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DOE OSTI2023

Modeling of Chemical Slurry Rheology in DWPF Sludge Batch (SB) 10 Simulants

The Defense Waste Processing Facility (DWPF) treats high-activity radionuclides from sludge through a process called vitrification. This process converts radioactive liquid waste currently stored in tank farms into a solid glass form that is suitable for long-term storage and disposal. Due to the complexities involved in vitrifying this waste within each operation of the Chemical Processing Cell (CPC), waste rheology is studied to characterize the fluid-mechanical properties as it passes through the CPC and into the Melter. To better understand the waste and validate flow behavior, slurry rheology of simulants that represents the waste was studied at various acid stoichiometry percentages and solids concentrations to determine the simulant’s yield stress and viscosity. This research work has been supported by the DOE-FIU Science & Technology Workforce Development Initiative, an innovative program developed by the U.S. Department of Energy’s Office of Environmental Management (DOE-EM) and Florida International University’s Applied Research Center (FIU-ARC). During the spring of 2022, a DOE Fellow intern, Brendon Cintas, spent 10 weeks doing a summer internship at Savannah River National Laboratory (SRS) under the supervision and guidance of Dan Lambert, Chemical Flowsheet Development. The intern’s project was initiated on June 6, 2022, and continued through August 11, 2022 with the objective of assisting scientists at SRNL’s Rheology and Grout Laboratory at Aiken Country Technology Lab (ACTL) better understand the sludge composition on the rheology of a simulant slurry using a HAAKE RheoStress 6000 rheometer and extrapolate the results to the real-waste data.

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DOE OSTI2017

Defense waste processing facility (DWPF) liquids model: revisions for processing higher TIO 2 containing glasses

Radioactive high level waste (HLW) at the Savannah River Site (SRS) has successfully been vitrified into borosilicate glass in the Defense Waste Processing Facility (DWPF) since 1996. Vitrification requires stringent product/process (P/P) constraints since the glass cannot be reworked once it is poured into ten foot tall by two foot diameter canisters. A unique “feed forward” statistical process control (SPC) was developed for this control rather than statistical quality control (SQC). In SPC, the feed composition to the DWPF melter is controlled prior to vitrification. In SQC, the glass product would be sampled after it is vitrified. Individual glass property-composition models form the basis for the “feed forward” SPC. The models transform constraints on the melt and glass properties into constraints on the feed composition going to the melter in order to guarantee, at the 95% confidence level, that the feed will be processable and that the durability of the resulting waste form will be acceptable to a geologic repository. This report documents the development of revised TiO 2 , Na 2 O, Li 2 O and Fe 2 O 3 coefficients in the SWPF liquidus model and revised coefficients (a, b, c, and d).

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DOE OSTI2020

Innovative Projects on RW Recycling from the French 'Investments for the Future' Program - 20023

The extent and duration of French dismantling operations require to anticipate at the earliest possible stage the management of the large volumes of radioactive waste that will be generated, and to engage collectively in sustainable innovation processes. To anticipate all constraints and requirements related to radioactive waste management and secure the long-term compatibility with waste disposals, the French State has decided to give to Andra, within the 'Investissements d'Avenir' program, an endowment to support innovation at all steps of radioactive waste life cycle with a special emphasis on the dismantling process. Currently, some 30 R and D projects dealing with the optimization of dismantling radioactive waste management are underway. They aim at providing strong innovations in terms of characterization, processing, conditioning and disposal of post dismantling radioactive waste. A specific focus will be made on those dealing with radioactive waste and radionuclides recycling. (authors)

