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All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Statistical Tools and Lines of Evidence Used to Differentiate FUSRAP/Non-FURAP Related Material at an Inactive Landfill - 17553

The Middlesex Municipal Landfill (MML) Site is a former municipal landfill in Middlesex, New Jersey that operated between approximately 1940 and 1974; it is currently part of FUSRAP. Because of the lack of waste management regulations at the time, unregulated disposal of municipal and industrial waste occurred, including the acceptance of radioactively contaminated soils from the Middlesex Sampling Plant (MSP) that were used as cover material on the northwest side of the landfill. The radioactive MSP contaminated soils are described as 'FUSRAP-related materials.' In some areas of the landfill, FUSRAP-related material exists comingled with other non-FUSRAP radioactive waste (such as an antique radio-luminescent marker and concrete construction material with elevated gamma readings). However, in other areas of the landfill (specifically the street's right-of-way, called Pershing Avenue, which borders the landfill), non-FUSRAP material is isolated. Several lines of evidence and statistical tools were used to differentiate FUSRAP-related materials from non-FUSRAP-related material, including Z-Score analysis, Tukey-Kramer Honestly Significant Difference test, box-and-whisker plots, Wilcoxon Signed Rank test, and Principal Component Analysis. (authors)

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

Development of the Waste Acceptance Criteria for the Idaho National Laboratory Remote-Handled Low-Level Waste Disposal Facility - 19143

The Idaho National Laboratory (INL) Remote-Handled Low-Level Waste (RH LLW) Disposal Facility has been constructed, has received operational approval from the Department of Energy (DOE), and is expected to receive its first waste shipment in FY 2019. The RH LLW Disposal Facility will support the needs of INL and the Naval Reactors Facility (NRF). RH LLW destined for the facility includes waste currently in storage from previous programs and projects conducted at INL; newly generated waste from programs and projects conducted at INL; and waste from NRF spent fuel processing activities. This new facility supports RH LLW disposal needs for the ongoing mission of INL and NRF. Waste to be disposed of at the RH LLW Disposal Facility may exhibit dose rates that range from 0.002 Sv/hr to 600 Sv/hr. Waste types include resins and activated metals as well as surface-contaminated debris generated from existing and new missions. The total number of canisters currently planned for disposal at the RH LLW Disposal Facility is 939 with the possibility of future expansion. The projected waste volume to be disposed of is 3,100 m{sup 3}. The RH LLW Disposal Facility will receive a wide variety of RH LLW and is in preparations to receive its first canisters in FY 2019. Waste acceptance criteria (WAC) and the radioactive waste management basis have been developed and implemented in accordance with the applicable DOE Order. In order to ensure that DOE requirements and guidance were appropriately implemented in the RH LLW Disposal Facility WAC, the project worked closely with the DOE Low Level Waste Disposal Facility Federal Review Group (LFRG) representatives and followed the applicable DOE standard. This paper provides an overview of the development of the WAC, including waste and waste generator certification requirements, for the RH LLW Disposal Facility. The paper illustrates the importance of maintaining strong ties to the development and revision of the RH LLW Disposal Facility Performance Assessment (PA), Documented Safety Analysis, Facility Procedures, Readiness Reviews, DOE Order, and applicable Technical Standards. The paper also provides an overview of the RH LLW Disposal Facility, WAC generator certification requirements, distinctive challenges to acceptance of legacy waste, and the unique approach the RH LLW Disposal Facility has taken to ensure compliance with PA driven limits. (authors)

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DOE OSTI2020

Update on Development of a U.S. Rail Transport Capability for Spent Nuclear Fuel and High-Level Waste - 20466

