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At least 613 records · Page 34
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Impact of axial velocity and transmembrane pressure (TMP) on ARP filter performance

The Savannah River Site (SRS) is currently treating radioactive liquid waste with the Actinide Removal Process (ARP) and the Modular Caustic Side Solvent Extraction Unit (MCU). Recently, the low filter flux through the ARP of approximately 5 gallons per minute has limited the rate at which radioactive liquid waste can be treated. Salt Batch 6 had a lower processing rate and required frequent filter cleaning. Savannah River Remediation (SRR) has a desire to understand the causes of the low filter flux and to increase ARP/MCU throughput. One potential method for increasing filter flux is to adjust the axial velocity and transmembrane pressure (TMP). SRR requested SRNL to conduct bench-scale filter tests to evaluate the effects of axial velocity and transmembrane pressure on crossflow filter flux. The objective of the testing was to determine whether increasing the axial velocity at the ARP could produce a significant increase in filter flux. The authors conducted the tests by preparing slurries containing 6.6 M sodium Salt Batch 6 supernate and 2.5 g MST/L, processing the slurry through a bench-scale crossflow filter unit at varying axial velocity and TMP, and measuring filter flux as a function of time.

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

RADIOACTIVE DEMONSTRATIONS OF FLUIDIZED BED STEAM REFORMING WITH ACUTAL HANFORD LOW ACTIVITY WASTES VERIFYING FBSR AS A SUPPLEMENTARY TREATMENT

The U.S. Department of Energy's Office of River Protection is responsible for the retrieval, treatment, immobilization, and disposal of Hanford's tank waste. Currently there are approximately 56 million gallons of highly radioactive mixed wastes awaiting treatment. A key aspect of the River Protection Project cleanup mission is to construct and operate the Waste Treatment and Immobilization Plant (WTP). The WTP will separate the tank waste into high-level waste (HLW) and low-activity waste (LAW) fractions, both of which will subsequently be vitrified. The projected throughput capacity of the WTP LAW Vitrification Facility is insufficient to complete the cleanup mission in the time frame required by the Hanford Federal Facility Agreement and Consent Order, also known as the Tri-Party Agreement (TPA). Therefore, Supplemental Treatment is required both to meet the TPA treatment requirements as well as to more cost effectively complete the tank waste treatment mission. Fluidized Bed Steam Reforming (FBSR) is one of the supplementary treatments being considered. FBSR offers a moderate temperature (700-750 C) continuous method by which LAW and other secondary wastes can be processed irrespective of whether they contain organics, nitrates/nitrites, sulfates/sulfides, chlorides, fluorides, and/or radio-nuclides like I-129 and Tc-99. Radioactive testing of Savannah River LAW (Tank 50) shimmed to resemble Hanford LAW and actual Hanford LAW (SX-105 and AN-103) have produced a ceramic (mineral) waste form which is the same as the non-radioactive waste simulants tested at the engineering scale. The radioactive testing demonstrated that the FBSR process can retain the volatile radioactive components that cannot be contained at vitrification temperatures. The radioactive and nonradioactive mineral waste forms that were produced by co-processing waste with kaolin clay in an FBSR process are shown to be as durable as LAW glass.

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DOE OSTI2020

Computationally Aided Design of a Small-Scale Waste-Glass Melter - 20452

To provide mission support to the WTP for the vitrification of legacy nuclear tank waste at the Hanford site, a reduced-scale vitrification pilot system is being designed to process simulated and actual radioactive tank wastes. The waste stored in underground tanks will be separated into HLW and LAW fractions, where the majority is LAW by mass (approximately 90%) and HLW by activity (greater than 95%). The first tank waste to be processed at the WTP will be LAW. Because Hanford tank waste and the resultant melter feed compositions are known to vary widely, pilot-scale operations are essential to identify potential problems, provide needed data, confirm assumptions, and determine impacts to full-scale melters and off-gas systems from these different feeds. In addition, the reduced-scale melter system must be capable of quickly providing results to operations without the typically high cost of full-scale radioactive operations and to provide an engineering platform in close proximity to local operations staff. The objectives of this computational modeling effort are to produce process conditions like those encountered in the WTP LAW melters and to minimize the volume of radioactive waste necessary for the evaluations. To this end, melter design features are being explored to increase production without significantly increasing surface area or melter-glass volumes or compromising data quality. In support of these objectives, a set of computational fluid dynamic models was developed to provide insight into the flow patterns within these small melters and evaluate design features to increase production without significantly increasing glass inventory. Computed velocities were evaluated at the interface between the cold cap and the molten-glass pool and within the bulk-glass pool. Since melt rate correlates with the velocity underneath the cold cap, this information can be used to select design features that produce the highest gains. (authors)

