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DOE OSTI2016

Development of an Execution Strategy Analysis (ESA) Capability and Tool for the Management of Spent Nuclear Fuel (SNF) - 16575

An Execution Strategy Analysis (ESA) capability and tool are being developed to evaluate alternative execution strategies for future deployment of a consolidated Interim Storage Facility using a consent-based siting process per the Administration's Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste (Strategy). Application of an ESA approach not only leverages but also goes beyond traditional project analysis tools. The ESA tool allows for ongoing performance assessment of the evolving project execution plan that takes into account significant assumptions, risks, and uncertainties throughout the project lifecycle. The ESA process and tool will be used to support the development of plans, budgets, and alternative execution/implementation strategies for meeting the goals of the Strategy. This paper describes the development of the ESA capability and tool and identifies the value of applying such a capability in implementing a long-term strategy for managing spent nuclear fuel and high-level radioactive waste. (authors)

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

Development of a Technical Standard for Radiological Characterization of DAW - 18062

Throughout its 70+-year history, the U.S. Department of Energy (DOE) and its predecessor agencies have performed a wide mission involving nuclear materials and weapons development, production, testing, and modernization. While the initial mission components relied heavily upon reactor physics and chemistry, the environmental programs have increasingly relied upon development of more accurate and effective measurement methodologies to determine the radiological and chemical properties of the unique waste streams that were generated during the weapons production and testing mission. Over time, the DOE laboratories and former production sites have developed a wide range of site-specific processes and procedures for conducting radiological characterization of dry active waste (DAW) that is managed as low-level radioactive waste. This has resulted in a Complex-wide variability among sites with respect to the methodologies and instrumentation used to characterize individual waste streams. The DAW-type waste was selected for development of a characterization standard since it is a simple, voluminous, and repetitive waste stream with a substantial knowledge base, established experience history at multiple site locations, and the potential for realization of programmatic benefits though standardization. The Energy Facility Contractors Group (EFCOG) Waste Management Working Group (WMWG) has proposed the development of a new DOE Technical Standard for radiological characterization of DAW. The purpose of the proposed new standard is to provide a technical resource that integrates and enhances the utilization of current site-specific methodologies and processes. A Technical Basis Document (TBD) has been prepared to support the development and utilization of the new standard by providing detailed information for use by site personnel when implementing the guidance provided in the standard. It also provides the rationale for and details concerning the recommended technical approaches for radiological characterization of representative DAW-type waste streams. The TBD will be included as an appendix to the proposed standard. Adoption of a standardized approach to DAW characterization will take advantage of multiple applications for a technology that is both mature and stable. The DAW characterization process is repetitive and lends itself to implementation at a wide range of operating locations - resulting in economies of scale, compliance with regulatory requirements, and consistent implementation of procedures which support design, operations, maintenance, and management of facilities, activities and programs. The purpose of the proposed new Technical Standard for DAW Radiological Characterization is to integrate and improve the utilization of current site-specific methodologies and processes and will include a comprehensive evaluation and comparison of current radiological characterization methodologies to determine which ones are most reliable and effective in producing relevant and valid data. In view of the long history and participation by several agencies in the regulation of nuclear programs, existing Voluntary Consensus Standards have also been reviewed for relevancy and applicability to the new proposed standard. The proposed new Standard will be developed in concert with existing scientific and regulatory guidance to produce a document that incorporates the state of current knowledge and site practices with a graded approach to applying those methodologies that are applicable and effective for characterizing a wide range of DAW-type wastes - including, but not limited to, debris, low specific activity (LSA) material, and surface contaminated objects (SCO). Under the leadership of the National Laboratory Subgroup of the EFCOG/WMWG), an EFCOG team will develop the proposed technical standard. This process will incorporate the information gathered during preparation of the TBD, which will serve as a technical resource to support the technical approach for selection of DAW characterization methodologies. Preparation of the TBD included a detailed survey of historical site practices and examination of the characterization approaches and measurement methodologies currently being applied to DAW-type wastes. Future activities include the development and submittal of a Project Justification Statement that will be submitted to the DOE Technical Standards Program for review and comment. The EFCOG/WMWG Team will respond to reviewer comments, as required. Following final concurrence by affected organizations through the DOE Technical Standards Coordination process, the EFCOG/WMWG Team will prepare the draft technical standard and respond to comments generated during RevCom review and final format review by the DOE Technical Standards Management Program. (authors)