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DOE OSTI2015

Support of SRS salt-stone regulatory compliance - 15586

The SRS salt-stone Facility is designed and permitted to immobilize and dispose of low-level radioactive and hazardous liquid waste (salt solution) remaining from the processing of radioactive material at the Savannah River Site. The salt-stone Processing Facility (SPF) is permitted as a wastewater treatment facility per the South Carolina Department of Health and Environmental Control Regulations (SCDHEC) South Carolina Hazardous Waste Management Regulations (SCHWMR) R.61-67. The salt-stone Disposal Facility is permitted as a solid waste landfill site, as defined by SCDHEC Regulations R61-107.19. The salt-stone Sampling and Analysis Plan provides the SCDHEC with the chemical and physical characterization strategy for the salt solution and salt-stone waste form. During operation, samples are collected from Tank 50H on a quarterly basis and are analyzed by the Savannah River National Laboratory (SRNL) for over chemical and radioactive components. The analytical results are compared to the salt-stone waste acceptance criteria in a technical report. SRNL also prepares salt-stone samples from the quarterly sample and premix binder materials provided by the SPF to determine the non-hazardous nature of the grout to meet the requirements of the SCDHEC regulation for the 'Characteristic of Toxicity', R.61-79.261.24(b) and the 'Universal Treatment Standards', R.61-79.268.48.4. salt-stone samples are collected, size reduced, packaged, and shipped to an independent United States EPA certified laboratory, or a laboratory that holds an equivalent certification such as one from the National Environmental Laboratory Accreditation Program to perform the EPA Toxicity Characteristic Leaching Procedure (TCLP) extraction of the salt-stone sample. SRNL reviews the results of the TCLP analysis and prepares a report detailing the sample preparation and analysis that is used by the salt-stone Facility to demonstrate the nonhazardous nature of the salt-stone waste form. (authors)

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DOE OSTI2021

Saltstone Waste Characterization Analysis - Salt Waste Processing Facility (SWPF) Waste Stream (Q4 CY2020)

The Saltstone Production and Disposal Facility is designed and permitted by the state of South Carolina Department of Health and Environmental Control (SCDHEC) to treat and dispose of low-level radioactive and hazardous liquid waste (salt solution) remaining from the processing of radioactive material at the Savannah River Site (SRS). Low-level waste (LEW) aqueous streams from the Effluent Treatment Project (ETP) and decontaminated solutions from the Tank Closure Cesium Removal Unit (TCCR) and the Salt Waste Processing Facility (SWPF) are stored in Tank 50 until the LEW can be transferred to the Saltstone Facility for treatment and disposal. In the past, decontaminated solution from the Modular Caustic Side Solvent Extraction Unit (MCU) was stored in Tank 50 until the LEW could be transferred to the Saltstone Facility for treatment and disposal. MCU is currently in a suspended operations state. LEW that meets the Waste Acceptance Criteria (WAC) can be transferred, stored, and treated in the Saltstone Production Facility (SPF) for subsequent disposal as saltstone grout in the Saltstone Disposal Facility (SDF). Sampling will be conducted as new waste streams are identified for treatment and disposal at the Saltstone Industrial Wastewater Treatment Facility (IWTF) and Z-Area Industrial Solid Waste Landfill (ISWLF), Facility ID# 025500-1603, General Condition B.9 or every six years in accordance with South Carolina (SC) Regulation 61-107.19 Parti C, “Solid Waste Management: Solid Waste Landfills and Structural Fill - General Requirements.”

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DOE OSTI2013

Handling Radioactive Waste from the Proton Accelerator Facility at the Paul Scherrer Institut (PSI) - Always Surprising? - 13320

The Paul Scherrer Institut (PSI) is the largest national research centre in Switzerland. Its multidisciplinary research is dedicated to a wide field in natural science and technology as well as particle physics. In this context, PSI is operating, amongst others, a large proton accelerator facility since more than 30 years. In two cyclotrons, protons are accelerated to high speeds and then guided along roughly 100 m of beam line to three different target stations to produce secondary particles like mesons and neutrons for experiments and a separately beam line for UCN. The protons induce spallation processes in the target materials, and also at other beam loss points along the way, with emission of protons, neutrons, hydrogen, tritium, helium, heavier fragments and fission processes. In particular the produced neutrons, due to their large penetration depth, will then interact also with the surrounding materials. These interactions of radiation with matter lead to activation and partly to contamination of machine components and the surrounding infrastructures. Maintenance, operation and decommissioning of installations generate inevitably substantial amounts of radioactive operational and dismantling waste like targets, magnets, collimators, shielding (concrete, steel) and of course secondary waste. To achieve an optimal waste management strategy for interim storage or final disposal, radioactive waste has to be characterized, sorted and treated. This strategy is based on radiation protection demands, raw waste properties (size, material, etc.), and requirements to reduce the volume of waste, mainly for legal and economical reasons. In addition, the radiological limitations for transportation of the waste packages to a future disposal site have to be taken into account, as well as special regulatory demands. The characterization is a task of the waste producer. The conditioning processes and quality checks for radioactive waste packages are part of an accredited waste management process of PSI, especially of the Section Dismantling and Waste Management. Strictly proven and accepted methods needed to be developed and enhanced for safe treatment, transport, conditioning and storage. But in the field of waste from research activities, individual and new solutions have to be found in an increasingly growing administrative environment. Furthermore, a wide variety of components, with a really large inventory of radioactive nuclides, has to be handled. And there are always surprising challenges concerning the unusual materials or the nuclide inventory. In case of the operational and dismantling radioactive accelerator waste, the existing conditioning methods are in the process of a continuous enhancement - technically and administratively. The existing authorized specifications of conditioning processes have to be extended to optimize and fully describe the treatment of the inevitably occurring radioactive waste from the accelerator facility. Additional challenges are the changes with time concerning the legal and regulatory requirements - or do we have to consider it as business as usual? This paper gives an overview of the current practices in radioactive waste management and decommissioning of the existing operational accelerator waste. (authors)

