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DOE OSTI2024

Layered Double Hydroxide Derivatives for Polyolefin Upcycling

While Ru-catalyzed hydrogenolysis holds significant promise in converting waste polyolefins into value-added alkane fuels, a major constraint is the high cost of noble metal catalysts. Here, in this work, we propose, for the first time, that Co-based catalysts derived from CoAl-layered double hydroxide (LDH) are alternatives for efficient polyolefin hydrogenolysis. Leveraging the chemical flexibility of the LDH platform, we reveal that metallic Co species serve as highly efficient active sites for polyolefin hydrogenolysis. Furthermore, we introduced Ni into the Co framework to tackle the issue of restricted hydrogenation ability associated with contiguous Co-Co sites. In-situ analysis indicates that the integration of Ni induces electron transfer and facilitates hydrogen spillover. This dual effect synergistically enhances the hydrogenation/desorption of olefin intermediates, resulting in a significant reduction in the yield of low-value CH 4 from 27.1 to 12.6%. Through leveraging the unique properties of LDH, we have developed efficient and cost-effective catalysts for the sustainable recycling and valorization of waste polyolefin materials.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2024

The simultaneous removal of technetium and iodine from Hanford tank waste

The simultaneous removal of radionuclides technetium-99 and iodine-129 from an actual decontaminated Hanford tank waste sample (a mixture of decontaminated waste from tanks 241-AP-105 and 241-AP-107) was demonstrated for the first time in this work. A series of commercially available ion exchange resins were evaluated in batch contact tests in the tank waste, and all showed removal of both Tc and I. The highest Tc removal was observed for Purolite A530e while the highest iodine removal was observed for ResinTech SIR-110-MP. Batch tests in simulated tank waste with these two resins showed that the SIR-110-HP-MP had consistently higher K d for both pertechnetate and iodide and much higher K d than previous works on Tc removal from Hanford waste. As such, the SIR-110-MP was evaluated in a dual -column (lead/lag) test processing 5.2L of the tank waste mixture showing 60% breakthrough of Tc on the lead column and no significant breakthrough on the lag after 625 bed volumes (BV, 6 mL size) while significant iodine breakthrough (>50%) occurred after 28 BV. The limited iodine uptake was attributed to the column conditions generating mass transfer limitations. A fraction of the Tc and I was not captured by the resin (<10%) in either the batch tests or column tests. The iodine fraction was identified to be an iodide, likely organo-iodide. The fraction of the Tc was identified as a non-pertechnetate species, which is the first time non-pertechnetate has been identified in AP-105 and AP-107 tanks, although the exact species is still unknown.

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

Managing work flow and logistic at waste processing facilities using the waste compliance and tracking system - 15392

Work flow and logistics at a busy radioactive waste processing complex or permitted treatment, storage, and disposal facility (TSDF) can pose a monumental challenge. An efficient operation requires good communication between engineering, logistics, and waste operators. The Los Alamos National Laboratory (LANL) has implemented an application that facilitates seamless communication of work flow requirements, between engineering, logistics, and waste operations personnel in the field. The Waste Compliance and Tracking System (WCATS), a comprehensive waste management system designed at LANL, provides four key elements to support logistics: (a) the container work path, (b) directed unit operation requests, (c) notification of pending operations, and (d) real-time status tools for office and field workers. The combination of these elements allows everyone to function as a team, with a shared knowledge of what is needed to accomplish facility objectives. The first element is like a road map indicating which unit operations are needed and in what sequence they should be performed; this is referred to as a 'work path'. It automatically routes a waste item or container from operation to operation. For example, a given work path might indicate that waste items require radiological survey, radioassay, and a technical review, in that order. When the work path is assigned to a container, the system enforces the correct order of operations. Directed unit operation 'requests', the second element, fills a gap between pre-planned, formal 'work path' activities and the need for flexibility to direct preparatory operations or deal with unexpected issues, like staging or mining containers in preparation for a shipment or transferring containers from one storage unit to another to clear a path for building maintenance. Logistics engineers prepare these 'requests', and the system monitors the status and completion of the work. The third element, notifications, allows team members to subscribe to and receive notice that an operation is pending, initiated, or complete. This functionality supports a range of communication options, including e-mail, cell phone, pager, or application in-box messaging. Notifications allow waste operations personnel to multi-task, and remain in the loop on the status of facility-wide activities. Finally, the fourth element provides tools that communicate work path pending operations and directed requests to field workers on their mobile computer, allowing them to mine or pick containers from storage units, and leverage the capabilities of a bar-code scanner to avoid selection mistakes. Oversight and summary tools are provided, allowing logistics personnel to monitor work path and request task status, identify work-flow bottlenecks, and maximize the efficiency of waste operations. The Waste Compliance and Tracking System (WCATS) provides an innovative, comprehensive set of tools that allow large-scale radioactive waste processing facilities to manage logistics, maximize efficient use of the workforce, and create a team environment, where everyone from management to the waste handlers have an understanding of the work load and what needs to be accomplished. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Structure and Potential–Dependent Selectivity in Redox–Metallopolymers: Electrochemically Mediated Multicomponent Metal Separations