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

Residual waste detection in HLW tanks - 15701

This research uses commercial sonar technology to monitor residual waste in the U.S. Department of Energy's (DOE) Hanford Site high-level-waste staging tanks, with primary focus on the detection and imaging of the settled solids at specified areas of interest along the tank surface within a limited amount of time. High-level waste (HLW) is generated at the US-DOE Hanford Site as a by-product from the processing of nuclear materials. Waste is currently stored in 177 underground carbon-steel waste tanks. Pulse jet mixers (PJM) are often used to break loose and suspend the solids that have settled at the bottom of the tanks. A typical tank can have 6-8 pulse tubes. The Solid Liquid Interface Monitor (SLIM) must be capable of measuring dynamic movements in Region 1 zone between each PJM cycle. SLIM was originally developed to map the sludge/supernate interface within the U.S. DOE Hanford Site's HLW tanks. SLIM is a dual-axis mechanically rotating sonar and consists of an acoustic transducer, rotation motor with gearbox, tilt motor, position reference sensor, and pressure balancing mechanisms and its associated deployment platform. The sonar gathers a swath of data in the horizontal plane and rotates the transducer through a programmed angle around the vertical axis. Oil-filled device. No user serviceable parts. No semiconductors. Hard anodized aluminum and polyurethane delrin. May be deployed in areas of significant radioactivity. Kaolin will be used during mixing operations in order to simulate the physical properties of the expected slurry. In order to reduce the amount of time needed to image and object, all settings were decreased to their minimal state. The low resolution settings provide insufficient data for the manufacturer's 3D profiler to compile a visible point cloud and much less to allow the meshing algorithm to work properly. Several filtering algorithms have been applied to each scan in order to remove erroneous spikes and ghosting effects. Data has been collected for the mixing process of kaolin. This data consists of scans executed while the pump was active and inactive. The imaging algorithm has also been used with data derived from scans of 3% or less of kaolin. Future work: Mixing of kaolin within an experimental tank will be conducted in order to simulate the settling time of the HLW within the tank. The same filtering algorithms will be applied to each of these scans to evaluate the SLIM's efficiency to image solids at the bottom of a tank through increasing volume of kaolin.