This paper provides an overview of the progress to date, and discussion of the path forward, related to designing, fabricating and testing prototype railcars that will comply with the safety standard S-2043. This standard was developed by the Association of American Railroads (AAR) specifically for railcars used to transport High-Level Radioactive Material (HLRM). AAR defines the term HLRM to include both spent nuclear fuel (SNF) and high-level radioactive waste (HLW). DOE is in the process of developing and testing prototype railcars that will satisfy Standard S-2043. This is a technical paper that does not take into account contractual limitations or obligations under the Standard Contract for Disposal of Spent Nuclear Fuel and/or High-Level Radioactive Waste (Standard Contract) (10 CFR Part 961). For example, under the provisions of the Standard Contract, spent nuclear fuel in multi-assembly canisters is not an acceptable waste form, absent a mutually agreed to contract amendment. To the extent discussions or recommendations in this paper conflict with the provisions of the Standard Contract, the Standard Contract governs the obligations of the parties, and this paper in no manner supersedes, overrides, or amends the Standard Contract. This paper reflects technical work which could support future decision making by the Department of Energy (DOE or Department). No inferences should be drawn from this paper regarding future actions by DOE, which are limited both by the terms of the Standard Contract and a lack of Congressional appropriations for the Department to fulfill its obligations under the Nuclear Waste Policy Act including licensing and construction of a spent nuclear fuel repository. (authors)

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

Modeling the Long-Term Behavior of Cementitious Barriers - 18188

Cementitious materials are commonly included as backfill/barrier materials in geological repositories for radioactive waste. An understanding of the long-term behavior of cement-based barriers is therefore required to provide confidence in their safety functions within an environmental safety case. Geochemical modelling has been undertaken on several different radioactive waste repository concepts (and their natural/industrial analogues), to explore the nature of cement degradation over timescales relevant to postclosure safety (many tens of thousands of years to c. 1 Ma). In radionuclide transport calculations undertaken as part of performance assessments, it is typically considered that the properties of cementitious engineered barrier materials change as they degrade, potentially resulting in an increase in porosity hydraulic conductivity, and effective diffusion coefficients. However, reactive transport modelling (supported by some analogue data) suggests that, in some repository systems, a degree of clogging of the cement pore space could occur and that some advective flow paths (cracks in the barriers, fractures in the host rock) could become (to some extent) infilled by cement and rock alteration products. This could result in a reduction in water flow and contaminant transport in the vicinity of the cement-rock interface, but some uncertainties remain (especially process couplings and the nature of water flow in complex fracture networks that are becoming clogged). In addition, the armoring of fracture surfaces could result in decreased potential for diffusion of solutes into the rock matrix. It is envisaged that ongoing work will continue to improve reactive-transport model capabilities, such that the implications cement degradation has for contaminant transport can be predicted with greater confidence. (authors)

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DOE OSTI2015

Evaluation of waste arising from future nuclear fuel cycle - 15197

A comprehensive study was recently completed at the request of the US Department of Energy Office of Nuclear Energy (DOE-NE) to evaluate and screen nuclear fuel cycles. The final report was issued in October 2014. Uranium- and thorium-based fuel cycles were evaluated using both fast and thermal spectrum reactors. Once-through, limited-recycle, and continuous-recycle cases were considered. This study used nine evaluation criteria to identify promising fuel cycles. Nuclear waste management was one of the nine evaluation criteria. The waste generation criterion from this study is discussed herein. The fundamental characteristics of nuclear fuel cycles were used to create a comprehensive set of fuel cycle options. These options were categorized into 40 Evaluation Groups based on similarities in their physics-based performance. These Evaluation Groups, listed below, comprehensively represent all fuel cycle options: - 8 once-through Evaluation Groups: 5 uranium-based fuel cycles and 3 thorium-based fuel cycles - 10 limited-recycle Evaluation Groups: 6 uranium-based fuel cycles and 4 thorium-based fuel cycles, and - 22 continuous-recycle Evaluation Groups: 14 uranium-based fuel cycles and 8 thorium-based fuel cycles. The study focused on the quantity and characteristics of radioactive wastes generated by the different fuel cycles, examining 5 metrics characterizing waste generation: - mass of spent nuclear fuel (SNF) plus high level waste (HLW) disposed per energy generated, - activity of SNF + HLW (at 100 years) per energy generated, - activity of SNF + HLW (at 100,000 years) per energy generated, - mass of depleted uranium (DU) + recovered uranium (RU) + recovered thorium (RTh) disposed per energy generated, and - volume of low level radioactive waste (LLW) per energy generated. (authors)

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DOE OSTI2015

Technetium incorporation in glass for the Hanford Tank Waste treatment and immobilization plant - 15066