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

Radioactive Waste Characterization Strategies; Comparisons Between AK/PK, Dose to Curie Modeling, Gamma Spectroscopy, and Laboratory Analysis Methods- 12194

In the coming fiscal years of potentially declining budgets, Department of Energy facilities such as the Los Alamos National Laboratory (LANL) will be looking to reduce the cost of radioactive waste characterization, management, and disposal processes. At the core of this cost reduction process will be choosing the most cost effective, efficient, and accurate methods of radioactive waste characterization. Central to every radioactive waste management program is an effective and accurate waste characterization program. Choosing between methods can determine what is classified as low level radioactive waste (LLRW), transuranic waste (TRU), waste that can be disposed of under an Authorized Release Limit (ARL), industrial waste, and waste that can be disposed of in municipal landfills. The cost benefits of an accurate radioactive waste characterization program cannot be overstated. In addition, inaccurate radioactive waste characterization of radioactive waste can result in the incorrect classification of radioactive waste leading to higher disposal costs, Department of Transportation (DOT) violations, Notice of Violations (NOVs) from Federal and State regulatory agencies, waste rejection from disposal facilities, loss of operational capabilities, and loss of disposal options. Any one of these events could result in the program that mischaracterized the waste losing its ability to perform it primary operational mission. Generators that produce radioactive waste have four characterization strategies at their disposal: - Acceptable Knowledge/Process Knowledge (AK/PK); - Indirect characterization using a software application or other dose to curie methodologies; - Non-Destructive Analysis (NDA) tools such as gamma spectroscopy; - Direct sampling (e.g. grab samples or Surface Contaminated Object smears) and laboratory analytical; Each method has specific advantages and disadvantages. This paper will evaluate each method detailing those advantages and disadvantages including; - Cost benefit analysis (basic materials costs, overall program operations costs, man-hours per sample analyzed, etc.); - Radiation Exposure As Low As Reasonably Achievable (ALARA) program considerations; - Industrial Health and Safety risks; - Overall Analytical Confidence Level. The concepts in this paper apply to any organization with significant radioactive waste characterization and management activities working to within budget constraints and seeking to optimize their waste characterization strategies while reducing analytical costs. (authors)

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DOE OSTI2016

Special Analysis for the Disposal of the Transuranic Waste Processing Center Mixed Low-Level Waste Stream at the Area 5 Radioactive Waste Management Site, Nevada National Security Site, Nye County, Nevada

This special analysis (SA) evaluates whether the Transuranic Waste Processing Center (TWPC) Mixed Low-Level waste stream is suitable for disposal by shallow land burial (SLB) at the Area 5 Radioactive Waste Management Site (RWMS) on the Nevada National Security Site (NNSS). Disposal of the TWPC Mixed Low-Level waste was meets all U.S. Department of Energy (DOE) Manual DOE M 435.1-1, “Radioactive Waste Management Manual,” Chapter IV, Section P performance objectives. The TWPC Mixed Low-Level waste stream is recommended for acceptance without conditions.

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DOE OSTI2018

Final Report: Vitrification of Inorganic Ion-Exchange Media, VSL-16R3710-1

One of the primary roles of waste pretreatment at the Hanford Tank Waste Treatment and Immobilization Plant (WTP) is to separate the majority of the radioactive components from the majority of the nonradioactive components in retrieved tank wastes, producing a high level waste (HLW) stream and a low activity waste (LAW) stream. This separation process is a key element in the overall strategy to reduce the volume of HLW that requires vitrification and subsequent disposal in a national deep geological repository for high level nuclear waste. After removal of the radioactive constituents, the LAW stream, which has a much larger volume but smaller fraction of radioactivity than the HLW stream, will be immobilized and disposed of in near surface facilities at the Hanford site.

