory acceptance criteria. (The framework in the U.K. and Canada is evolving towards that in the USA). Special emphasis is placed on the fact that an entombed facility is not considered a waste disposal facility; rather it is a decommissioning end-state option. ISD does not eliminate the need for effective and careful management of contaminated materials and structures, nor does it serve to abandon contaminated buildings in place. Further, ISD is safe and cost-effective for a relatively small number of facilities across the DOE Complex. This paper explores the current programmatic considerations and technical aspects of the ISD process and discuss the measures taken to ensure that an 'entombed' facility is safe for both site workers and the general population. (authors)

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DOE OSTI2024

Design and Materials of Reference Electrodes for Radioactive Waste Tank Service – A Literature Review

The Hanford site stores approximately 55 million gallons of radioactive and chemically hazardous wastes from the production of weapons materials. The wastes are stored in 177 underground, carbon steel storage tanks, 149 of these are single shell tanks (SSTs) and 28 of these are double shell tanks (DSTs). The DSTs provide critical retrieval and interim storage before the waste is vitrified in the Waste Treatment and Isolation Plant (WTP). The DSTs have been in service for 38 to 56 years and current plans indicate that WTP operations will be completed in 2075. Thus, the tanks will need to remain in service far beyond the initial 40-year life expectancy. For life extension of the tanks, effective corrosion control practices must remain in force. This effort includes direct measurements of the extent of corrosion (e.g., ultrasonic measurements and corrosion coupons) and electrochemical processes (e.g., linear polarization measurements).

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

Use of Groundwater Flow, Solute Transport, and Geochemical Modeling To Evaluate Long-Term Nitrate Plume Concentrations Following Phreatophyte Source Control - 17262

The Monument Valley, Arizona, Processing Site is a former uranium mill site within the Navajo Nation in northeastern Arizona. Milling activities between 1955 and 1968 resulted in contamination of groundwater in the shallow alluvium at the site. Remedial action was completed under Title I of the Uranium Mill Tailings Radiation Control Act of 1978. Site remediation activities between 1992 and 1994 removed surficial source material and other site-related contamination, but contamination of deeper soil, primarily by nitrate, remained between the excavation depth and the water table. Residual nitrate in these soils, from uranium ore processing at the mill, has continued to be a source of groundwater contamination. To evaluate the potential for native phreatophytes to control remaining source material, four-wing salt-bush and black grease-wood were planted in 2000 and 2005 within the footprints of the former tailings piles. The objectives of the plantings were to increase nitrate removal from soil water through root uptake and, because phreatophyte water usage is expected to equal or exceed annual precipitation, to halt nitrate loading to groundwater. Recent evaluation showed more than an 80% decrease in soil nitrate concentrations and no effective groundwater recharge in the planting area. Numerical groundwater flow, solute transport, and geochemical modeling were performed to evaluate the long-term response of the existing nitrate plume to the phreatophyte source control measures. This modeling will help determine if 100-year natural flushing is a viable compliance strategy for the nitrate plume. Results showed that nitrate concentrations in the alluvial aquifer are expected to drop below the standard of 10 mg/L as N in approximately 70 years as a result of the phreatophyte planting. Companion sensitivity analysis was also performed to better understand how uncertainty in groundwater flow model and solute transport model inputs would impact cleanup times. Even considering input uncertainty, nitrate concentrations are expected to reach the standard in the alluvial aquifer within 100 years. Recognizing that model predictions are not absolute, but represent the general potential for an event to occur, the model was used to identify key sampling locations to verify that the nitrate plume will attenuate as predicted. This evaluation identified locations that can be used to assess the viability of the modeling predictions within the next 50 years. If the source diminishes or no longer exists, modeling predicts that concentrations of nitrate and ammonium in these wells should show significant declines over the next 10-55 years. If downward concentration trends do not occur as predicted, other remedial strategies may need to be evaluated to meet groundwater standards within the specified 100-year duration. (authors)

