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DOE OSTI2020

Sol–gel synthesis of iodosodalite precursors and subsequent consolidation with a glass binder made from oxides and sol–gel routes

Radioiodine accumulates in aqueous solutions and off-gas streams during nuclear fuel reprocessing due to its solubility in aqueous solutions and volatility, respectively. Additionally, radioiodine is highly mobile in geological environments and iodine-129 has been found to be one of the largest long-term contributors to chronic dose to humans in scenarios of radionuclide migration from geological repositories. Most of the radioiodine can be captured during fuel reprocessing in off-gas streams using solid sorbents and scrubbing solutions. Once iodine is captured, it must be stored in a durable form for eventual disposal. Iodosodalite has been investigated as a waste form for radioiodine, however these synthesis processes typically result in mixed products and iodine volatilization during consolidation. Therefore, this paper proposes a novel approach to synthesizing iodosodalite utilizing a sol-gel method. This method was chosen to lower processing temperatures and improve product yield. Preliminary experiments conducted to determine the viability of this synthetic method are presented. Additionally, consolidation of sol-gel derived iodosodalite with a glass-binder was explored using three different methods: 1) incorporating the glass-binder during gel preparation using alkoxide precursors; 2) separately preparing the glass binder using a sol-gel method; and 3) separately preparing the glass-binder using a melt-quench technique. Glass-bonded iodosodalite was successfully synthesized using these novel sol-gel approaches.

36 MATERIALS SCIENCE↗

DOE OSTI2017

The Reactor Graphite Handling. Problems and Solutions - 17208

The problem of industrial uranium-graphite reactors and NPP units decommissioning is one of the urgent tasks that require efficient and cost-effective solutions. To date, only Russia has already shut down and subject to decommissioning more than 120 objects. World-wide, more than 260,000 ton of irradiated graphite have arisen from the commercial nuclear power operations, research reactors, and defense production reactors. Under the circumstances the problem to develop cost-effective methods of the irradiated graphite management has appeared. Alternative disposal options for the irradiated graphite were discussed for at least 25-30 years. To date, various methods of irradiated graphite reprocessing have been proposed. They are based on the application of different physical and chemical processes, most of which involve incineration of the entire mass of graphite. However, they all have disadvantages, and do not solve the whole complex of problems. The primary goal of the research presented here was to carry out a study aimed at the development of a more simple technology (as compared to the incineration of the entire mass of graphite) that would provide for the graphite purification and allow the possibility of its transfer to the less hazardous waste category followed by the disposal of it in a near-surface repository. To achieve it is proposed to consider two directions, each of them reducing the cost of graphite reprocessing and the waste volume disposed of in the repository. The first alternative is based on the principle of partitioning radionuclides of different nature from the graphite, thus allowing to transfer it into the LLW category and minimize the volume of secondary waste products. The second option to change the reprocessing technology is the idea to use reactor graphite as the starting material to produce stable final waste form compositions intended for the long-term storage of radioactive waste including transuranic elements. (authors)

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DOE OSTI2023

Prioritizing Off-Gas Metrics: A Guide for Comparable Off-Gas Capture Testing

The Material Recovery and Waste Form Development (MRWFD) off-gas team had a workshop, hosted by Idaho National Laboratory (INL), to align goals and expectations for off-gas research. The workshop included team members from four national laboratories. The workshop focused on defining distinct R&D phases with specific metrics, outlining standard test and measurement protocols for Iodine and Krypton/Xenon sorbents, brainstorming approaches to future disruptive technologies, and recognizing parameters with more inherent risk, requiring more rigorous evaluation. This report will serve as a guide for future off-gas work. Its purpose is to foster efficient collaboration across diverse research facilities and invite direct comparison of materials and results.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

Development of a Wide Range Segmented Gamma Scanner for Measuring Radioactive Wastes Arising in the Chinese Nuclear Industry - 16261