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DOE OSTI2020

A Dose Assessment Model for Radioactive Waste Exposed by Gully Erosion at West Valley - 20513

The Western New York Nuclear Service Center (WNYNSC), located approximately 48 km south of Buffalo, New York, is the site of a former nuclear fuel reprocessing and radioactive waste disposal facility. Spent nuclear fuel was processed there from 1966 to 1972, leaving behind radioactive and chemical wastes in two disposal areas and a waste tank farm. Site operations also resulted in releases of radioactivity to site soils, groundwater, and to surface waters draining the site. The New York State Energy Research and Development Authority (NYSERDA) and the U.S. Department of Energy (DOE) are collaborating in a process of decision making for decommissioning those facilities remaining at the WNYNSC following the completion of Phase 1 decommissioning. Neptune and Company, Inc. (Neptune) was contracted to develop a probabilistic performance assessment (PPA) computer model to assist the agencies in this process. The PPA Model includes a contaminant transport component focusing on the movement of contaminants within and among environmental media including groundwater and surface water transport, contaminant translocation by plants and animals, diffusion, and erosion. The model also includes evaluation of potential exposure and health effects for a Resident Farmer exposure scenario, where the Resident Farmer represents a critical group, described as that group of individuals reasonably expected to receive the greatest exposure to residual radioactivity for any applicable set of circumstances. The West Valley Site is located in the glaciated Allegheny Plateau region of western New York State. The waste reprocessing and disposal areas were constructed on a relatively fat area of plateau dissected by drainages of Buttermilk Creek, including Erdman Brook, Franks Creek, and Quarry Creek. An important aspect of the contaminant transport component of the PPA model is consideration of erosive processes such as slumping of the stream slopes and the advance of gullies from these streams. These erosion processes remove material from the plateau, growing the size of the creek valleys and making them wider and deeper. Of particular interest for the impact of erosion is radioactive waste contained in the Nuclear Regulatory Commission (NRC)-Licensed Disposal Area (NDA), the New York State-Licensed Disposal Area (SDA), and residual radiological inventory in the underground storage tanks at the Waste Tank Farm (WTF). The PPA Model is organized around geographically-defined facilities which were constructed upon the plateau, including the NDA, SDA, and WTF. Ongoing stream erosion processes will potentially transfer radioactive waste and residual inventory from these facilities to the ground surface on adjacent hillslope areas where erosion has breached the facility. Hence, it is important to evaluate the consequences of potential exposures to a Resident Farmer on the hillslopes below a breached facility. Two interrelated aspects of the dose assessment model related to hillslope exposure are discussed: 1) representation of the physical processes related to transport of radionuclides from facilities onto the hillslopes, and from the hillslopes into adjoining creeks where contaminated material migrates downstream with surface water and sediment, and; 2) adaptation of the activities associated with the Resident Farmer scenario to assess potential exposures to contamination in the hillslope areas. This discussion will cover the conceptual basis of the hillslope exposure and transport models, and also implementation in the PPA computer model. (authors)

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DOE OSTI2018

Evaluation of Simple Chemical Interactions in the Defense Waste Processing Facility (DWPF) Chemical Process Cell (CPC) Under the Glycolic Acid Flowsheet