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

First-Principles-Based Computational Study on Nucleation and Growth Mechanisms of U on Mo(110) Surface Solvated in an Eutectic LiCl-KCl Molten Salt

In nuclear engineering and industry electrochemical methods have been intensively employed in pyro-processing technology to recycle used nuclear fuels for substantial alleviation of the concern with the safe treatment of high-level radioactive nuclear wastes. Pyro-processing is based on sequential electrochemical reactions at high temperature (typically above 700 K at ambient pressure) to enhance the efficiency for decoupling nuclear fuels from unfavorable fission fragments. Using a molten salt in electrorefining process, at which actinides can be selectively deposited in solid cathode from ionic states, gives several advantages over an aqueous solution: higher thermal conductivity (it is liquid phase at high temperature, molten salt plays a role as medium at high temperature over 773 K in electrochemical reaction, lower kinetic barrier in ionic diffusion mobility and better stability against undesired side reactions with radioactive materials. Therefore, the volume of high-level nuclear wastes to be, otherwise, disposed into secured sites is dramatically reduced. Most of all it conforms to nuclear non-proliferation treaty. Electrochemical reactions in a molten salt medium are, however, much more complicated than in aqueous solution rendering the accurate understanding of thermodynamic and kinetic properties formidably challenging. It is represented by extensive researches in experiments and computations to identify underlying mechanisms on how U and other fission fragments behave in the electrorefining system. Regardless of the researches there is a general agreement that the initial stage of the nucleation and growth process of the U deposit is critical for determining the properties of final product and the electrorefining efficiency. In spite of the focused studies it still remains unclear. Theoretical and computational researches have been often carried out with various empirical parameter for interatomic potential energy, which are unknown variables until numerical fits to experimental measurements or just assumption is made. Thus, empirical simulations sometimes predicted outcomes are inconsistent depending on materials of interest with experimental measurements. This may be largely because the accuracy of an empirical interatomic potential energy models critically depends on the input parameters and thus, can be limited in transferability of the results. Furthermore, it may be devoid of fundamental physics or chemistry. This situation requires special caution when an electrorefining system is studied to recycle spent nuclear fuels because of the extreme difficulty in reliable acquisition of thermodynamic and kinetic data from direct experimental measurements. In this presentation, we extensively utilized first principles density functional theory (DFT) calculations and ab-initio molecular dynamic (AIMD) simulations to figure out underlying mechanisms of the initial stage of nucleation and growth behaviors of U on Mo(110) surface in a eutectic LiCl-KCl molten salt at T = 773 K. To rigorously investigate the effect of the molten salt electrolyte we compared the results with those calculated under vacuum condition.

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DOE OSTI2017

Commissioning of a GeoMelt Vitrification Plant at the Central Laboratory on the Sellafield Site - 17561