A priority of the United States Department of Energy (U.S. DOE) is to dispose of nuclear wastes accumulated in 177 underground tanks at the Hanford Nuclear Reservation in eastern Washington State. These nuclear wastes date from the Manhattan Project of World War II and from plutonium production during the Cold War. The DOE plans to separate high-level radioactive wastes from low activity wastes and to treat each of the waste streams by vitrification (immobilization of the nuclides in glass) for disposal. The immobilized low-activity waste will be disposed of here at Hanford and the immobilized high-level waste at the national geologic repository. Included in the inventory of highly radioactive wastes is large volumes of {sup 99}Tc (∼9 x 10 E2 TBq or ∼2.5 x 10{sup 4} Ci or ∼1500 kg). A problem facing safe disposal of Tc-bearing wastes is the processing of waste feed into in a chemically durable waste form. Technetium incorporates poorly into silicate glass in traditional glass melting. It readily evaporates during melting of glass feeds and out of the molten glass, leading to a spectrum of high-to-low retention (ca. 20 to 80%) in the cooled glass product. DOE-ORP currently has a program at Pacific Northwest National Laboratory (PNNL), in the Department of Materials Science and Engineering at Rutgers University and in the School of Mechanical and Materials Engineering at Washington State University that seeks to understand aspects of Tc retention by means of studying Tc partitioning, molten salt formation, volatilization pathways, and cold cap chemistry. Another problem involves the stability of Tc in glass in both the national geologic repository and on-site disposal after it has been immobilized. The major environmental concern with {sup 99}Tc is its high mobility in addition to a long half-life (2.1 x 10{sup 5} yrs). The pertechnetate ion (TcO{sup 4-}) is highly soluble in water and does not adsorb well onto the surface of minerals and so migrates nearly at the same velocity as groundwater. Long-term corrosion of glass waste forms is an area of current interest to the DOE, but attention to the release of Tc from glass has been little explored. It is expected that the release of Tc from glass should be highly dependent on the local glass structure as well as the chemistry of the surrounding environment, including groundwater pH. Though the speciation of Tc in glass has been previously studied, and the Tc species present in waste glass have been previously reported, environmental Tc release mechanisms are poorly understood. The recent advances in Tc chemistry that have given rise to an understanding of incorporation in the glass giving rise to significantly higher single-pass retention during vitrification are presented. Additionally, possible changes to the baseline flowsheet that allow for relatively minor volumes of Tc reporting to secondary waste treatment will be discussed. (authors)

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DOE OSTI2019

Investigation of Thermolytic Production of Hydrogen Applicable to Savannah River Site High Level Waste - 19443

Several experiments have been performed to measure the generation rates of hydrogen gas due to reaction (thermolysis) of organic compounds in both simulated and radioactive waste. Initial measurements of hydrogen generation rates at 100 deg. C were made to downselect molecules and functional groups that produce significant amounts of hydrogen for subsequent kinetic determination. A statistically-designed test matrix was employed to quantitatively determine chemical conditions of highest reactivity at 100 deg. C. Additional experiments were performed at 100 deg. C to determine the impact of specific chemical components on hydrogen generation rates for relevant organic compounds. A semi-empirical model for thermolytic hydrogen generation rate from glycolate at 100 deg. C was developed and compared to results from radioactive waste and literature reports. (authors)

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DOE OSTI2016

Best Practice on Facility Characterisation from a Material and Waste End- State Perspective - 16041

Radiological characterisation plays a significant role in the process of decommissioning of shut-down nuclear facilities in order to ensure the protection of the environment and radiation safety. At all stages of a decommissioning programme or project, adequate radiological characterisation is of crucial importance, not least from a material and waste perspective. The radiological characterisation is a key element for planning, controlling and optimising decommissioning and dismantling activities. Experience has shown that data and information from the operation of a facility can - supplemented by recently collected and analysed data and information - be of crucial importance for decisions on waste management and for characterisation of radioactive waste. Once the dismantling has been done, some information may be hard, costly or even impossible to obtain later in the waste management process. This was the reason why the Working Party on Decommissioning and Dismantling (WPDD) of the OECD Nuclear Energy Agency (NEA) decided in late 2013 to extend the mandate of the Task Group on Radiological Characterisation and Decommissioning (TGRCD) for a second phase focusing on nuclear facility characterisation from a waste and material end-state perspective whereas the first phase focused on overall strategies of radiological characterisation. This paper gives an overview of the activities and findings within both phases up to now. (authors)