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

Caustic Recycling Pilot Unit to Separate Sodium from LLW at Hanford Site - 12279

As part of the Department of Energy (DOE) sponsored Advanced Remediation Technologies initiative, a scheme was developed to combine Continuous Sludge Leaching (CSL), Near-Tank Cesium Removal (NTCR), and Caustic Recycling Unit (CRU) using Ceramatec technology, into a single system known as the Pilot Near-Tank Treatment System (PNTTS). The Cesium (Cs) decontaminated effluent from the NTCR process will be sent to the caustic recycle process for recovery of the caustic which will be reused in another cycle of caustic leaching in the CSL process. Such an integrated mobile technology demonstration will give DOE the option to insert this process for sodium management at various sites in Hanford, and will minimize the addition of further sodium into the waste tanks. This allows for recycling of the caustic used to remove aluminum during sludge washing as a pretreatment step in the vitrification of radioactive waste which will decrease the Low Level Waste (LLW) volume by as much as 39%. The CRU pilot process was designed to recycle sodium in the form of pure sodium hydroxide. The basis for the design of the 1/4 scale pilot caustic recycling unit was to demonstrate the efficient operation of a larger scale system to recycle caustic from the NTCR effluent stream from the Parsons process. The CRU was designed to process 0.28 liter/minute of NTCR effluent, and generate 10 M concentration of 'usable' sodium hydroxide. The proposed process operates at 40 deg. C to provide additional aluminum solubility and then recover the sodium hydroxide to the point where the aluminum is saturated at 40 deg. C. A system was developed to safely separate and vent the gases generated during operation of the CRU with the production of 10 M sodium hydroxide. Caustic was produced at a rate between 1.9 to 9.3 kg/hr. The CRU was located inside an ISO container to allow for moving of the unit close to tank locations to process the LLW stream. Actual tests were conducted with the NTCR effluent simulant from the Parsons process in the CRU. The modular CRU is easily scalable as a standalone system for caustic recycling, or for NTTS integration or for use as an In-Tank Treatment System to process sodium bearing waste to meet LLW processing needs at the Hanford site. The standalone pilot operation of the CRU to recycle sodium from NTCR effluent places the technology demonstration at TRL level 6. Multiple operations were performed with the CRU to process up to 500 gallons of the NTCR effluent and demonstrate an efficient separation of up to 70 % of the sodium without solids precipitation while producing 10 M caustic. Batch mode operation was conducted to study the effects of chemistry variation, establish the processing rate, and optimize the process operating conditions to recycle caustic from the NTCR effluent. The performance of the CRU was monitored by tracking the density parameter to control the concentration of caustic produced. Different levels of sodium were separated in tests from the effluent at a fixed operating current density and temperature. The voltage of the modules remained stable during the unit operation which demonstrated steady operation to separate sodium from the NTCR effluent. The sodium transfer current efficiency was measured in testing based on the concentration of caustic produced. Measurements showed a current efficiency of 99.8% for sodium transfer from the NTCR effluent to make sodium hydroxide. The sodium and hydroxide contents of the anolyte (NTCR feed) and catholyte (caustic product) were measured before and after each batch test. In two separate batch tests, samples were taken at different levels of sodium separation and analyzed to determine the stability of the NTCR effluent after sodium separation. The stability characteristics and changes in physical and chemical properties of the NTCR effluent chemistry after separation of sodium hydroxide as a function of storage time were evaluated. Parameters such as level of precipitated alumina, total alkalinity, analysis of Al, Na, K, Cs, Fe, OH, nitrate, nitrite, total dissolved and undissolved solids, viscosity, density, and other parameters of the NTCR effluent were measured. Changes in rheology and properties of NTCR stream to support downstream handling of the effluent after sodium separation was the basis for the analysis. The results show that the NTCR effluent is stable without the precipitation of aluminum hydroxide after 70% of the sodium was separated from the effluent. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Process optimization of caustic scrubber and iodine-129 immobilization in sodalite-based waste forms

Caustic scrubbers are proposed to remove vaporized iodine from nuclear fuel reprocessing plants. Resulting caustic slurries of radioactive iodine ( 129 I) and other anions can be captured and immobilized into the cage structures of sodalite-type minerals. Here, mixed anion sodalites were hydrothermally synthesized from a simulated caustic scrubber solution containing hydroxide/water, carbonate, chlorine, bromine, iodine, nitrate, and nitrite ions with added kaolinite. Experiments were conducted with twenty-five different sets of process conditions. Resulting powders were characterized by quantitative X-ray diffraction (XRD) and multi-functional thermal analysis, including gravimetry, calorimetry, and mass spectrometry. Using statistical analysis, process variables were tailored to maximize conversion of kaolinite into cage silicates, immobilize caustic scrubber anions, and minimize amorphous content. Further, chemical analysis showed the incorporation of all the targeted ions into the solids. Detailed investigation of a batch characterized after washing (creating secondary waste) and one without washing but subsequent calcination is given. This work contributes to defining the constraints required for subsequent consolidation process steps of the waste form to avoid volatilization.