The National Nuclear Laboratory (NNL) and Kurion Veolia (Kurion) have installed a GeoMelt{sup R} In-Container Vitrification (ICV){sup R} radioactive waste treatment system in the Central Laboratory Active Rig Hall at Sellafield. This installation was completed in 2016. GeoMelt incorporates a range of patented and proprietary vitrification processing technologies that are configured in a variety of ways depending on the project and the waste. GeoMelt transforms hazardous chemical and radioactive wastes into an inert vitreous monolith. Hazardous organics are destroyed by pyrolysis and heavy metals and radionuclides are immobilized within the glass matrix where they are not preferentially leached into the environment. GeoMelt IVC produces a durable wasteform for a large variety of input wastes. The batch nature of the process which involves single-use refractory offers a wide range of process benefits over traditional vitrification technologies including: - a wide range of waste chemistries allowing high waste loadings; - compatibility with heterogeneous wastes including wastes with high metal content; - a technological simple solution without stirrers, bubblers, etc.; - No refractory corrosion concerns or melter drain failure scenarios; - Flexibility in glass chemistry and melter/disposal container configuration allows the wasteform to be specifically tailored to the storage or disposal site requirements. As part of the 2016 system commissioning at NNL's Central Laboratory, three demonstration melts were performed, one funded by Sellafield Ltd. and two by the Nuclear Decommissioning Authority (NDA). The first melt processed inactive soil with an inactive caesium tracer. The second melt processed inactive soil with a Cs-137 tracer. The third melt processed active waste soil from a variety of NDA nuclear licensed sites. The work conducted during commissioning was multi-purpose; it successfully demonstrated the benefits that the GeoMelt technology could bring to the UK NDA estate while performing equipment testing/troubleshooting and operator training. A significant benefit of the GeoMelt ICV technology is volume reduction for wastes which would otherwise experience a volume increase if encapsulated. Another benefit is the ability to treat difficult radioactive waste streams not amenable to encapsulation (e.g., reactive and/or organic wastes) with a very low energy consumption. The purpose of the first commissioning melt, Plant Performance Demonstration (PPD)-1, was to operate the GeoMelt system under typical conditions and to perform a post-melt mass balance of the caesium as a surrogate for Cs-137 in order to access retention of caesium in the final glass product. The melt processed a mixture of inactive soil obtained near the Sellafield site (as analogue for contaminated Sellafield soil), additives, and glass frit. An inactive caesium tracer was added to this melt as Cs{sub 2}CO{sub 3}. The purpose of the second commissioning melt, PPD-2, was to operate the GeoMelt system under typical conditions and to perform a post-melt activity balance using a Cs-137 tracer in order to assess retention of Cs-137 in the final glass product. Materials processed were inactive soil, glass frit, additives, and a radioactive caesium tracer (CsCl). The purpose of the third melt, Experiment-1 (EX-1) was to treat active waste and inactive waste simulant, as a demonstration of simultaneous processing of multiple waste streams. EX-1 consisted of processing active soil, Magnox sludge simulant, inactive soil, additives, and frit. The active soil used for this melt was obtained from the Hunterston A, Bradwell, Capenhurst, and Sellafield nuclear licensed sites. Since most soils are high in silica, the concept for this experiment was to use the active soils to provide glass formers while co-processing Magnox sludge simulant which does not contain glass formers. Beyond demonstration of capability areas of focus of this series of demonstrations included processing efficiency (kWh/kg treated) and caesium retention (%). This paper summarizes operations and results of the three melts. (authors)

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

Impact Of Sodium Oxalate, Sodium Aluminosilicate, and Gibbsite/Boehmite on ARP Filter Performance

The Savannah River Site (SRS) is currently treating radioactive liquid waste with the Actinide Removal Process (ARP) and the Modular Caustic Side Solvent Extraction Unit (MCU). Recently, the low filter flux through the ARP of approximately 5 gallons per minute has limited the rate at which radioactive liquid waste can be treated. Salt Batch 6 had a lower processing rate and required frequent filter cleaning. Savannah River Remediation (SRR) has a desire to understand the causes of the low filter flux and to increase ARP/MCU throughput. SRR requested SRNL to conduct bench-scale filter tests to evaluate whether sodium oxalate, sodium aluminosilicate, or aluminum solids (i.e., gibbsite and boehmite) could be the cause of excessive fouling of the crossflow or secondary filter at ARP. The authors conducted the tests by preparing slurries containing 6.6 M sodium Salt Batch 6 supernate, 2.5 g MST/L slurry, and varying concentrations of sodium oxalate, sodium aluminosilicate, and aluminum solids, processing the slurry through a bench-scale filter unit that contains a crossflow primary filter and a dead-end secondary filter, and measuring filter flux and transmembrane pressure as a function of time. Among the conclusions drwn from this work are the following: (1) All of the tests showed some evidence of fouling the secondary filter. This fouling could be from fine particles passing through the crossflow filter. (2) The sodium oxalate-containing feeds behaved differently from the sodium aluminosilicate- and gibbsite/boehmite-containing feeds.