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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 OSTI2019

Treatment of Uranium Catalyst Waste for High Volume Reduction of the Final Waste to be Disposed - 19023

A large volume of problematic radioactive waste of spent uranium catalyst, approximately 7,000 drums was generated by a chemical company in Korea. The spent uranium catalyst has since been stored onsite awaiting a management strategy that is to minimize the volume/cost requirements for final disposal of the uranium-bearing waste. This work developed for the first time a process to greatly reduce the waste volume to be disposed while the secondary wastes generated from the process being released to the environment are controlled to meet the discharge criteria in Korea. (authors)

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DOE OSTI2015

Feasibility studies for the design of disposal vaults for multi-purpose containers (MPCs): construction, operation, ventilation and management of heat - 15175

The Nuclear Decommissioning Authority (NDA) subsidiary, Radioactive Waste Management Ltd (RWM) is responsible for the safe implementation of a Geological Disposal Facility (GDF) for radioactive waste in the UK. As part of their generic programme to underpin the planned designs for a GDF, work is ongoing to explore a range of possible geological disposal concepts. This work developed GDF concept designs for Multi-Purpose Containers (MPC) disposal in the generic host rock types considered: higher strength (crystalline), lower strength sedimentary (sedimentary / clay formations) and evaporite rocks. This paper is specific to concepts for the disposal of MPC for Pressurised Water Reactor (PWR) spent nuclear fuel (SF). The key outcomes of the feasibility studies to improve the understanding of the MPC disposal concept and the specific challenges to construct, operate, ventilate, manage heat and backfill associated with this concept are summarised in this paper. The management of the high loading of radiogenic heat (from 12 PWR SF assemblies in MPCs compared to 4 in 'standard' disposal containers) during an extended open phase (delayed backfilling of the disposal area) of up to 300 years was investigated. The concept designs developed demonstrate it is feasible to construct, operate and ventilate disposal vaults to accommodate open 'storage' periods required for transitional thermal decay of high heat generating waste in higher strength and lower strength sedimentary host rock types considered by RWM. This work focused on developing two contrasting disposal concepts which are applicable to both higher strength and lower strength rock environments; both of which illustrate how vault temperature, stability and groundwater inflows can be controlled to delay backfilling and allow MPC package thermal output to decrease. MPC disposal in evaporite rock does not require a cooling period due to its high thermal conductivity and favourable performance at higher temperatures; therefore the concept developed for this rock includes immediate backfilling of disposal tunnels. A clear understanding of the interactions and limitations associated with heat management during the open period was developed by investigation of a 'Heating Ventilation and Air-Conditioning System' (HVAC). It was concluded that removal of heat from the disposal tunnels / vaults is feasible, but introduces significant operational complexity. (authors)

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DOE OSTI2015

CEA decommissioning and clean-up activities legacy waste management and lessons learned - 15506

The CEA has generated radioactive waste in both liquid and solid form for more than 50 years within the scope of its R and D, clean-up and dismantling programs. The vast majority of the waste produced is transferred to disposal centers in approved waste packages. Nonetheless, a large quantity of legacy waste still remains in ageing facilities in which the storage conditions are sometimes no longer satisfactory with respect to the safety requirements that are constantly evolving. As part of a broad clean-up and dismantling process focusing on these ageing facilities managed by the CEA, the top priority is to shut down nuclear activities on two symbolic sites: Grenoble and Fontenay-aux-Roses, which are both located in a dense urban environment. After about ten years of dismantling, the nuclear facilities on the Grenoble site are on their way to be completely decommissioned in the regulatory sense of the term. Furthermore, the radwaste issue on this site has raised no specific problems, though it did require developing innovative and sometimes costly waste outlets. The older Fontenay-aux-Roses site is where the first experimental research on minor actinide separation took place. The facilities on this site have reached a stage in dismantling which necessitates the recovery of irradiating legacy waste containing alpha, beta and gamma emitters. The transfer of this waste to interim storage centers more than 600 kilometers away is a decisive factor in reducing the site's nuclear source term and in ensuring the satisfactory progress of the dismantling operations. This paper details the issues raised by the recovery of legacy waste, which requires greater characterization to better understand its radiological and chemical status. This waste also requires better characterization to meet the storage requirements of new sites designed for storage periods exceeding 50 years prior to its final disposal in a deep geological repository. The issue of transporting waste on public roads is specifically covered, along with the strategy defined to securely transfer all legacy waste resulting from the CEA's very diverse R and D programs. Emphasis is placed on finding the right balance between collecting the necessary data on the waste to be transported, ensuring the reconditioning of this waste so it can be transported safely, and maintaining an efficient industrial-scale waste transfer process to limit the lead times and costs related to legacy waste recovery operations. (authors)