This report describes results from testing developed for the increased understanding of simple chemical interactions relevant to the processing and vitrification of radioactive waste at the Savannah River Site (SRS) in Aiken, SC. This work is outlined in task 3 of the Technical Task Request X-TTR-S-00024, Rev. 0 and task 4 of the Task Technical & Quality Assurance Plan SRNL- RP-2014-01183 Rev. 0. The testing was conducted from May 2016 through February 2017 at the Aiken County Technology Laboratory (ACTL) and performed by Savannah River National Laboratory (SRNL) personnel. This testing involved the investigation of simple reactions and interactions of glycolic or nitric acids with combinations of sodium nitrite, manganese dioxide, mercuric oxide, and noble metals (silver, palladium, rhodium, and ruthenium). The aim of this testing was to better understand the reactions and chemical phenomena previously observed in the Sludge Receipt and Adjustment Tank (SRAT) cycle of the Chemical Process Cell (CPC) in the Defense Waste Processing Facility (DWPF).

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DOE OSTI2023

Cesium Removal from 5.5 and 7.0 M Na AP-105 Using Crystalline Silicotitanate

The Tank Side Cesium Removal (TSCR) system, currently operational by Washington River Protection Solutions LLC (WRPS), prepares initial low-activity Hanford waste tank supernate feeds for the Hanford Waste Treatment and Immobilization Plant (WTP) Low-Activity Waste (LAW) Facility. In addition to entrained solids removal from the supernate, the primary goal of TSCR is to remove cesium-137 ( 137 Cs) by ion exchange, allowing contact handling of the liquid effluent product at the WTP as governed by a waste acceptance criterion (WAC). Specific to 137 Cs, this requirement is < 3.18E-5 Ci 137 Cs/mole of Na. Crystalline silicotitanate (CST) ion exchange media, manufactured by Honeywell UOP, LLC (product IONSIVTM R9140-B), has been selected as the ion exchange media for TSCR. CST is a non-elutable inorganic material that has demonstrated robust chemical, physical, and radiation tolerance while maintaining functionality. However, testing to date on actual tank waste samples has been limited to Na concentrations between 5 and 6 M Na (Fiskum et al. 2019a, 2021b and Westesen et al. 2021a, 2021b, 2022) while actual tank conditions can reach upwards of 9 M Na. Testing with feed from Hanford tank AP-105 incorporated testing at both 5.5 and 7 M Na in order to evaluate the impact of Na concentration on volume of waste processed before reaching the WAC. A 6-L volume of 5.5 M Na AP-105 and an 8-volume of 7 M Na AP-105 was processed through the Radioactive Waste Test Platform system, established at Pacific Northwest National Laboratory to support small-scale waste qualification efforts. The columns consisted of 6- to 9-mL CST beds (CST Lot 2002009604, sieved to screen out >30-mesh particles) placed in 1.5-cm-inner-diameter columns. Feed was processed at 1.9 bed volumes (BV) per hour; the flowrate, in terms of contact time with the CST bed, matched the expected flowrate at TSCR. Table S.1 and Figure S.1 summarize the measured AP-105 Cs load performance for each feed condition.

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DOE OSTI2012

Radionuclide and contaminant immobilization in the fluidized bed steam reforming waste products

The goal of this chapter is to introduce the reader to the Fluidized Bed Steam Reforming (FBSR) process and resulting waste form. The first section of the chapter gives an overview of the potential need for FBSR processing in nuclear waste remediation followed by an overview of the engineering involved in the process itself. This is followed by a description of waste form production at a chemical level followed by a section describing different process streams that have undergone the FBSR process. The third section describes the resulting mineral product in terms of phases that are present and the ability of the waste form to encapsulate hazardous and radioactive wastes from several sources. Following this description is a presentation of the physical properties of the granular and monolith waste form product including and contaminant release mechanisms. The last section gives a brief summary of this chapter and includes a section on the strengths associated with this waste form and the needs for additional data and remaining questions yet to be answered. The reader is directed elsewhere for more information on other waste forms such as Cast Stone (Lockrem, 2005), Ceramicrete (Singh et al., 1997, Wagh et al., 1999) and geopolymers (Kyritsis et al., 2009; Russell et al., 2006).