There is a requirement to measure a variety of challenging radioactive waste forms generated by the Chinese nuclear industry. These range from low activity waste to high activity waste with a high content of Cs-137 and waste in high density 200 litre drums. This paper describes the design, development and testing of an updated and automated Wide Range Segmented Gamma Scanner (WR-SGS) specifically configured for Chinese nuclear industry radioactive waste measuring requirements. The instrument retains the capabilities of the original SGS developed by Parker and Martin at the Los Alamos National Laboratory and now widely used for radioactive waste assay. It also incorporates additional features to improve measurement efficiency and reduce cost. These features include continuous helical scanning of the waste drum, and separate lifting pillars to lift the detector and the transmission source. In place of a transmission source shutter mechanism, the source is placed in a shielded safe when not in use. The design also incorporates a variable aperture collimator, and tungsten filters that can be deployed for the measurement of high activity, high dose-rate drums. Updated motion control electronics, based on the use of a programed logic controller have been incorporated to address potential future obsolescence. The WR-SGS employs a 40% high purity Germanium detector with electro-mechanical cooling and state of the art digital spectroscopy counting electronics for gamma ray spectrum data acquisition. Testing of the unit has been performed with both point sources and rod sources, which are traceable to international standards. Typical test drums, each with six distributed re-entrant tubes and with different density matrices, have been employed with the rod sources to validate both the calibration and the measurement performance of the WR-SGS for drums of different density. The measurement results have been compared with the reference decay corrected activity values and very good agreement is achieved, confirming the measurement performance of the instrument. (authors)

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DOE OSTI2020

Panel Session 98B: Treatment of Low-Activity Tank Waste at Hanford: Considering the Alternatives

This panel focused on alternative treatment of the Hanford Low-Activity Tank Waste. Currently, Tank Waste at the Hanford Site is to be separated into low-activity (LAW) and high-level waste (HLW) fractions, and then vitrified to produce immobilized LAW glass for onsite disposal, and HLW glass for ultimate geologic disposal. This panel discussed several commissioned studies by the National Laboratories, National Academies, and General Accountability Office to evaluate the risks, costs, and benefits associated with considering options for immobilizing portions of the Hanford LAW waste forms. Panelists with presentations: Analysis of Supplemental Treatment Approaches for Low-Activity Waste at the Hanford Nuclear Reservation Overview and Conclusions (Michael Stone); Analysis of Supplemental Treatment Approaches for Low-Activity Waste at the Hanford Nuclear Reservation: Disposal Performance Evaluation (Tom Brouns); National Academies Review of the Study on Supplemental Treatment Approaches of Low-Activity Waste at the Hanford Nuclear Reservation (Charles Ferguson); GAO Observations on Immobilizing Low-Activity Waste at DoE's Hanford Site (David Trimble)

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DOE OSTI2024

Development of Magnesium Oxysulfate Formulation for SRPPF Aqueous Recovery Liquid Solidification

The liquid effluent from the Savannah River Plutonium Processing Facility (SRPPF) Aqueous Recovery Processes will be solidified into a stable form that is acceptable by Waste Isolation Pilot Plant (WIPP) for disposal. The current Aqueous Recovery flow sheet proposes to solidify the liquid effluent using a grout formula that was developed and tested for the former Waste Solidification Building process. This Portland cement based mixture results in a high pH (~13) leachate from the solidified waste form which is not acceptable to WIPP in the large quantities expected from production at SRPPF. Various cementitious materials were previously evaluated as alternative grout formulations to Portland cement and a MgO-based mix was identified as a promising alternative. A magnesium oxysulfate (MOS) cement formulation comprised of reactive magnesium oxide (MgO), anhydrous magnesium sulfate (MgSO 4 ), and sand, as a non-reactive heat sink provided good mixability, similar density to the original Portland-cement based mix, and a leachate pH of 9.4, within the assumed WIPP brine pH range. However, the MOS formulation exhibited an appreciable amount of heat generation, which resulted in premature setting of a large-scale test.