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DOE OSTI2019

Development of an Alternative Approach to Demonstrating Environmental Performance in the UK - 19354

European basic safety standards require that exposure of humans to the potentially harmful effects of ionising radiation are optimised so that they are as low as reasonably achievable, taking account of economic and societal factors. In the United Kingdom the responsibility for ensuring that exposures to humans are optimised is split between two regulatory functions. The Office of Nuclear Regulation regulates exposures to workers in all scenarios and to members of the public in the event of an accident. Environmental regulators in each of the four nations comprising the United Kingdom regulate exposures to members of the public from normal operations. Since the early 1990's, all permissions granted to nuclear sites in the UK for the discharge/disposal of radioactive waste to the environment have contained conditions requiring holders to demonstrate that they minimise the generation of radioactive waste, apply appropriate abatement and discharge/dispose of it in a way that minimises impacts on members of the public and the environment. For many years, the industry standard for providing this demonstration, which was heavily influenced by regulatory expectations and feedback, was to undertake one or a series of multi-attribute decision assessments (MADA). Wood's team of environmental specialists has been involved in undertaking such MADA studies for many years. Since the mid-2000's, our team has become increasingly aware of some of the limitations of undertaking MADA to demonstrate environmental performance for large scale new nuclear build projects and, in particular, the next generation of new nuclear power stations that are being planned and implemented in the UK. We have collaborated with the vendors of reactor technologies, the prospective/owner operators of new nuclear power stations and the developers of new facilities on existing sites to develop an alternative approach to demonstrating environmental performance. Our alternative approach comprises three main pillars which are summarised as: - Evolution - benefit from progressive improvements, operational experience and recent permitting activities. Our alternative approach uses the claim, argument, evidence approach which is commonly used for nuclear safety cases and which allows us to 'tell the story' of how environmental performance has been achieved. We have successfully deployed this approach for all new nuclear power stations proposed in the UK. - Integration - ensure that optimisation is integrated into the project programme and forms part of the 'big picture'. Our alternative approach allows environmental performance to be managed better throughout the life cycle of the project and ensures that our outputs are 'living' documents and do not 'sit on the shelf'. We have successfully demonstrated how this can be applied during rationalisation of major schemes and significant design changes on nuclear new build projects. - Opportunity - identify the best time to deliver elements of the programme. Our alternative approach aims to minimise the overall time, trouble and costs associated with demonstrating environmental performance by ensuring that decisions are made at the time when they offer greatest benefit and by minimising unnecessary rework. Our advice has helped to reduce the project costs and minimise delays associated with regulatory queries and interventions. This paper describes the journey that we have taken with our clients in developing our alternative approach to demonstrating environmental performance. We explain the UK regulatory context and explore the need for change. We present a number of case studies related to new nuclear power stations and other new facilities on existing nuclear sites that demonstrate the successful application of our alternative approach. We provide examples of how we have successfully implemented the outputs of our alternative approach into project and corporate management arrangements. Finally, we examine feedback from clients, regulators and other interested parties to propose future refinements. (authors)