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

Overview of Novel Canadian Approaches for Decontamination of Solid Radioactive and Mixed Waste - 19476

The availability of decontamination-for-release technology options provide a path forward to economically decontaminate radioactive metallic waste and enable its disposal as clean scrap metal, to decontaminate tools/equipment for unrestricted refurbishment or reuse, and to carry out inspection of in-service components. Projects involving management of solid mixed waste pose complex technical challenges that require the development and application of novel decontamination processes. The objective of this paper is to present project case studies that provide an overview of Kinectrics' experience with the decontamination of solid radioactive and mixed waste. Novel approaches to decontamination, recent developments, key results and lessons learned will also be discussed. (authors)

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

Final Report For PDRD SR16009: Durable Water Splitting Using Thermochemical Cycles Of Nanostructured Metal Oxides

The Tritium Facility process for stripping tritium from the glovebox atmosphere currently relies on the use of hot magnesium beds (mag beds) to “crack” tritiated water, releasing hydrogen isotopes as a gas. This is a safe and effective process for recovering hydrogen isotopes from water for further processing. However, due to the thermodynamic properties of magnesium, mag beds are single use - once the magnesium metal is converted to magnesium oxide the bed must be replaced. Replacing mag beds is a labor-intensive and operationally complex task. In order to streamline operations and reduce costs, it is desirable to develop a reusable system that recovers hydrogen isotopes from water, while safely sequestering oxygen in a stable form. The following document reviews the results from the Plant Directed Research, Development, and Demonstration (PDRD) project “Durable Water Splitting using Thermochemical Cycles of Nanostructured Metal Oxides” (SR16009), which had the specific goal of developing a process for stripping tritium from the glovebox atmosphere that eliminates magnesium beds.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Why SRS Matters - K Area

A video series presenting an overview of the Savannah River Site (SRS) mission and operations. Each episode features a specific area/operation and how it contributes to help make the world safer. This episode features K Area's mission and operations. K area is the former production reactor that's been re-purposed to serve as a plutonium processing and storage facility.

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

Calcined Waste Storage at the Idaho Nuclear Technology and Engineering Center

This comprehensive report provides definitive information regarding the volume, mass, and composition (chemical and radioactivity) of the calcined waste stored at the Idaho Nuclear Technology and Engineering Center on the Idaho National Laboratory Site. The calcine composition data are required for current regulatory compliance and safe operation of the INTEC’s six Calcined Solids Storage Facilities (CSSFs), as well as future retrieval, treatment, and shipping the calcine to an off-Site facility (such as a geologic repository). This report also contains a description of the calcine storage bins that compose each of the CSSFs, emphasizing the features that may be of concern to future calcine retrieval. The CSSFs were designed by different architectural engineering firms and built at different times. Each CSSF has a unique design, reflecting varying design criteria and lessons learned from historical CSSF operation. The varying CSSF designs will affect future calcine retrieval processes and equipment. This revision (5) of the report presents refinements of the databases used to calculate the volume, mass, chemical content, and radioactivity of calcined waste produced and stored within the CSSFs. This revision corrects minor errors and inconsistencies noted in an extensive validation and verification effort of the prior version of the Microsoft Excel workbooks upon which the calcine composition described in this report is based. Historical calcine samples are insufficient in number and scope of analysis to fully characterize the entire inventory of calcine in the CSSFs. Sample data exist for all of the liquid wastes that were calcined. This report provides calcine composition data based on liquid waste sample analyses, the volume of liquid waste calcined, calciner operating data, and CSSF operating data using several large Microsoft Excel (Microsoft 2010) databases and spreadsheets. The Excel spreadsheets are collectively called the Historical Processing Model (HPM). The calcine composition determined by this method compares favorably with historical calcine sample data.