Electro-responsive functional materials can play a critical role in selective metal recovery and recycling due to the need for molecular differentiation between transition metals in complex mixtures. Redox-active metallopolymers are a promising platform for electrochemical separations, offering versatile structural tuning and fast electron transfer. First, through a judicious selection of polymer structure between a main-chain metallopolymer (polyferrocenylsilane) and a pendant-group metallopolymer (polyvinylferrocene), charge-transfer interactions and binding strength toward competing metal ions are tuned, which as a result, dictate selectivity. For example, almost an order of magnitude increase in separation factor between chromate and meta-vanadate can be achieved, depending on polymer structure. Second, these metallopolymer electrodes exhibit potential-dependent selectivity that can even flip ion preference, based solely on electrical means—indicating a control parameter that is orthogonal to structural modifications. Finally, this work presents a framework for evaluating electrochemical separations in multicomponent ion mixtures and elucidates the underlying charge-transfer mechanisms resulting in molecular selectivity through a combination of spectroscopy and electronic structure calculations. Furthermore, the findings demonstrate the applicability of redox-metallopolymers in tailored electrochemical separations for environmental remediation, value-added metal recovery, waste recycling, and even mining processing.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2021

Thermodynamic non-ideality and disorder heterogeneity in actinide silicate solid solutions

Non-ideal thermodynamics of solid solutions can greatly impact materials degradation behavior. We have investigated an actinide silicate solid solution system (USiO 4 –ThSiO 4 ), demonstrating that thermodynamic non-ideality follows a distinctive, atomic-scale disordering process, which is usually considered as a random distribution. Neutron total scattering implemented by pair distribution function analysis confirmed a random distribution model for U and Th in first three coordination shells; however, a machine-learning algorithm suggested heterogeneous U and Th clusters at nanoscale (~2 nm). The local disorder and nanosized heterogeneous is an example of the non-ideality of mixing that has an electronic origin. Partial covalency from the U/Th 5f–O 2p hybridization promotes electron transfer during mixing and leads to local polyhedral distortions. The electronic origin accounts for the strong non-ideality in thermodynamic parameters that extends the stability field of the actinide silicates in nature and under typical nuclear waste repository conditions.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2022

Supramolecular Control of Redox-Mediated Interactions for Ion-Selective Electroseparations

This report summarizes efforts from the Su Group at the University of Illinois at Urbana-Champaign in researching and developing materials and processes for the recovery of rare earth elements and other metal elements that are critical to the United States economy and national security, under DOE Award No. DE-SC0021409. This report highlights efforts in the recovery of: • Transition metal elements (including arsenic, chromium, molybdenum, rhenium, vanadium): Redox-active metallopolymers are able to capture these elements in their oxyanion form. Direct control of target ion capture/release through electrochemical-mediated switching of metallocene binding sites reduces chemical input and lowers generation of waste in element recovery processes. We study the effects of metallopolymer structure on targeted ion binding through a combination of experimental screening of ion selectivity from multi-component mixtures, computational efforts which elucidate the charge-transfer origins of metallocene binding and provide insight on how binding site structure influences selectivity, and state-of-the-art neutron reflectivity of metallocene polymer films to understand the interplay of polymer-ion (charge-transfer), polymer-solvent (polymer hydrophilicity/hydrophobicity), and ion-solvent (ion hydration) interactions with spatial resolution of polymer films. • Rare earth elements (including yttrium, neodymium, europium, gadolinium, dysprosium, cerium): Here we employ redox-copolymers for recovery of rare earth elements by electrochemically regenerated ion-exchange. Rare earth elements require an alternative route to capture/release than directly with metallopolymers since rare earth elements typically exist in cationic form in water. To address this, copolymerization of traditional ion-exchange moieties with metallocene moieties allowed for capture of rare earth elements with the ion-exchange moieties and electrochemical-mediated release of rare earth elements through oxidation of metallocene moieties.