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DOE OSTI2020

Simulation of DWPF Chemical Process Cell Using a Reaction Calorimeter

The Defense Waste Processing Facility (DWPF) converts liquid radioactive waste into glass using a Sludge Receipt Adjustment Tank (SRAT) and Slurry Mix Evaporator (SME) in addition to other processes. The SRAT and SME processes can be simulated on a small-scale using the Mettler-Toledo Reaction Calorimeter (RC1). This equipment is able to automate the execution and monitoring of experiments involving nonradioactive chemical simulants. During the chemical processing of sludge in the SRAT process, foam begins to form and can foam over the reaction vessel. The current antifoam used at DWPF creates a risk of increased flammability and has a limited pH range. Usually, mercury is circulated back into the reaction vessel during the SRAT process, making it difficult to remove. To combat foam over, novel antifoams will be tested during the boil-up process using the RC1. To improve mercury extraction, 3D-printed coalescer columns will be tested within the Mercury Water Wash Tank (MWWT) of the RC1. Testing of Simulant Sludge Batch 6 (SB6) was performed to ensure: the sludge would form a stable foam at boiling during full SRAT/SME runs; the antifoam would reduce the amount of foam within the vessel The sludge and a Precipitate Reactor Feed Tank (PRFT) simulant with monosodium titanate (MST) were mixed separately, combined, and then heated to boiling. As shown below, the antifoam used was able to eliminate the foam within the vessel, as expected. The FormLabs Form2 3D Printer can be used to create roughly 3-inch tall and 1-inch diameter columns from clear and ceramic resins. The columns will utilize chemical and physical forces to draw the mercury out of a condensed offgas stream leaving the RC1 reaction vessel. These columns will be tested during full SRAT/SME runs on the basis of how well they are able to collect elemental mercury. In preparation for full-scale runs, the RC1 equipment was assembled onto a custom table and tested with water runs. The software used to control automation, iControl, has a user-friendly interface that allows the experiment to be altered as needed. An example of the data recorded during an experimental run is represented in Figure 1 below. Water runs were utilized as a way to prove if the RC1 system was functioning in the following areas: Boil-up rates, Chemical additions using pumps, Temperature readings, Balance readings, Probe readings, Offgas system, Agitation. Future Plans: Five novel antifoams will be tested during full SRAT/SME runs. The most effective antifoam will be utilized in DWPF's Chemical Process Cell. The design of the Mercury Extraction Coalescer can be further manipulated to increase the extraction of not only elemental mercury, but other heavy metals in environmental remediation processes as well.

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DOE OSTI2018

Inorganic Ba-Sn Nanocomposite Materials for Sulfate Sequestration from Complex Aqueous Solutions

Selective sequestration of sulfate (SO42-) in the form of barite (BaSO4) from alkaline solutions of high ionic strength containing carbonate at high concentrations resembling legacy radioactive waste generated during plutonium cold war era production is problematic due to the preferential formation of BaCO3. Incorporation of sulfate into the insoluble and thermally stable BaSO4 phase can potentially benefit radioactive waste processing in reducing operational challenges and suppressing volatilization of other waste components such as technetium-99. To enhance selectivity of SO42- sequestration, a series of Ba(II)-Sn(IV) materials was prepared using simple hydrothermal synthesis from different Sn(II) and Sn(IV) precursors. Structural characterization indicated that all obtained products predominantly contained BaSn(OH)6 and Ba2SnO2(OH)4(H2O)10 crystalline phases which were disrupted upon exposure to SO42- due to formation of BaSO4. Performance of the Ba(II)-Sn(IV) materials was tested using complex alkaline solutions simulating radioactive waste containing 0.094 M SO42- and 0.5 M CO32- among other constituents. About 54 – 66% of SO42- was converted to BaSO4 when a quantity of Ba(II)-Sn(IV) material containing approximately a stoichiometric amount of Ba2+ relative to SO42- was used. In comparison, previous studies indicate negligible BaSO4 formation under similar conditions when a simple Ba2+ salt is used. This improvement is attributed to the selective replacement of the Ba(II) stannate by SO42-. Thermal stability of the sulfate-loaded product material up to 1100 °C was demonstrated. Obtained materials offer a convenient and economical option for the selective sequestration or removal of SO42- from carbonate containing solutions.