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DOE OSTI2013

Geological Disposal Concept Selection Aligned with a Voluntarism Process - 13538

The UK's Radioactive Waste Management Directorate (RWMD) is currently at a generic stage in its implementation programme. The UK site selection process is a voluntarist process and, as yet, no communities have decided to participate. RWMD has set out a process to describe how a geological disposal concept would be selected for the range of higher activity wastes in the UK inventory, including major steps and decision making points, aligned with the stages of the UK site selection process. A platform of information is being developed on geological disposal concepts at various stages of implementation internationally and, in order to build on international experience, RWMD is developing its approach to technology transfer. The UK has a range of different types of higher activity wastes with different characteristics; therefore a range of geological disposal concepts may be needed. In addition to identifying key aspects for considering the compatibility of different engineered barrier systems for different types of waste, RWMD is developing a methodology to determine minimum separation distances between disposal modules in a co-located geological disposal facility. RWMD's approach to geological disposal concept selection is intended to be flexible, recognising the long term nature of the project. RWMD is also committed to keeping alternative radioactive waste management options under review; an approach has been developed and periodic reviews of alternative options will be published. (authors)

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

The ELIPSE Process: An Underwater Plasma Technology for Liquid Treatment - 16026

Radioactive liquid organic wastes are very various and produced in different quantities. Some of them are treated through specific processes when others are still waiting from outlet to be destroyed and stabilized. Their different composition (chlorinated, phosphate, sulfated, fluorinated, ...) make their treatment difficult in a same technology. The ELIPSE process has been design to propose an innovative solution leading to a global treatment of what is usually called 'orphan liquids'. This brand new technology involved a non-transferred plasma torch working under a water column ensuring the cooling, the filtration and the scrubbing of the gases coming from the combustion of the liquids. Furthermore, the water that ensures the global cooling of the system leads to low or to avoid the corrosion in a compact process. Only the nozzle of the torch may be affected by the corrosion and may become a wearing part of the system of treatment. After giving a detailed description of the ELIPSE process, the present paper will focus on the results obtained for the treatment of such different liquids as tributyl phosphate, perfluorinated oil and trichloroethylene. In addition to obtaining destruction efficiency upper than 99%, the corrosion seems to have been significantly limited. These features lead to the first conclusion that the ELIPSE process could be a solution for the future to the problem of the hazardous liquid treatment, nuclear or not. (authors)

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

The DECOVALEX international collaboration on modeling of coupled subsurface processes and its contribution to confidence building in radioactive waste disposal

Abstract The long-lived radiotoxicity of the high-level radioactive waste generated by nuclear power plants requires safe isolation from the biosphere for many hundreds of thousands of years. An international consensus has emerged that such isolation can best be provided by disposal in mined geologic repositories, a strategy that today is pursued by most countries dealing with radioactive waste. However, the need to predict the performance of such repositories over very long time periods generates large uncertainties that have to be accounted for in safety assessments. The findings from such safety assessments need to be conveyed to all stakeholders in a clear way, such that public confidence in geologic disposal solutions can be achieved. It is suggested here that close international collaboration on the technical aspects of geologic waste disposal has helped, and will continue to help, building trust and increasing confidence. This paper discusses a particular international collaboration initiative referred to as DECOVALEX, which brings together multiple teams and disciplines to collectively tackle complex experimental and modeling challenges related to geologic disposal. By describing how DECOVALEX works and by providing joint research examples, a case is made that such international collaboration contributes to knowledge transfer and confidence building in radioactive waste disposal science.