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

Potential scenarios for broadening stakeholder involvement in the implementing geological disposal technology platform

This paper analyses the potential for the involvement of different types of stakeholders in the Implementing Geological Disposal Technology Platform (IGD-TP). This analysis was conducted as part of the InSOTEC project, a three-year (2011- 2014) collaborative research project funded under the 7. Euratom Framework Programme (Grant Agreement nr. 269906). In our analysis, we consider the extent to which the IGDTP's practice as regards to stakeholder involvement matches its discourse, and what potential for improvement exists given its structural organisation as a European Technology Platform (ETPs). Technology Platforms (TPs) can be understood as knowledge networks, deliberately set up to influence (research) policy in a specific domain. We therefore use knowledge networks as a conceptual approach and look at the IGD-TP as a complex network which includes actors, knowledge and practices across different countries, focusing on a very specific topic (i.e. implementing geological disposal). We compare the way different stakeholders are involved in the IGD-TP to the practice of other ETPs, and explore how the IGD-TP is viewed by its members and by outsiders to the platform Applying Callon's framework of knowledge co-production (1999) we come to define different degrees of interaction between science, society and policy in view of defining research and development (R and D) priorities [1]. Subsequently we describe how these interactions could be conceptualised and interpreted for the IGD-TP. The current approach of the IGDTP can be mainly understood as classical model involving mainly expert stakeholders and scientists. Where there seems to be a good representation among IGD-TP members of industry, research institutes, and some members of the academic community this is not the case for other types of stakeholders, such as public authorities or civil society. At this stage, the overall approach of the IGD-TP would seem to restrict the scope of stakeholder involvement, as it narrows participation down to uniquely technology experts, hindering socio-technical manifestations. Our analysis nevertheless shows that there is room for engaging with a broader range of stakeholders in the field of radioactive waste management, if this is the intention of the IGD-TP. However, this would require a commitment to developing a common knowledge base including other stakeholders through a process of mutual adjustment and negotiation. (authors)

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

Durable Tellurite Glasses for Vitrification of Technetium Wastes - 18270

Technetium-99 (Tc-99) can bring a significant environmental risk because it has a long half-life and high fission yield. The present work has developed alumino tellurite glass to immobilize Tc wastes generated by the pyrochemical processing technology. Rhenium (Re) was used as a non-radioactive surrogate for Tc. The Re{sub 2}O{sub 7} concentration in the tellurite glasses was as high as 9.11 mass% with retention of 86% on average. X-ray absorption spectroscopy analysis revealed that Re exists as Re{sup 7+} and is surrounded by 4 oxygen atoms. The 7-day PCT results were satisfied with the immobilized low-activity waste regulatory limit with Ca(ReO{sub 4}){sub 2} content less than 12 mass%. The long-term dissolution test was conducted for 256 days and a rate drop has occurred in which the rate of leaching gradually decreased. An alteration layer depleted soluble element was observed on the leached glass surface. These results suggest that tellurite glass can be a good candidate as a matrix for immobilizing Tc. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Radioactive Waste Sludge Washing and Demonstration of the Nitric-Glycolic Acid Flowsheet for Sludge Batch 10 Qualification

For each sludge batch that is processed in the Defense Waste Processing Facility (DWPF), the Savannah River National Laboratory (SRNL) performs qualification testing to demonstrate that the sludge batch (SB) is processible. During processing of SB9, DWPF will be transitioning from the Nitric-Formic Acid (NFA) flowsheet to the Nitric-Glycolic Acid (NGA) flowsheet. Thus, the qualification of SB10 was requested to only be performed using the NGA flowsheet. In order to qualify the batch for the NGA flowsheet, Sludge Receipt and Adjustment Tank (SRAT) and Slurry Mix Evaporator (SME) cycles, designated SC-19, were performed using SB10 Tank 51 sample material. SRNL received Tank 51 material in the midst of Tank Farm washing. SRNL continued the washing in the SRNL Shielded Cells. The SRNL process included the addition of Sodium Reactor Experiment (SRE) material from H Canyon, simulating the transfer of SRE from H Canyon to Tank 51 during Tank Farm washing. The washed SB10 Tank 51 material, with SRE, was characterized prior to flowsheet qualification testing.