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DOE OSTI2015

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

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

composition constraint↗

DOE OSTI2017

Integration of National Laboratory and Low-Activity Waste Pre-Treatment System Technology Service Providers - 16435

The National Laboratories are a critical partner and provide expertise in numerous aspects of the successful execution of the Direct-Feed Low Activity Waste Program. The National Laboratories are maturing the technologies of the Low-Activity Waste Pre-Treatment System (LAWPS) consistent with DOE Order 413.3B “Program and Project Management for the Acquisition of Capital Assets” expectations. The National Laboratories continue to mature waste forms, i.e. glass and secondary waste grout, for formulations and predictions of long-term performance as inputs to performance assessments. The working processes with the National Laboratories have been developed in procurements, communications, and reporting to support the necessary delivery-based technology support. The relationship continues to evolve from planning and technology development to support of ongoing operations and integration of multiple highly coordinated facilities.

Low-Activity Waste↗

DOE OSTI2013

GeoMelt{sup R} ICV{sup TM} Treatment of Sellafield Pond Solids Waste - 13414

Kurion, Inc., in partnership with AMEC Ltd., is demonstrating its GeoMelt{sup R} In-Container Vitrification (ICV){sup TM} Technology to Sellafield Ltd. (SL). SL is evaluating the proposition of directly converting a container (skip/box/drum) of raw solid ILW into an immobilized waste form using thermal treatment, such that the resulting product is suitable for interim storage at Sellafield and subsequent disposal at a future Geological Disposal Facility. Potential SL feed streams include sludges, ion-exchange media, sand, plutonium contaminated material, concrete, uranium, fuel cladding, soils, metals, and decommissioning wastes. The solid wastes have significant proportions of metallic constituents in the form of containers, plant equipment, structural material and swarf arising from the nuclear operations at Sellafield. GeoMelt's proprietary ICV process was selected for demonstration, with the focus being high and reactive metal wastes arising from solid ILW material. A composite surrogate recipe was used to demonstrate the technology towards treating waste forms of diverse types and shapes, as well as those considered difficult to process; all the while requiring few (if any) pre-treatment activities. Key strategic objectives, along with their success criterion, were established by SL for this testing, namely: 1. Passivate and stabilize the raw waste simulant, as demonstrated by the entire quantity of material being vitrified, 2. Immobilize the radiological and chemo-toxic species, as demonstrated via indicative mass balance using elemental analyses from an array of samples, 3. Production of an inert and durable product as evidenced by transformation of reactive metals to their inert oxide forms and satisfactory leachability results using PCT testing. Two tests were performed using the GeoMelt Demonstration Unit located at AMEC's Birchwood Park Facilities in the UK. Post-melt examination of the first test indicated some of the waste simulant had not fully processed, due to insufficient processing time and melt temperature. A second test, incorporating operational experience from the first test, was performed and resulted in all of the 138 kg of feed material being treated. The waste simulant portion, at 41 kg, constituted 30 wt% of the total feed mass, with over 90% of this being made up of various reactive and non-reactive metals. The 95 liters of staged material was volume reduced to 41 liters, providing a 57% overall feed to product volume reduction in a fully passivated two-phase glass/metal product. The GeoMelt equipment operated as designed, vitrifying the entire batch of waste simulant. Post-melt analytical testing verified that 91-99+% of the radiological tracer metals were uniformly distributed within the glass/cast refractory/metal product, and the remaining fraction was captured in the offgas filtration systems. PCT testing of the glass and inner refractory liner showed leachability results that outperform the DOE regulatory limit of 2 g/m{sup 2} for the radiological species of interest (Sr, Ru, Cs, Eu, Re), and by more than an order of magnitude better for standard reference analytes (B, Na, Si). (authors)

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DOE OSTI2019

Annual Status Report (FY 2018): Performance Assessment for the Environmental Restoration Disposal Facility