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DOE OSTI2013

Integrated High-Level Waste System Planning - Utilizing an Integrated Systems Planning Approach to Ensure End-State Definitions are Met and Executed - 13244

The Savannah River Site (SRS) is a Department of Energy site which has produced nuclear materials for national defense, research, space, and medical programs since the 1950's. As a by-product of this activity, approximately 37 million gallons of high-level liquid waste containing approximately 292 million curies of radioactivity is stored on an interim basis in 45 underground storage tanks. Originally, 51 tanks were constructed and utilized to support the mission. Four tanks have been closed and taken out of service and two are currently undergoing the closure process. The Liquid Waste System is a highly integrated operation involving safely storing liquid waste in underground storage tanks; removing, treating, and dispositioning the low-level waste fraction in grout; vitrifying the higher activity waste at the Defense Waste Processing Facility; and storing the vitrified waste in stainless steel canisters until permanent disposition. After waste removal and processing, the storage and processing facilities are decontaminated and closed. A Liquid Waste System Plan (hereinafter referred to as the Plan) was developed to integrate and document the activities required to disposition legacy and future High-Level Waste and to remove from service radioactive liquid waste tanks and facilities. It establishes and records a planning basis for waste processing in the liquid waste system through the end of the program mission. The integrated Plan which recognizes the challenges of constrained funding provides a path forward to complete the liquid waste mission within all regulatory and legal requirements. The overarching objective of the Plan is to meet all Federal Facility Agreement and Site Treatment Plan regulatory commitments on or ahead of schedule while preserving as much life cycle acceleration as possible through incorporation of numerous cost savings initiatives, elimination of non-essential scope, and deferral of other scope not on the critical path to compliance. There is currently a premium on processing and storage space in the radioactive liquid waste tank system. To enable continuation of risk reduction initiatives, the Plan establishes a processing strategy that provides tank space required to meet, or minimizes the impacts to meeting, programmatic objectives. The Plan also addresses perturbations in funding and schedule impacts. (authors)

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DOE OSTI2018

Replacement of Melter 2 at the Defense Waste Processing Facility (DWPF) - 18608

The original DWPF melter was replaced in 2002 with an identical unit, and radioactive operation of Melter 2 commenced in 2003. After nearly 14 years of service, a lower electrode failure of Melter 2 occurred in February 2017 which required its removal and replacement with Melter 3. The scope of the replacement outage involves several significant evolutions: - Shutdown of Melter 2 and placement into a safe state; - Remote external and internal inspection of Melter 2 to chronicle its condition prior to removal; - Disassembly and removal of Melter 2 from the process cell; - Placement of Melter 2 into a storage box and transport to a storage vault; - Preparation and initial testing of Melter 3 prior to transport to the DWPF; - Transport to the DWPF and installation of Melter 3 in the process cell; - Final startup testing and initiation of radioactive operation. Upon determination that repair of Melter 2 was not feasible, DWPF created an outage team to integrate schedules and resources needed to execute melter replacement as well as including maintenance scope for facility work that can only be performed when the melter is de-energized. The execution of this scope required integration with other mission critical outage activities already in progress, and planned to start, including initial facility tie-in scope for the new Salt Waste Processing Facility (SWPF). This unplanned addition of the melter replacement scope to the existing outage scope resulted in the longest facility outage period since the initiation of radioactive operations at the DWPF in 1996. Successful completion of this scope across all Liquid Waste facilities would require balance - for schedules, resources, priority, funding, etc. To overcome the significant challenges created by this additional scope, a WAR Room was established to ensure adequate integration and coordination of critical resources, and daily progress. The execution of this scope has required the integration of nearly two dozen work groups across multiple facilities at Savannah River Site (SRS), in addition to support from nearly a dozen vendors. The melter replacement outage was executed with a high degree of disciplined operation to ensure the facility was ready to resume glass pouring operation in time to support restart of the Liquid Waste program at the Savannah River Site (SRS). This paper will provide a brief overview of Melter 2 operational history. The execution of the removal and replacement operations will be described with a specific focus on lessons learned and opportunities where the outage execution diverged from the original Melter 1 replacement outage. It is expected to be of interest to vitrification facilities, as well as facilities, which require remote maintenance and replacement of large process equipment. (author)