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

lNAC's OPTIMUS{sup TM}-L Package - the Next Transformation - 19396

NAC has developed the OPTIMUS{sup TM}-L package, the newest addition to the OPTIMUS{sup TM} packaging product line. OPTIMUS{sup TM}-L is a Type B(U)F-96 transportation package designed for maximum flexibility and cost-efficiency to support shipment of a wide range of challenging low-activity wastes. The OPTIMUS{sup TM}-L package utilizes the same cask containment vessel (CCV) as its predecessor, OPTIMUS{sup TM}-H, packaged inside lightweight outer packaging. The large cavity size of the CCV, which can accommodate a 110-gallon drum, combined with the small size and modularity of the OPTIMUS{sup TM}-L package provides unmatched flexibility for the user. OPTIMUS{sup TM}-L can accommodate more radioactive waste in each drum (up to a Fissile Gram Equivalent (FGE) limit of 395 g Pu-239) than larger packages can in 10 drums. Furthermore, because of its small size and weight, up to ten (10) OPTIMUS{sup TM}-L packages can be accommodated per legal-weight truck shipment. Thus, one Legal Weight Truck (LWT) shipment of 10 OPTIMUS{sup TM}-L packages can accommodate more than 10 times the FGE as a large package containing 10 drums. OPTIMUS{sup TM}-L contents include low level waste, mixed low-level waste, intermediate level waste and contact handled transuranic (CH-TRU) waste. Furthermore, all contents can include aerosol cans with compressed gas and liquefied gas propellants and standard Department of Transportation (DOT) 3E lecture bottles for shipment to treatment facilities. The OPTIMUS{sup TM}-L is also designed to accommodate irradiated fuel debris and other fissile material waste, which may be packaged inside shield inserts if required for supplemental radiation shielding. In this paper, NAC provides a technical overview of the OPTIMUS{sup TM}-L package and identifies the design features and technology advancements that makes the OPTIMUS{sup TM} product line a readily adaptable and flexible solution for packaging DOE and other orphan wastes and enables the integration of these wastes into the waste management system process and disposition infrastructure. (authors)

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

Modeling Streamflow and Sediment Transport Under Current and Future Climates at the West Valley Site - 20231

Performance Assessment models often require inputs representing specific hydrologic processes such as streamflow and sediment flux. This paper details the development of a process-level hydrologic model and its use in a Probabilistic Performance Assessment (PPA) system-level model for the West Valley Site in western New York. For this PPA model, a watershed model was developed using SWAT (Soil and Water Assessment Tool) and used to evaluate water and sediment fluxes in nearby streams draining the Site. A particular challenge of probabilistic modeling over long periods of time is scaling parameters appropriately so that outcomes are not grossly over- or underestimated. This challenge is especially important when the spatial and temporal scales of process and system level models differ. In this instance, the SWAT watershed model and GoldSim PPA model have similar spatial scales of stream reaches and sub-catchments within a single watershed (approximately hundreds to tens of thousands of square meters, but the temporal scales differ considerably. The SWAT model calculates streamflow and sediment transport rates at the daily scale, while the PPA model requires long-term average values that are applied over hundreds to thousands of year-long periods. As such, distributions of daily climatic parameters such as average precipitation rate, temperature, wind speed, and relative humidity were developed to force a transient SWAT model that captures the dynamic behavior of the watershed in response to daily changes in the weather. Output from the model, which includes stream flow (m{sup 3}/y) and sediment transport rate (Mg/y) is then averaged over a 100-year period as an approximation for the long-term average values which can be applied in GoldSim. A single SWAT model was run 5000 times with varying climatic inputs and deterministic soil, land use, and elevation data, and the range of outputs were compiled into a distribution which can be implemented stochastically into the PPA Model. Additionally, this paper includes a high-level discussion of how changes in future climate affect weather input distributions, and a summary of how those changes in weather affect the outcomes from the SWAT model and the inputs into the PPA model. In this analysis, future climate distributions were developed based on climate research documenting the expected changes to precipitation and temperature in western New York. The 5,000 SWAT simulations were then repeated with these future climate distributions, and the resulting output was implemented into the PPA Model as hydrologic conditions under a future climate. In the current version of the PPA Model, hydrologic parameters change linearly from the 'current climate' conditions to the 'future climate' conditions over a period of 100 years, after which they remain constant. (authors)