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

Conditioning and Requalification of Legacy Waste Packages - 18648

In most German interim storage facilities, hundreds of waste packages are waiting for re-conditioning and qualification to comply with the waste acceptance criteria (WAC) for the Konrad repository. Especially the waste classification as well as some characteristics and the description of the material composition in its current state do not comply with the Konrad requirements since the conditioning of these waste packages took place up to 40 years ago. The German name for these waste packages is 'Altgebinde', hereafter referred to as 'legacy waste packages'. The present article will provide an overview of the measures for the qualification of these legacy waste packages to fulfill the Konrad requirements. The following measures, when necessary, are carried out for the qualification of legacy waste packages: - Collection and evaluation of all documents related to the primary conditioning process, e. g. the used equipment, documented waste and packaging properties, conditioning and radiology, if available (hereafter 'old documents'). - Preparation of a conceptual report and a process control quality plan based on existing data, with appropriate measures to qualify the waste packages. Assessment and approval of the process control quality plan by the competent authority. - Characterization of the waste and packaging properties by visual inspection or tomography, recording or updating of dose rate, contamination and mass, verification of product stability, for example by gas analysis of the interior atmosphere of the container. - Characterization of the nuclide inventory by gammas scans, sample taking to analyze nuclides which can not be measured by means of in-situ gamma spectrometry, and for material declaration. - Qualification or post-conditioning of the legacy waste packages if necessary, e. g. by drying the waste packages and packaging them in containers approved for disposal. - Compilation of all results of the above mentioned tests and steps to create a waste package documentation, including the calculations for the nuclide inventory and the description of the material composition of the waste packages. - Receiving approval by the competent authority after evaluation of the final documentation. The aim behind this structured approach is to prevent additional, unnecessary measures and treatment of the waste. Consequently, the additional characterization is preferably carried out on the basis of the available documents. (authors)

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

Elimination of the Characterization of DWPF Pour Stream Sample and the Glass Fabrication and Testing of the DWPF Sludge Batch Qualification Sample

A recommendation to eliminate all characterization of pour stream glass samples and the glass fabrication and Product Consistency Test (PCT) of the sludge batch qualification sample was made by a Six-Sigma team chartered to eliminate non-value-added activities for the Defense Waste Processing Facility (DWPF) sludge batch qualification program and is documented in the report SS-PIP-2006-00030. That recommendation was supported through a technical data review by the Savannah River National Laboratory (SRNL) and is documented in the memorandums SRNL-PSE-2007-00079 and SRNL-PSE-2007-00080. At the time of writing those memorandums, the DWPF was processing sludge-only waste but, has since transitioned to a coupled operation (sludge and salt). The SRNL was recently tasked to perform a similar data review relevant to coupled operations and re-evaluate the previous recommendations. This report evaluates the validity of eliminating the characterization of pour stream glass samples and the glass fabrication and Product Consistency Test (PCT) of the sludge batch qualification samples based on sludge-only and coupled operations. The pour stream sample has confirmed the DWPF's ability to produce an acceptable waste form from Slurry Mix Evaporator (SME) blending and product composition/durability predictions for the previous sixteen years but, ultimately the pour stream analysis has added minimal value to the DWPF's waste qualification strategy. Similarly, the information gained from the glass fabrication and PCT of the sludge batch qualification sample was determined to add minimal value to the waste qualification strategy since that sample is routinely not representative of the waste composition ultimately processed at the DWPF due to blending and salt processing considerations. Moreover, the qualification process has repeatedly confirmed minimal differences in glass behavior from actual radioactive waste to glasses fabricated from simulants or batch chemicals. In contrast, the variability study has significantly added value to the DWPF's qualification strategy. The variability study has evolved to become the primary aspect of the DWPF's compliance strategy as it has been shown to be versatile and capable of adapting to the DWPF's various and diverse waste streams and blending strategies. The variability study, which aims to ensure durability requirements and the PCT and chemical composition correlations are valid for the compositional region to be processed at the DWPF, must continue to be performed. Due to the importance of the variability study and its place in the DWPF's qualification strategy, it will also be discussed in this report. An analysis of historical data and Production Records indicated that the recommendation of the Six Sigma team to eliminate all characterization of pour stream glass samples and the glass fabrication and PCT performed with the qualification glass does not compromise the DWPF's current compliance plan. Furthermore, the DWPF should continue to produce an acceptable waste form following the remaining elements of the Glass Product Control Program; regardless of a sludge-only or coupled operations strategy. If the DWPF does decide to eliminate the characterization of pour stream samples, pour stream samples should continue to be collected for archival reasons, which would allow testing to be performed should any issues arise or new repository test methods be developed.