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DOE OSTI2015

Public Preferences Related to Radioactive Waste Management Nuclear Energy and Environment: Methodology and Response Reference Report for the 2014 Energy and Environment Survey

The results described in this report are from nationwide surveys between 2006 and 2014 on preferences of US residents concerning the environment and energy sources. The most recent 2014 survey was undertaken to determine how consent, in the context of nuclear facility siting, is understood and evaluated by a cross-section of the American public. Continuing attention to the events at the Fukushima nuclear facility, coupled with its negative implications for public support for nuclear energy, has changed the balance of risk and benefit perceptions, and thus the context in which nuclear facility siting efforts will occur. In addition, the portion of the public most concerned about climate change have typically been those most concerned about the environment in general, and in turn, are those that have traditionally been hostile to nuclear energy. The overall survey results show that the broader public is not well informed about the nuclear fuel cycle including energy production from nuclear reactors and current policies for the management of used nuclear fuel. In addition, the surveys find that respondents are reluctant to continue to rely on temporary on-site storage of used nuclear fuel, and that there is moderate support for developing one or more interim storage facilities. The survey responses also suggest that the level of trust accorded a new nuclear waste authority by the public will be sensitive to how it is institutionally defined. For a hypothetical community, which had volunteered to host an interim storage facility within 50 miles of their homes, the majority of respondents indicated that veto authority in the siting process should be limited to (a) majorities of local and statewide voters, (b) state and federal environmental regulatory authorities, and (c) governors. The respondents also indicated that the state and local host community should be permitted to withdraw consent up to the point at which a license to build the facility is submitted, but not after the license is received and facility construction initiated. The 2013 and 2014 surveys posed a set of questions concerning respondents' expectations about engaging in the process for siting an interim storage facility. Roughly half of the respondents said they would likely attend informational meetings, and nearly half said they would likely communicate with elected officials or would likely express their views on the topic via social media. Relatively few -- about one in five -- said they would be likely to actively participate in support for or opposition of an ISF. Finally, the 2014 results suggest that while respondents are more likely to support a citizen-led deliberative panel process than one led by experts, this preference does not seem to influence their opinion about the expected outcome of the siting process.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

PREPARATION OF U-PLANT FOR FINAL DEMOLITION AND DISPOSAL - 12109E

The U-Plant is one of the five major nuclear materials processing facilities at Hanford and was chosen as a pilot project to develop the modalities for closure of the other four facilities at Hanford and the rest of the Department of Energy (DOE) complex. The remedy for this facility was determined by a Record of Decision (ROD) pursuant to the Comprehensive Environmental Response, Compensation and Liability Act of 1980 (CERCLA). That remedy was to 'Close in Place - Partially Demolished Structure'. The U-Plant facility is identified as the 221-U Building and is a large, concrete structure nominally 247m (810 ft) long, 20 M (66 ft) wide and 24 m (77 ft) high with approximately 9 m (30 ft) being below grade level. It is a robust facility with walls ranging from 0.9 m to 2.7 m (3 ft to 9 ft) thick. One large room extends the entire length of the building that provides access to 40 sub-grade processing cells containing tanks, piping and other components. The work breakdown was divided into three major deliverables: (1) Tank D-10 Removal: removal of Tank D-10, which contained TRU waste; (2) Equipment Disposition: placement of contaminated equipment in the sub-grade cells; and (3) Canyon Grouting: grouting canyon void spaces to the maximum extent practical. A large number of pieces of contaminated equipment (pumps, piping, centrifuges, tanks, etc) from other facilities that had been stored on the canyon operating floor were placed inside of the sub-grade cells as final disposition, grouted and the cell shield plug reinstalled. This action precluded a large volume of waste being transported to another burial site. Finally, {approx}19,000 m3 ({approx}25,000 yd3) of grout was placed inside of the cells (in and around the contaminated equipment), in the major galleries. the ventilation tunnel, the external ventilation duct, and the hot pipe trench to minimize the potential for void spaces and to reduce the mobility, solubility, and/or toxicity of the grouted waste. The interim condition of the facility is 'cold and dark'. Upon availability of funding the structure will have contamination fixative applied to all contaminated surfaces and may be explosively demolished, with the remaining structure buried under an engineered barrier.