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

Cradle to grave: the importance of the fuel cycle to molten salt reactor sustainability

Advanced reactor technologies are being considered for the next-generation of nuclear power plants. These plants are designed to have a smaller footprint, run more efficiently at higher temperatures, have the flexibility to meet specific power or heating needs, and have lower construction costs. This paper offers a perspective on molten salt reactors, promoted as having a flexible fuel cycle and close-to-ambient pressure operation. A complexity introduced by reducing the reactor footprint is that it may require low-enriched fuel for efficient operation, available from enrichment of the feed salt or by reusing actinides from existing used nuclear fuel (UNF). Recycling UNF has the potential to reduce high-level waste, if done correctly. Release limits from UNF processing are stringent, and processes for waste reduction, fission gas trapping, and stable waste-form generation are not yet ready for commercial deployment. These complex processes are expensive to develop and troubleshoot because the feed is highly radioactive. Thus, fuel production and supply chain development must keep abreast of reactor technology development. Another aspect of reactor sustainability is the non-fuel waste streams that will be generated during operation and decommissioning. Some molten salt reactor designs are projected to have much shorter operational lifetimes than light-water reactors: less than a decade. A goal of the reactor sustainability effort is to divert these materials from a high-level waste repository. However, processing of reactor components should only be undertaken if it reduces waste. Economic and environmental aspects of sustainability are also important, but are not included in this perspective.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Management of Metals from Decommissioning of Nuclear Facilities - 18411

The EDF group companies within the Cyclife platform have for decades performed treatment and recycling of metals. With different approaches but all considered international Good Practice, as valuable materials are recovered for safe recycling and volumes in the repositories for radioactive waste are saved. For an efficient decommissioning project, a well-defined waste management process is one of the critical paths, not least when it comes to routes for clearance and recycling of materials. It is easy to underestimate the complexity and bottlenecks caused by shortage of logistic space. In order to mitigate these risks, activities such as strengthening categorization, quantification and adopting different routes based on the material properties should be implemented. By categorization of the of waste and materials considered/classified as contaminated or potentially contaminated it is obvious that the management of the metals is a key parameter in nuclear decommissioning. Analyses performed by the EDF companies shows that a majority of the materials, classified as contaminated, from a BWR/PWR plant are metals. Experience gives that a high percentage of these metals from decommissioning can be subject to recycling either directly upon completion of the clearance process or after certain treatment. Non-technical key parameters like the regulatory requirements on applying the waste hierarchy, the impact on repository programs as well as, not at least, the regulatory and public acceptance of recycling of materials from the nuclear industry are, together with costs essential factors when forming the routing of metals and other waste materials from a decommissioning project. Today the EDF group companies are positioned to provide a wide range of services for management of the contaminated metals from decommissioning projects, spanning from industrial management of metals which directly will be subject general clearance after verification measurements up to decontamination and melting of selected reactor internals for recycling to public domain or within the nuclear industry. In this paper the typical waste material composition for BWR and PWR plants, the different ways to treat/manage the radioactive and potentially radioactive metals including a comparison of the different approach commonly used in a technical, logistical and economical perspectives. A few success stories will be resented as well as lessons learned. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2011