DOE O 435.1 and DOE M 435.1-1 require that a determination of continued adequacy of the performance assessment (PA), composite analysis, and disposal authorization statement be made on an annual basis, and it must consider the results of data collection and analysis from research, field studies, and monitoring. Beginning in 1996, the Environmental Restoration Disposal Facility (ERDF) started accepting low-level radioactive, hazardous, and mixed wastes that were generated during cleanup activities at the Hanford Site. ERDF is composed of a series of cells or disposal areas and can accommodate future design expansions as needed. Currently, there are 10 cells. During this reporting period (fiscal year 2018, which extends from October 1, 2017, through September 30, 2018), approximately 193,358 U.S. tons of waste was disposed at ERDF. From ERDF inception through September 30, 2018, approximately 18.3 million U.S. tons of waste had been disposed at ERDF, which equates to consumption of approximately 87% of the disposal volume. As a condition of the Disoposal Authorization Statement, disposal operations within ERDF must be in accordance with the waste acceptance criteria (WAC) that provides specific radionuclide disposal limits, waste form restrictions, and descriptions of acceptable waste packages in compliance with the requirements of DOE 435-1.1. The ERDF WAC stipulates that waste destined for disposal at ERDF be controlled based on source, physical form, and contaminant concentration and activity levels. Concentration limits included in the ERDF WAC were extracted from the ERDF PA. A sum of fraction analysis of of the ERDF inventory against the threshold concentration limits from inception through fiscal year 2018 produces a result <1. This result provides confirmation of compliance with PA imposed radionuclide limits.

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DOE OSTI2013

Treatment of Bottled Liquid Waste During Remediation of the Hanford 618-10 Burial Ground - 13001

A problematic waste form encountered during remediation of the Hanford Site 618-10 burial ground consists of bottled aqueous waste potentially contaminated with regulated metals. The liquid waste requires stabilization prior to landfill disposal. Prior remediation activities at other Hanford burial grounds resulted in a standard process for sampling and analyzing liquid waste using manual methods. Due to the highly dispersible characteristics of alpha contamination, and the potential for shock sensitive chemicals, a different method for bottle processing was needed for the 618-10 burial ground. Discussions with the United States Department of Energy (DOE) and United States Environmental Protection Agency (EPA) led to development of a modified approach. The modified approach involves treatment of liquid waste in bottles, up to one gallon per bottle, in a tray or box within the excavation of the remediation site. Bottles are placed in the box, covered with soil and fixative, crushed, and mixed with a Portland cement grout. The potential hazards of the liquid waste preclude sampling prior to treatment. Post treatment verification sampling is performed to demonstrate compliance with land disposal restrictions and disposal facility acceptance criteria. (authors)

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DOE OSTI2012

The Windscale Advanced Gas Cooled Reactor (WAGR) Decommissioning Project A Close Out Report for WAGR Decommissioning Campaigns 1 to 10 - 12474