36 MATERIALS SCIENCE↗
DOE OSTI2014

Beyond Human Capital Development: Balanced Safeguards Workforce Metrics and the Next Generation Safeguards Workforce

Since its establishment in 2008, the Next Generation Safeguards Initiative (NGSI) has achieved a number of objectives under its five pillars: concepts and approaches, policy development and outreach, international nuclear safeguards engagement, technology development, and human capital development (HCD). As a result of these efforts, safeguards has become much more visible as a critical U.S. national security interest across the U.S. Department of Energy (DOE) complex. However, limited budgets have since created challenges in a number of areas. Arguably, one of the more serious challenges involves NGSI’s ability to integrate entry-level staff into safeguards projects. Laissez fair management of this issue across the complex can lead to wasteful project implementation and endanger NGSI’s long-term sustainability. The authors provide a quantitative analysis of this problem, focusing on the demographics of the current safeguards workforce and compounding pressures to operate cost-effectively, transfer knowledge to the next generation of safeguards professionals, and sustain NGSI safeguards investments.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2019

Dose Calculation Methodology and Data for Solid Waste Performance Assessment and Composite Analysis at the Savannah River Site

This report provides a detailed description of the methodology developed to perform dose calculations for E-Area Low Level Waste Facility Performance Assessments and Savannah River Site Composite Analyses. Chapters 2 – 8 give a complete set of equations to be used to calculate dose for individual exposure pathways. The first set of dose scenarios assume that a member of the public establishes residence near the waste site and uses contaminated groundwater or contaminated surface water for personal consumption and to irrigate a garden and pasture where produce and farm animals are raised. Products from the garden and farm animals are used for personal consumption. This scenario applies to both E-Area Performance Assessments and Savannah River Site Composite Analyses. General dose exposure pathways for the resident farmer scenario are: 1. Ingestion pathways described in Chapter 2, 2. Inhalation pathways described in Chapter 3, 3. External exposure pathways described in Chapter 4, and 4. Recreational pathways described in Chapter 5 (Composite Analysis only). Doses from all of the above pathways are directly related to the contaminant concentration in the water. Additional dose scenarios considered for E-Area Performance Assessment assume that an inadvertent intruder encroaches on the waste disposal site after loss of institutional control. General dose exposure pathways considered for the inadvertent intruder are: 1. Ingestion pathways described in Chapter 6, 2. Inhalation pathways described in Chapter 7, and 3. External exposure pathways described in Chapter 8. Doses from all of the intruder pathways are directly related to the contaminant concentration in the buried waste. Special dose considerations related to human consumption of meat and dairy, and treatment of tritium in dose calculations are discussed in Chapter 9. Chapter 10 describes an Excel database named “Radionuclide, Element and Dose Parameter Data Package” that has been created to provide a system for maintaining the data required to perform radionuclide transport and dose calculations. This data includes: Isotope physical parameters and radioactive decay data needed for both transport and dose calculations, Isotope-specific dose coefficients and soil shielding factors associated with exposure pathways, Radionuclide drinking water concentration limits either published by the Environmental Protection Agency or derived from internal dose coefficients to meet the groundwater protection requirements in Department of Energy Order 435.1, Element-specific bio-transfer factors used in dose calculations, Dose equations and values of physical parameters and human factors used in dose calculations, and Key physical constants used in dose calculations. Revision 1 incorporates updated information developed since this report was issued for use in the next revision of the E-Area Low-Level Waste Facility Performance Assessment.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2019