ISOLOK VALVE ACCEPTANCE TESTING FOR DWPF SME SAMPLING PROCESS

Evaluation of the Defense Waste Processing Facility (DWPF) Chemical Process Cell (CPC) cycle time identified several opportunities to improve the CPC processing time. Of the opportunities, a focus area related to optimizing the equipment and efficiency of the sample turnaround time for DWPF Analytical Laboratory was identified. The Mechanical Systems & Custom Equipment Development (MS&CED) Section of the Savannah River National Laboratory (SRNL) evaluated the possibility of using an Isolok{reg_sign} sampling valve as an alternative to the Hydragard{reg_sign} valve for taking process samples. Previous viability testing was conducted with favorable results using the Isolok sampler and reported in SRNL-STI-2010-00749 (1). This task has the potential to improve operability, reduce maintenance time and decrease CPC cycle time. This report summarizes the results from acceptance testing which was requested in Task Technical Request (TTR) HLW-DWPF-TTR-2010-0036 (2) and which was conducted as outlined in Task Technical and Quality Assurance Plan (TTQAP) SRNL-RP-2011-00145 (3). The Isolok to be tested is the same model which was tested, qualified, and installed in the Sludge Receipt Adjustment Tank (SRAT) sample system. RW-0333P QA requirements apply to this task. This task was to qualify the Isolok sampler for use in the DWPF Slurry Mix Evaporator (SME) sampling process. The Hydragard, which is the current baseline sampling method, was used for comparison to the Isolok sampling data. The Isolok sampler is an air powered grab sampler used to 'pull' a sample volume from a process line. The operation of the sampler is shown in Figure 1. The image on the left shows the Isolok's spool extended into the process line and the image on the right shows the sampler retracted and then dispensing the liquid into the sampling container. To determine tank homogeneity, a Coliwasa sampler was used to grab samples at a high and low location within the mixing tank. Data from the two locations were compared to determine if the contents of the tank were well mixed. The Coliwasa sampler is a tube with a stopper at the bottom and is designed to obtain grab samples from specific locations within the drum contents. A position paper (4) was issued to address the prototypic flow loop issues and simulant selections. A statistically designed plan (5) was issued to address the total number of samples each sampler needed to pull, to provide the random order in which samples were pulled and to group samples for elemental analysis. The TTR required that the Isolok sampler perform as well as the Hydragard sampler during these tests to ensure the acceptability of the Isolok sampler for use in the DWPF sampling cells. Procedure No.L9.4-5015 was used to document the sample parameters and process steps. Completed procedures are located in R&D Engineering job folder 23269.

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

Characterization of Tank 16 Annulus Samples

The closure of Tank 16H will require removal of material from the annulus of the tank. Samples from Tank 16H annulus were characterized and tested to provide information to evaluate various alternatives for removing the annulus waste. The analysis found all four annulus samples to be composed mainly of Si, Na, and Al and lesser amounts of other elements. The XRD data indicate quartz (SiO 2 ) and sodium aluminum nitrate silicate hydrate (Na 8 (Al 6 Si 6 O 24 )(NO 3 ) 2 .4H 2 O) as the predominant crystalline mineral phases in the samples. The XRD data also indicate the presence of crystalline sodium nitrate, sodium nitrite, gibbsite, hydrated sodium bicarbonate, and muscovite. Based on the weight of solids remaining at the end of the test, the water leaching test results indicate approximately 20-35% of the solids dissolved after three contacts with an approximately 3:1 volume of water at 45 C. The chemical analysis of the leachates and the XRD results of the remaining solids indicate sodium salts of nitrate, nitrite, sulfate, and possibly carbonate/bicarbonate make up the majority of the dissolved material. The majority of these salts were dissolved in the first water contact and simply diluted with each subsequent water contact. The water leaching removed large amounts of the uranium in two of the samples and {approx}1/3 of the {sup 99}Tc from all four samples. Most of the other radionuclides analyzed showed low solubility in the water leaching test. The preliminary data on the oxalic acid leaching test indicate the three acid contacts at 45 C dissolved from {approx}34-47% of the solids. The somewhat higher dissolution found in the oxalic acid leaching test versus the water leaching test might be offset by the tendency of the oxalic acid solutions to take on a gel-like consistency. The filtered solids left behind after three oxalic acid contacts were sticky and formed large clumps after drying. These two observations could indicate potential processing difficulties with solutions and solids from oxalic acid leaching. The gel formation might be avoided by using larger volumes of the acid. Further testing would be recommended before using oxalic acid to dissolve the Tank 16H annulus waste to ensure no processing difficulties are encountered in the full scale process.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