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

ITEAM - An international mechanistic approach to modeling glass behavior over geologic time scales - 15527

Vitrification is a common worldwide solution for radioactive waste immobilization. Although the excellent chemical durability of glass has been demonstrated in ancient man-made and natural analogues, the lack of an international consensus on a glass corrosion rate law limits the application of glass in high-level waste repository performance models. The contributions of physical and chemical processes controlling the long-term glass dissolution rate in geologic disposal remain uncertain resulting in this non-consensus. A thorough mechanistic understanding of waste form durability in geologic environments would improve public and regulator confidence. In addition, this understanding could lead to cost savings if it becomes possible to take credit for the true durability of the waste form itself in system evaluations rather than requiring expensive engineered barrier systems. To this end, five nations (US, UK, France, Japan, and Belgium) have joined together to formulate a joint plan for collaborative research into the mechanisms controlling the long-term corrosion of glass. Over the last five years, the coordinated research performed by these countries has produced a new level of understanding about the glass corrosion mechanisms, specifically as they apply to glasses similar in composition to the French R7T7 glass. Some recent results are presented as they relate to several mechanisms that are thought to control glass alteration at long time periods. It is important to understand the effects of each particular mechanism on the overall radionuclide release and the relative importance of each process model to the performance assessment calculation. In addition, a new series of experiments calls out potential sources of error due to the use of irregular shapes (such as crushed glass powder) in dissolution experiments. (authors)

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

Preparation of Four Large-format Hot Isostatically Pressed I-AgZ Waste Form Samples for Performance Testing

Hot isostatic pressing (HIP) is being considered as a process to directly convert 129 I-bearing materials to a durable radiological waste form. Zeolites, specifically silver-exchanged mordenite (AgZ), have been studied extensively as potential radioactive iodine sorbents and will contain 129 I as chemisorbed AgI, a relatively insoluble compound with a high melting point. Oak Ridge National Laboratory has conducted several recent studies on the HIP of both I–AgZ and other iodine-bearing zeolites, resulting in an inventory of samples prepared with a suite of zeolites with varied iodine incorporation methods, pressing conditions, and sizes. These samples have been made available to researchers at other national laboratories in support of test method development to assess the stability of heterogenous iodine-bearing waste forms. The work described here generated four large-format HIPed I–AgZ samples. These samples are 1.5 in. diameter and 2 in. length and contain 49 g of I–AgZ. Two samples were prepared at an iodine loading of 64 mg I/g I– AgZ, and two samples were prepared at an iodine loading of 135 mg I/g I–AgZ. These two sets of duplicate samples will be made available upon request in support of iodine waste form durability method development efforts.

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

Thermal properties and Thermodynamic Equilibrium Modeling of Cementitious Waste Forms

INTRODUCTION Cementitious matrices are used in the US DOE complex and worldwide to solidify aqueous radioactive, hazardous, mixed salt solutions, and sludges to meet low-level radioactive waste (LLW) disposal requirements. Blended formulations are used for most waste form mix designs. Substitution of pozzolans for Class F fly ash has the potential to alter the processing properties and stabilization properties of the resulting waste forms because they add chemical and mineralogical complexity to the final material. The Savannah River Site saltstone waste form was selected as the test case material. CemGEMS software was chosen as the model for predicting hydrated cementitious phases assemblages and phase evolution. Isothermal calorimetry was used as the method for evaluating processing and properties of fresh, uncured waste forms. The test cases consisted of reference case saltstone, and five natural pozzolan substituted saltstone mixes.

Bustamante, Michael E. [Savannah River National La↗