A Practical Approach of Mining Operational Risks in Cigeo - A field-test - 19349

Andra is currently completing the detailed engineering phase of the stepwise design development of Cigeo (the French HLW and IL-LLW Deep Geological Repository, aka DGR). Various technical issues are at stake, including the risk management requirements linked to the safety of 'Mining Operations' (i.e. occupational safety issues related to underground excavation/dismantling works at some 500 m of depth), at different stages of the DGR life cycle (i.e. from its construction start-up to its final closure and future monitoring/or conversely a decision of waste retrieval). These occupational safety requirements have a significant impact on Cigeo's design, its progressive construction and later closure. On a shorter term, one must demonstrate that the design bases are technically sound and that the solutions selected (to build, operate, close the DGR infrastructures or conversely retrieve the waste) are both feasible and safe. In that sense (and following an identification of operational risks during construction), Andra focused on two sensitive cases: - The construction of large diameter drift crossings, in particular those related to the connection between a transfer drift (main) and a IL-LLW disposal vault (i.e. during construction operations), - The partial dismantling of a drift concrete lining (made of concrete segment rings, erected and placed by a TBM) at time of construction of a horizontal seal (i.e. during closure operations). The discussion starts with a description of the geotechnical conditions prevailing at the (French) Bure site, in the Callovo-Oxfordian (COx) claystone formation. Then the paper presents the Cigeo underground architecture (at large) and the specific structural choices made (at the end of the detailed engineering phase) for drifts, disposal vaults and seals. The discussion on construction - deconstruction challenges for each of the critical issues cited above provides an opportunity to explain the operational sequences (and how the methodology selected enables to reduce the risks). The following part of the paper focuses the technological demonstration (an in-situ test to be planned in detail) of a partial dismantling of the concrete segments rings forming the support lining of two tunnels previously built in the Meuse/Haute-Marne URL (Underground Research Laboratory) infrastructures, combined with the construction of a specific drift crossing. In this dismantling and construction test, the technical demonstration objective is three-fold: - The capacity to lay down compressible arch segments and support (by bolts and mesh) the rock after the lining removal, - The capacity to excavate and line the drift crossing, - The structural / geomechanical behavior (evolution with time) of the rock and lining compared with that of 'neighboring' zones. This paper provides a timeline of the tests. Should the results obtained deliver some positive perspectives of technical feasibility (and convincing operational safety) Andra will incorporate this dismantling/construction solution in the 'design freeze' of the DGR underground infrastructures. The positive outcomes of this technical test campaign will also provide 'elements of proof' for the instruction of the Cigeo licensing (a process likely to take place between 2020 and 2024) by the French Nuclear Safety Authority (ASN). (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2014

Conversion of Nuclear Waste into Nuclear Waste Glass: Experimental Investigation and Mathematical Modeling

The melter feed, slurry, or calcine charged on the top of a pool of molten glass forms a floating layer of reacting material called the cold cap. Between the cold-cap top, which is covered with boiling slurry, and its bottom, where bubbles separate it from molten glass, the temperature changes by up to 1000 K. The processes that occur over this temperature interval within the cold cap include liberation of gases, conduction and consumption of heat, dissolution of quartz particles, formation and dissolution of intermediate crystalline phases, and generation of foam and gas cavities. These processes have been investigated using thermal analyses, optical and electronic microscopies, x-ray diffraction, as well as other techniques. Properties of the reacting feed, such as heat conductivity and density, were measured as functions of temperature. Investigating the structure of quenched cold caps produced in a laboratory-scale melter complemented the crucible studies. The cold cap consists of two main layers. The top layer contains solid particles dissolving in the glass-forming melt and open pores through which gases are escaping. The bottom layer contains bubbly melt or foam where bubbles coalesce into larger cavities that move sideways and release the gas to the atmosphere. The feed-to-glass conversion became sufficiently understood for representing the cold-cap processes via mathematical models. These models, which comprise heat transfer, mass transfer, and reaction kinetics models, have been developed with the final goal to relate feed parameters to the rate of glass melting.