The reactor core of the Windscale Advanced Gas-Cooled Reactor (WAGR) has been dismantled as part of an ongoing decommissioning project. The WAGR operated until 1981 as a development reactor for the British Commercial Advanced Gas cooled Reactor (CAGR) power programme. Decommissioning began in 1982 with the removal of fuel from the reactor core which was completed in 1983. Subsequently, a significant amount of engineering work was carried out, including removal of equipment external to the reactor and initial manual dismantling operations at the top of the reactor, in preparation for the removal of the reactor core itself. Modification of the facility structure and construction of the waste packaging plant served to provide a waste route for the reactor components. The reactor core was dismantled on a 'top-down' basis in a series of 'campaigns' related to discrete reactor components. This report describes the facility, the modifications undertaken to facilitate its decommissioning and the strategies employed to recognise the successful decommissioning of the reactor. Early decommissioning tasks at the top of the reactor were undertaken manually but the main of the decommissioning tasks were carried remotely, with deployment systems comprising of little more than crane like devices, intelligently interfaced into the existing structure. The tooling deployed from the 3 tonne capacity (3te) hoist consisted either purely mechanical devices or those being electrically controlled from a 'push-button' panel positioned at the operator control stations, there was no degree of autonomy in the 3te hoist or any of the tools deployed from it. Whilst the ATC was able to provide some tele-robotic capabilities these were very limited and required a good degree of driver input which due to the operating philosophy at WAGR was not utilised. The WAGR box proved a successful waste package, adaptable through the use of waste box furniture specific to the waste-forms generated throughout the various decommissioning campaigns. The use of low force compaction for insulation and soft wastes provided a simple, robust and cost effective solution as did the direct encapsulation of LLW steel components in the later stages of reactor decommissioning. Progress through early campaigns was good, often bettering the baseline schedule, especially when undertaking the repetitive tasks seen during Neutron Shield and Graphite Core decommissioning, once the operators had become experienced with the equipment, though delays became more pronounced, mainly as a result of increased failures due to the age and maintainability of the RDM and associated equipment. Extensive delays came about as a result of the unsupported insulation falling away from the pressure vessel during removal and the inability of the ventilation system to manage the sub micron particulate generated during IPOPI cutting operations, though the in house development of revised and new methodologies ultimately led to the successful completion of PV and I removal. In a programme spanning over 12 years, the decommissioning of the reactor pressure vessel and core led to the production 110 ILW and 75 LLW WAGR boxes, with 20 LLW ISO freight containers of primary reactor wastes, resulting in an overall packaged volume of approximately 2500 cubic metres containing the estimated 460 cubic metres of the reactor structure. (authors)

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DOE OSTI2016

The Evolution of Chemical Waste Treatment in the Nuclear Industry Simple Solutions for Complex Problems - 16694

The U.S. nuclear waste treatment industry was born in the 1980s and early 1990s with the end of the Cold War, the peak of commercial nuclear power, and after implementation of federal regulations that drove new disposal requirements for radioactive waste generators. New regulations resulted in soaring costs for radioactive waste disposal, and thus an industry was born that focused on recycling, volume reduction, and waste form improvement. New treatment facilities and technologies were started that primarily dealt with low-level radioactive wastes; however, there was little to no treatment capacity for wastes commingled with hazardous chemicals. Large volumes of 'mixed wastes', of which many became known as 'orphan wastes', were being stored on U.S. Department of Energy (DOE) sites with no path for disposal and thus posed a risk to workers, the public, and the environment. Perma-fix Environmental Services, Inc. (Perma-Fix), which started in the hazardous waste industry in 1990 and had developed facilities and technologies to treat hazardous wastes, evolved in the late 90's to meet this mixed waste market opportunity. By adapting our technologies and collaborating with waste generators to understand their needs, we successfully developed treatment and disposition solutions for some of the most challenging mixed wastes in the U.S. By establishing a pathway for treatment and disposal of orphan wastes, Perma-Fix played an important role in the safe treatment and disposition of orphan wastes, which facilitated closure of DOE sites such as Rocky Flats, Mound, and Fernald. We found that understanding chemistry and having experience in managing the hazardous constituents of mixed wastes allowed us to apply solutions that were not overly complex and expensive. Further, by adapting proven hazardous waste treatment solutions to wastes commingled with contaminants such as polychlorinated biphenyls (PCBs), mercury, and organics we were able to safely manage even more problematic mixed wastes. During the 1980's, commercial low-level radioactive waste treatment and disposal solutions were being developed. At the same time, the federal government began dealing with increasing stakeholder demands for permanent solutions of more complex radioactive wastes such as high-level waste (HLW) and transuranic (TRU) wastes. HLW and TRU waste disposal became a contentious political issue between the federal government, regulatory agencies, and the public. As a result, DOE was mandated to develop treatment and disposal methodologies to address these wastes. DOE's Waste Isolation Pilot Plant (WIPP) facility solved most issues with disposition of TRU wastes; however, for HLW (for example tank wastes at Hanford, Savannah River, and Idaho) the complexity of the wastes and the fact that government entities were developing political solutions for waste treatment resulted in a variety of complex and expensive solutions. Decades later we still do not have a permanent solution for HLW. Whether some simple commercial approaches are possible, what regulatory changes should occur, and what the implications may be if we do not change our current path, will be discussed. After two decades, we see the low-level nuclear waste industry maturing and replicating the path of the hazardous waste industry. When new hazardous waste federal regulations were enacted in the 1970's and 1980's, the hazardous waste industry began to boom with new company start-ups and construction of numerous treatment and disposal facilities. Initially, this was a good business to be in, managing dangerous wastes in an environmentally safe and responsible manner, and it was profitable. However, over-capacity and fierce competition drove down pricing, and companies failed. This ultimately led to a major consolidation within the hazardous waste industry. Similarly and more recently in the nuclear industry, over-capacity and competition has driven down profitability, companies have failed, and there is a major consolidation occurring within sectors of the industry. For consolidation to be successful and healthy for either industry, a leader must emerge to set the direction for the others and establish a sustainable path that is acceptable for waste generators, stakeholders, and investors. This was the case for the hazardous waste industry. This paper will examine the parallels between the hazardous and nuclear waste industries as they have matured and evolved, and what discuss the long term implications of consolidation within the nuclear industry. (authors)