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DOE OSTI2011

THREE-DIMENSIONAL THERMAL MODELING ANALYSIS OF CST MEDIA FOR THE SMALL ION EXCHANGE PROJECT

The Small Column Ion Exchange (SCIX) project is designed to accelerate closure of High Level Waste (HLW) tanks at the Savannah River Site (SRS). The SRS tanks store HLW in three forms: sludge, saltcake, and supernate. An in-tank ion exchange process is being designed to treat supernate and dissolved saltcake waste. Through this process, radioactive cesium from the salt solution is adsorbed into Crystalline Silicotitanate (CST) ion exchange media packed within a flow-through column. A packed column loaded with radioactive cesium generates significant heat from radiolytic decay. The waste supernate solution within the ion exchange bed will boil around 120 C. Solution superheating above the boiling point within the column could lead to violent hazardous energy releases. System heating from loaded CST is also of concern in other process modules, such as the waste tank. Due to tank structural integrity concerns, the wall temperature limit for the SRS waste tanks is 100 C. The transfer of cesium-loaded CST to the tank could result in localized hot spots on the tank floor and walls which may exceed this limit. As a result, thermal modeling calculations have been conducted to predict the maximum temperatures achievable both in the column and in the waste tank. As specified in the associated Technical Task Plan, one objective of the present work was to compute temperature distributions within the ion exchange column module under accident scenarios including loss of salt solution flow through the bed and loss of coolant system flow. The column modeling domain and the scope of the calculations in this case were broadened relative to previous two-dimensional calculations to include vertical temperature distributions within the packed bed of ion exchange media as well as the upper column plenum region containing only fluid. The baseline design conditions and in-column modeling domain for the ion-exchange column module are shown in Figure 1. These evaluations assumed the maximum bounding cesium loading considered possible based on current knowledge regarding CST media and the anticipated feed compositions. Since this cesium loading was considerably higher than the nominal loading conditions in SRS waste, cases with lower loading were also evaluated. Modeling parameters were the same as those used previously unless otherwise indicated. The current model does not capture multi-phase cooling mechanisms operative when solution boiling occurs. This feature is conservative in the sense that it does not account for the large cooling effects associated with phase transfer. However, the potential transfer of heat to the plenum region associated with vertical bubble ascension through the column during boiling is also neglected. Thermal modeling calculations were also performed for the entire waste storage tank for the case where loaded and ground CST was transferred to the tank. The modeling domain used for the in-tank calculations is provided in Figure 2. The in-tank domain is based on SRS Tank 41, which is a Type-IIIA tank. Temperature distributions were evaluated for cylindrical, ground CST mounds located on the tank floor. Media grinding is required prior to vitrification processing of the CST in the SRS Defense Waste Processing Facility (DWPF). The location of the heat source region on the tank floor due to the accumulation of CST material was assumed to be just under the grinder. The shape of the CST mound was assumed to be cylindrical. This shape is believed to be most representative of the actual mound shape formed in the tank, given that submersible mixing pumps will be available for media dispersion. Alternative configurations involving other geometrical shapes for the CST mound were evaluated in the previous work. Sensitivity analysis for the in-tank region was performed for different amounts of CST media. As was the case for the in-column model, the in-tank model does not include multi-phase cooling mechanisms operative when solution boiling occurs. The in-column and the in-tank evaluations incorporated recently updated maximum cesium loading levels calculated using the current SCIX feed compositions, which resulted in significantly higher cesium loading than previously calculated. The calculations were conducted to ensure conservative predictions for the maximum temperatures achievable using the current baseline design. The degree of conservatism was reduced for in-column calculations relative to the previous work by using a three-dimensional modeling approach and selecting parameters which were nearer to expected conditions. The degree of conservatism for the in-tank calculations was also reduced by lowering the soil penetration depth below the tank from 150 to 20 feet. The primary goals of the extended thermal modeling effort were to determine whether fluid boiling or superheating are possible within the column module and to determine the maximum floor temperatures within the tank loaded with spent CST.

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DOE OSTI2015

Radiological characterization of high Cs-137 waste items at LANL combining Cs-137 NDA results and acceptable process knowledge - 15592