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

Application of Generic Disposal System Models

This report describes specific GDSA activities in fiscal year 2015 (FY2015) toward the development of the enhanced disposal system modeling and analysis capability for geologic disposal of nuclear waste. The GDSA framework employs the PFLOTRAN thermal-hydrologic-chemical multi-physics code (Hammond et al., 2011) and the Dakota uncertainty sampling and propagation code (Adams et al., 2013). Each code is designed for massively-parallel processing in a high-performance computing (HPC) environment. Multi-physics representations in PFLOTRAN are used to simulate various coupled processes including heat flow, fluid flow, waste dissolution, radionuclide release, radionuclide decay and ingrowth, precipitation and dissolution of secondary phases, and radionuclide transport through the engineered barriers and natural geologic barriers to a well location in an overlying or underlying aquifer. Dakota is used to generate sets of representative realizations and to analyze parameter sensitivity.

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

Operational History of the Area 3 Radioactive Waste Management Site

Report presents information based on characterization and monitoring data, historical documentation and records, modeling studies, evaluations, and process knowledge for the Area 3 Radioactive Waste Management Site on the Nevada National Security Site.

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

Preparation of Waste Fingerprints for the Miscellaneous Beta Gamma Waste Feeds to the Box Encapsulation Plant at Sellafield - 16080

There are a number of legacy pond and silo storage facilities on Sellafield Site that contain a broad range of ILW arising from the early Windscale Pile and Magnox Fuel Reprocessing Programmes. The waste comprises of solid Miscellaneous Beta Gamma Waste (MBGW) from operations (including filters, containerised waste, thermocouple wire, support struts, pumps, soft wastes) and associated sludges that have arisen from the corrosion of the fuel, Magnox Swarf cladding and the MBGW. It is intended to retrieve and package this waste to meet the requirements of interim storage and final disposal in the Box Encapsulation Plant (BEP) at Sellafield. As part of this process each package must be assigned a realistic and justifiable radionuclide and material inventory. These inventories can either be assigned by radiometric measurements including the associated errors or developed by fundamental data underpinning these assessments and a 'chain of custody' approach of knowing where the waste has been and how it has evolved. The subject of this paper is concerned with the development of the waste fingerprints for these legacy waste streams. As the wastes have arisen over a ∼70 year period there have been an assortment of different working practices and developments in equipment design that have led to significant variations in the waste condition and its composition. During storage the waste items have become mixed, corroded and sludge deposits have arisen. Because of this, and that early consignment records are either incomplete or missing, the task of assigning material and radionuclide fingerprints has been very challenging. In addition, to comply with regulatory requirements the inventory assignment has to be sufficiently robust to meet the processing envelope of the encapsulation process i.e. to allow controls to be made on the amounts of reactive materials (e.g. aluminium, Magnox and uranium) that can be processed within each product and allow identification of excluded items from being consigned from the donor plants. Also, as the waste comprises actinides, fission and activation products it was noted that whilst the UK standard inventory code FISPIN10 predicts the actinides and fission products well, there was a significant discrepancy in the prediction of activation products when compared with those predicted by the specialised fusion code FISPACT. It should be noted that the use of FISPACT to predict actinides and fission products is not recommended. To address these issues, - position statements were produced for each of the waste streams that would be consigned to BEP; - gaps within the data were assessed and where possible inactive trial work was identified to address it; - material and radionuclide fingerprint justification documents were developed from the position statements based upon available data, the experimental work and justified assumptions; - where FISPIN runs were required, these were run with a modified activation data library based on the European Activation File 2010 (EAF2010) to model the activation products. This paper summarises the BEP process, discusses examples of how the material and radionuclide waste stream fingerprints have been justified by revisiting the sample data, original consignment records, engineering drawings and recent remotely operated vehicle (ROV) pond surveys and making reasoned assumptions based on operational practices and non-active experimental work. Incorporation of the EAF2010 library within the FISPIN code has offered significant benefits to BEP and indeed other plants processing mixed fission and activated wastes as it allows comparable predictions of activation products compared with that made by FISPACT and for the plant to maintain one data processing stream rather than two if it had to utilise both modelling codes. This work will underpin the inventory submission in support of the Interim Letter of Compliance (ILoC) to Radioactive Waste Management Limited (RWM). The process of examining the original records, plant operation schedules and engineering drawings has enabled removal of some of the pessimisms that have been made in previous assessments. (authors)

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

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

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

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