Johnson, Isaac↗
DOE OSTI2019

Decommissioning of a Historic Waste Storage Facility in Austria - 19059

Nuclear Engineering Seibersdorf Ltd. (NES) performed decommissioning of an old waste storage facility, the so called 'bunker', which was commissioned in 1968 and used until 1978 for the storage of radioactive waste from medical, industrial and research activities, as well as from the operation of the Austrian research reactor ASTRA. Due to the layout of the bunker the initial decommissioning plan was to decontaminate the inner walls of the caverns and afterwards perform clearance measurements on the standing structure. However, throughout the first phase of decommissioning it turned out, that contaminations penetrated deeper into the concrete structure as initially anticipated, which made an adoption of the decommissioning strategy necessary. By changing the workflow and to remove non-contaminated parts of the structure first, also the strategy for the clearance measurements had to be adapted. The selective decommissioning strategy and the demand that measurements shall be performed by an accredited body, required to set up an adequate communication structure to keep the workflow going. Moreover, a fast communication between NES, the competent authority and the authority's (external) experts was necessary to receive clearance licenses on time, and to prevent backlog of dismantled material. The paper reports on the initial and the adapted decommissioning strategy and the challenges within the changed approach. Applied decontamination and decommissioning methods and the implemented radiation protection measures will be discussed. Furthermore, information on the methodologies for clearance measurements, the collaboration of the accredited body with the decommissioning team, as well as on the communication between NES and the competent authority will be given. Finally, a balance of radioactive waste vs. conventionally disposed waste will be shown. (authors)

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

Hanford WTP LAW Melter Startup and Tuning Feed Material Development - 17065

The U.S. Department of Energy's (DOE) Hanford Site in southeast Washington state is home to 56 million gallons of chemical and radioactive waste stored in underground tanks, the result of more than four decades of plutonium production. To address this challenge, the U.S. Department of Energy contracted Bechtel National, Inc., to design and build the world's largest radioactive waste treatment plant. The Hanford Tank Waste Treatment and Immobilization Plant (WTP) will use vitrification to immobilize most of Hanford's waste. Vitrification involves blending the waste with glass-forming materials, and heating it to high temperature to form molten glass, then pouring it into stainless steel containers. In this glass form, its radioactivity will decay over hundreds to thousands of years. In advance of full operation of WTP, a sequential process of commissioning will be undertaken to ensure the plant operates safely and within design requirements. The first phase of WTP operations will focus on the Low-Activity Waste (LAW) Vitrification Facility and its supporting facilities and systems. Commissioning of the LAW Facility will employ a progressive approach to address the complexity of system operation and introduction of hazards. This approach starts with introduction of low hazard and simple testing configurations, and incrementally increasing the complexity and hazardous makeup of testing feed materials, culminating in fully integrated treatment of radioactive tank waste during hot commissioning. Testing with simulants that are chemically similar to radioactive tank waste introduces significant chemical hazards into treatment systems in the form of nitrogen oxides (NOx) and anhydrous ammonia. A method of testing to allow tuning and demonstration of equipment and to develop operator proficiency prior to introducing these chemical hazards was deemed necessary. To that end, the development and testing of a new 'tuning feed' material is in progress. The tuning feed is important because its use will enable the first-time integration of all LAW Facility process systems. The tuning feed must include key physical and chemical characteristics that correspond to the commissioning simulant and low activity tank waste, without the inherently hazardous materials or toxic off-gas constituents that result from tank waste treatment. Tuning feed is being developed to provide representative melter feed rheology, glass chemistry, cold cap formation, and off-gas generation rate in order to predict system response to actual tank waste to the extent practical. The tuning feed must also be compatible with process equipment and other process streams that will be encountered during the cold commissioning testing program. (authors)

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

Solid Waste Release Calculations for the Composite Analysis Baseline Assessment

The purpose of the Hanford Site Composite Analysis (CA) is to provide an estimate of the cumulative radiological impacts from active and planned low-level radioactive waste disposals and other potentially interacting radioactive waste disposal sources that will remain following Hanford Site closure. The need for a waste form release model addressing multiple solid waste forms is documented in CP-60410, Hanford Site Composite Analysis Technical Approach Description: Waste Form Release.

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