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DOE OSTI2017

Non-pertechnetate Technetium Sensor Research and Development

Several significant uncertainties remain regarding the understanding and modeling of the fate and speciation of technicium-99 ( 99 Tc) in Hanford waste tanks, glass, and low-temperature waste forms. A significant (2% to 25%) fraction of the 99Tc in the water-soluble portion of the tank waste may be present as one or more non pertechnetate species that have not been identified and to date, cannot be effectively separated from the wastes. This task will provide a sensor specifically tuned to detect the Tc(I)-carbonyl species believed to constitute the main fraction of the non-pertechnetate form of technetium. By direct measurement of the non-pertechnetate species, such a sensor will help reduce the uncertainties in the modeling of the fate and speciation of 99 Tc in Hanford tanks and waste forms. This report summarizes work performed in FY2016 that was sponsored by the Department of Energy’s Office of Environmental Management and demonstrates the protocol for using fluorescent Tc(I)-tricarbonyl complex as a means to detect the non-pertechnetate species within tank waste solutions. The protocol was optimized with respect to ligand concentration, solvent choice, reaction temperature and time. This work culminated in the quantitation of Tc(I)-tricarbonyl within a waste simulant, using a standard addition method for measurement. This report also summarizes the synthesis and high-yield preparation of the low-valence technetium species, [Tc(CO) 3 (H 2 O) 3 ] + , which will be used as the technetium standard material for the demonstration of the non-pertechnetate species in actual wastes.

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DOE OSTI2012

A Preliminary Performance Assessment for Salt Disposal of High-Level Nuclear Waste - 12173