Radiological characterization is an important component to disposition and disposal of radioactive waste at the Los Alamos National Laboratory. Gamma and neutron based non-destructive assay systems are available at LANL's TA-54 Area G for radiological characterization and are generally very effective in quantifying the transuranic radionuclide concentration required for a TRU or low level waste disposition determination. This is routinely performed by scaling some nuclides to direct assay results using acceptable process knowledge scaling from a directly measured key TRU nuclide, e.g., for gamma systems using a standard isotopic distribution to determine the amount of Pu-240 associated with a measured value of Pu-239. However, for two groups of LANL waste items that have high levels of Cs-137, these non-destructive assay systems were unable to make a standard TRU/LLW determination. Therefore, it was necessary to apply an extension of the typical scaling technique that keyed in on the direct Cs-137 gamma assay result to effectively characterize these waste items for TRU processing or LLW disposition. Two basic types of waste packages contained the high Cs-137 waste requiring special scaling: drums and boxes. The drums included 55-gallon, 85-gallon and 110-gallon volume sizes and the boxes included a corrugated metal box, a Standard Waste Box, and several overpack boxes containing oversized waste. All of these items contained radioactive waste generated from operations at the Chemistry and Metallurgical Research facility at LANL that included hot cell, metallographic sectioning and maintenance activities. All of the drums as well as the CMB, SWB and one oversized item in a plywood box were located above-ground at LANL's TA-54 Area G, while five of the oversized items were hot cell liners in metal boxes stored below-ground in shafts. All of the items were measured in the 2013-2014 time period at TA-54 Area G with LANL NDA systems. Most of the items were counted with LLW certified far-field gamma spectroscopy systems employing portable far-field HPGe detectors and carts and some of the items were counted on WIPP certified gamma only or gamma with neutron NDA systems. In all cases Cs-137 was quantified directly with a gamma measurement but the key TRU nuclides were not detectable. Thus, for each waste item AK/PK was studied and scaling factors to Cs-137 were established and applied to provide a comprehensive radiological characterization. A total of 49 waste items were successfully characterized as LLW based on this technique. The above ground waste items were dispositioned and disposed of at an offsite LLW repository and the below ground boxes were left in place as LLW in the TA-54 Area G shafts. This NDA with scaling application has proved effective and defensible in these high Cs-137 cases at LANL and will continue to be used as applicable and appropriate for future waste characterization at LANL. (authors)

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DOE OSTI2015

OECD NEA radioactive waste repository metadata management (RepMet) initiative (2014-2018) - 15614

The OECD/NEA Radioactive Waste Repository Metadata Management (RepMet) initiative aims to bring about a better understanding of the identification and administration of metadata - a key aspect of data management - to support national programmes in managing their radioactive waste repository data, information and records in a way that is both harmonised internationally and suitable for long-term management and use. At this point in the project, the main focus lies on 'waste packages at their storage sites ready for geological disposal' but excluding spent fuel and high level waste (HLW). Later work will deal with metadata collected on spent fuel and HLW as well as with different phases of repository development: design and construction, and operation. In order to develop metadata, RepMet members have provided lists of data items they currently collect and manage for packaged waste. The lists, albeit it non comprehensive, suggest that the number and type of data items varies importantly from organisation to organisation. At term, RepMet plans to generate a comprehensive list of such data items and their metadata. In parallel, RepMet is developing a document which examines a selected number of existing metadata standards. Because of the specialist nature of radioactive waste management, including waste processing facilities, disposal facilities and waste packages, existing metadata standards do not seem to be transferable as defined. Rather, RepMet will explore such standards to identify relevant principles for possible adaptation to the radioactive waste management community. (authors)

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DOE OSTI2023

Vitrification of Hanford Tank 241-AP-105 Waste at 7 M Na and Equivalent Simulant

Hanford Site nuclear waste is to be vitrified at the Waste Treatment and Immobilization Plant (WTP), which is a part of the safe and efficient retrieval, treatment, and disposal mission of the U.S. Department of Energy Office of River Protection. Hanford tank 241-AP-105 (referred to herein as AP-105) is one of the initial Hanford radioactive tank wastes planned to be processed and vitrified. A portion of AP-105 waste was retrieved by Washington River Protection Solutions, LLC (WRPS) and transferred to Pacific Northwest National Laboratory (PNNL). The waste went through dilution by Columbia River water to reach a target sodium (Na) concentration of 7 M, solids filtration, and cesium removal by ion exchange. A glass composition was calculated from the Kim et al. glass models to satisfy the WTP baseline requirements based on the as-received sample and the target dilution to 7 M, from which a simulant was calculated and glass forming chemical (GFC) additions were determined to form a liquid/solids mixture called melter feed. To prepare for the processing of the 7 M Na AP-105 waste melter feed and learn about the production expectations, the melter feed simulant of 7 M Na AP-105 waste was processed in a non-radioactive, continuous laboratory-scale melter (CLSM) system.

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