A salt repository is one of the four geologic media currently under study by the U.S. DOE Office of Nuclear Energy to support the development of a long-term strategy for geologic disposal of commercial used nuclear fuel (UNF) and high-level radioactive waste (HLW). The immediate goal of the generic salt repository study is to develop the necessary modeling tools to evaluate and improve the understanding of the repository system response and processes relevant to long-term disposal of UNF and HLW in a salt formation. The current phase of this study considers representative geologic settings and features adopted from previous studies for salt repository sites. For the reference scenario, the brine flow rates in the repository and underlying interbeds are very low, and transport of radionuclides in the transport pathways is dominated by diffusion and greatly retarded by sorption on the interbed filling materials. I-129 is the dominant annual dose contributor at the hypothetical accessible environment, but the calculated mean annual dose is negligibly small. For the human intrusion (or disturbed) scenario, the mean mass release rate and mean annual dose histories are very different from those for the reference scenario. Actinides including Pu-239, Pu-242 and Np-237 are major annual dose contributors, and the calculated peak mean annual dose is acceptably low. A performance assessment model for a generic salt repository has been developed incorporating, where applicable, representative geologic settings and features adopted from literature data for salt repository sites. The conceptual model and scenario for radionuclide release and transport from a salt repository were developed utilizing literature data. The salt GDS model was developed in a probabilistic analysis framework. The preliminary performance analysis for demonstration of model capability is for an isothermal condition at the ambient temperature for the near field. The capability demonstration emphasizes key attributes of a salt repository that are potentially important to the long-term safe disposal of UNF and HLW. The analysis presents and discusses the results showing repository responses to different radionuclide release scenarios (undisturbed and human intrusion). For the reference (or nominal or undisturbed) scenario, the brine flow rates in the repository and underlying interbeds are very low, and transport of radionuclides in the transport pathways is dominated by diffusion and greatly retarded by sorption on the interbed filling materials. I-129 (non-sorbing and unlimited solubility with a very long half-life) is the dominant annual dose contributor at the hypothetical accessible environment, but the calculated mean annual dose is negligibly small that there is no meaningful consequence for the repository performance. For the human intrusion (or disturbed) scenario analysis, the mean mass release rate and mean annual dose histories are very different from those for the reference scenario analysis. Compared to the reference scenario, the relative annual dose contributions by soluble, non-sorbing fission products, particularly I-129, are much lower than by actinides including Pu-239, Pu-242 and Np-237. The lower relative mean annual dose contributions by the fission product radionuclides are due to their lower total inventory available for release (i.e., up to five affected waste packages), and the higher mean annual doses by the actinides are the outcome of the direct release of the radionuclides into the overlying aquifer having high water flow rates, thereby resulting in an early arrival of higher concentrations of the radionuclides at the biosphere drinking water well prior to their significant decay. The salt GDS model analysis has also identified the following future recommendations and/or knowledge gaps to improve and enhance the confidence of the future repository performance analysis. - Repository thermal loading by UNF and HLW, and the effect on the engineered barrier and near-field performance. - Closure and consolidation of salt rocks by creep deformation under the influence of thermal perturbation, and the effect on the engineered barrier and near-field performance. - Brine migration and radionuclide transport under the influence of thermal perturbation in generic salt repository environment, and the effect on the engineered barrier and near-field performance and far-field performance. - Near-field geochemistry and radionuclide mobility in generic salt repository environment (high ionic strength brines, elevated temperatures and chemically reducing condition). - Degradation of engineer barrier components (waste package, waste canister, waste forms, etc.) in a generic salt repository environment (high ionic strength brines, elevated temperatures and chemically reducing condition). - Waste stream types and inventory estimates, particularly for reprocessing high-level waste. (authors)

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DOE OSTI2014

Non-Pertechnetate Technetium Sensor Research and Development

There remain several significant uncertainties in the understanding and modeling of the fate and speciation of 99 Tc in Hanford waste tanks, glass, and low-temperature waste forms. A significant (2% to 25%) fraction of the 99 Tc in the water-soluble portion of the tank waste may be present as a non-pertechnetate species that has not been identified and, based on experimentation to date, cannot be effectively separated from the wastes. This task will provide a sensor specifically tuned to detect the Tc(I)-carbonyl species believed to constitute the main fraction of the non-pertechnetate form of technetium. By direct measurement of the non-pertechnetate species, such a sensor will help reduce the uncertainties in the modeling of the fate and speciation of 99 Tc in Hanford tanks and waste forms. This report summarizes work done in FY 2014 exploring the chemistry of a low-valence technetium species, [Tc(CO) 3 (H 2 O) 3 ] + , a compound of interest due to its implication in the speciation of alkaline-soluble technetium in several Hanford tank waste supernatants. Progress made in FY 2014 was sponsored by the Department of Energy’s Office of Environmental Management and is summarized in this report.

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DOE OSTI2013

Area 3 and Area 5 Radioactive Waste Management Sites at the Nevada National Security Site, Nye County, Nevada: Review of the Performance Assessments and Composite Analyses (2012 Annual Summary Report)

The Maintenance Plan for the Performance Assessments and Composite Analyses for the Area 3 and Area 5 Radioactive Waste Management Sites at the Nevada Test Site (National Security Technologies, LLC 2007a) requires an annual review to assess the adequacy of the performance assessments (PAs) and composite analyses (CAs), with the results submitted to the U.S. Department of Energy (DOE) Office of Environmental Management. The Disposal Authorization Statements for the Area 3 and Area 5 Radioactive Waste Management Sites (RWMSs) also require that such reviews be made and that secondary or minor unresolved issues be tracked and addressed as part of the maintenance plan (DOE 1999a, 2000). The U.S. Department of Energy, National Nuclear Security Administration Nevada Site Office performed an annual review of the Area 3 and Area 5 RWMS PAs and CAs for fiscal year (FY) 2012. This annual summary report presents data and conclusions from the FY 2012 review, and determines the adequacy of the PAs and CAs. Operational factors (e.g., waste forms and containers, facility design, and waste receipts), closure plans, monitoring results, and research and development (RD) activities were reviewed to determine the adequacy of the PAs. Likewise, the environmental restoration activities at the Nevada National Security Site (NNSS) relevant to the sources of residual radioactive material that are considered in the CAs, the land-use planning, and the results of the environmental monitoring and RD activities were reviewed to determine the adequacy of the CAs. Important developments in FY 2012 include the following: Release of a special analysis for the Area 3 RWMS assessing the continuing validity of the PA and CA; Development of a new Area 5 RWMS closure inventory estimate based on disposals through FY 2012; Evaluation of new or revised waste streams by special analysis; and Development of version 4.114 of the Area 5 RWMS GoldSim PA model. The Area 3 RWMS has been in inactive status since July 1, 2006, with the last shipment received in April 2006. The FY 2012 review of operations, facility design, closure plans, monitoring results, and RD results for the Area 3 RWMS indicates no changes that would impact PA validity. A special analysis using the Area 3 RWMS v2.102 GoldSim PA model was prepared to update the PA results for the Area 3 RWMS in FY 2012. The special analysis concludes that all performance objectives can be met and the Area 3 RWMS PA remains valid. There is no need to the revise the Area 3 RWMS PA. Review of Area 5 RWMS operations, design, closure plans, monitoring results, and RD activities indicates no significant changes other than an increase in the inventory disposed. The FY 2012 PA results, generated with the Area 5 RWMS v4.114 GoldSim PA model, indicate that there continues to be a reasonable expectation of meeting all performance objectives. The results and conclusions of the Area 5 RWMS PA are judged valid, and there is no need to the revise the PA. A review of changes potentially impacting the CAs indicates that no significant changes occurred in FY 2012. The continuing adequacy of the CAs was evaluated with the new models, and no significant changes that would alter CA results or conclusions were found. The revision of the Area 3 RWMS CA, which will include the Underground Test Area source term (Corrective Action Unit [CAU] 97), is scheduled for FY 2024, following the completion of the Yucca Flat CAU 97 Corrective Action Decision Document/Corrective Action Plan in FY 2016. Inclusion of the Frenchman Flat CAU 98 results in the Area 5 RWMS CA is scheduled for FY 2016, pending the completion of the CAU 98 closure report in FY 2015. Near-term RD efforts will focus on continuing development of the Area 3 and Area 5 RWMS GoldSim PA/CA and inventory models.

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DOE OSTI2014

2013 Annual Summary Report for the Area 3 and Area 5 Radioactive Waste Management Sites at the Nevada National Security Site, Nye County, Nevada: Review of the Performance Assessments and Composite Analyses

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