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

Elimination Of Catalytic Hydrogen Generation In Defense Waste Processing Facility Slurries

Based on lab-scale simulations of Defense Waste Processing Facility (DWPF) slurry chemistry, the addition of sodium nitrite and sodium hydroxide to waste slurries at concentrations sufficient to take the aqueous phase into the alkaline region (pH > 7) with approximately 500 mg nitrite ion/kg slurry (assuming <25 wt% total solids, or equivalently 2,000 mg nitrite/kg total solids) is sufficient to effectively deactivate the noble metal catalysts at temperatures between room temperature and boiling. This is a potential strategy for eliminating catalytic hydrogen generation from the list of concerns for sludge carried over into the DWPF Slurry Mix Evaporator Condensate Tank (SMECT) or Recycle Collection Tank (RCT). These conclusions are drawn in large part from the various phases of the DWPF catalytic hydrogen generation program conducted between 2005 and 2009. The findings could apply to various situations, including a solids carry-over from either the Sludge Receipt and Adjustment Tank (SRAT) or Slurry Mix Evaporator (SME) into the SMECT with subsequent transfer to the RCT, as well as a spill of formic acid into the sump system and transfer into an RCT that already contains sludge solids. There are other potential mitigating factors for the SMECT and RCT, since these vessels are typically operated at temperatures close to the minimum temperatures that catalytic hydrogen has been observed to occur in either the SRAT or SME (pure slurry case), and these vessels are also likely to be considerably more dilute in both noble metals and formate ion (the two essential components to catalytic hydrogen generation) than the two primary process vessels. Rhodium certainly, and ruthenium likely, are present as metal-ligand complexes that are favored under certain concentrations of the surrounding species. Therefore, in the SMECT or RCT, where a small volume of SRAT or SME material would be significantly diluted, conditions would be less optimal for forming or sustaining the catalytic ligand species. Such conditions are likely to adversely impact the ability of the transferred mass to produce hydrogen at the same rate (per unit mass SRAT or SME slurry) as in the SRAT or SME vessels.

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

Closure Report for Corrective Action Unit 574: Neptune, Nevada National Security Site, Nevada

Corrective Action Unit (CAU) 574 is identified in the Federal Facility Agreement and Consent Order (FFACO) as 'Neptune' and consists of the following two Corrective Action Sites (CASs), located in Area 12 of the Nevada National Security Site: (1) CAS 12-23-10, U12c.03 Crater (Neptune); and (2) CAS 12-45-01, U12e.05 Crater (Blanca). This Closure Report presents information supporting closure of CAU 574 according to the FFACO (FFACO, 1996 [as amended March 2010]) and the Streamlined Approach for Environmental Restoration Plan for CAU 574 (U.S. Department of Energy, National Nuclear Security Administration Nevada Site Office [NNSA/NSO], 2011). The following activities were performed to support closure of CAU 574: (1) In situ external dose rate measurements were collected using thermoluminescent dosimeters at CAS 12-45-01, U12e.05 Crater (Blanca). (2) Total effective dose rates were determined at both sites by summing the internal and external dose rate components. (3) A use restriction (UR) was implemented at CAS 12-23-10, U12c.03 Crater (Neptune). Areas that exceed the final action level (FAL) of 25 millirems per year (mrem/yr) based on the Occasional Use Area exposure scenario are within the existing use restricted area for CAU 551. The 25-mrem/yr FAL is not exceeded outside the existing CAU 551 UR for any of the exposure scenarios (Industrial Area, Remote Work Area, and Occasional Use Area). Therefore, the existing UR for CAU 551 is sufficient to bound contamination that exceeds the FAL. (4) An administrative UR was implemented at CAS 12-45-01, U12e.05 Crater (Blanca) as a best management practice (BMP). The 25-mrem/yr FAL was not exceeded for the Remote Work Area or Occasional Use Area exposure scenarios; therefore, a UR is not required. However, because the 25-mrem/yr FAL was exceeded for the Industrial Area exposure scenario, an administrative UR was established as a BMP. UR documentation is included as Appendix B. The UR at CAS 12-23-10, U12c.03 Crater (Neptune), is within the existing UR for CAU 551. Additional postings were not installed, and annual post-closure inspections will be performed in conjunction with the inspections performed for CAU 551. At CAS 12-45-01, U12e.05 Crater (Blanca), the administrative UR does not require postings or inspections. NNSA/NSO requests the following: (1) A Notice of Completion from the Nevada Division of Environmental Protection to NNSA/NSO for closure of CAU 574; and (2) The transfer of CAU 574 from Appendix III to Appendix IV, Closed Corrective Action Units, of the FFACO

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

High Sodium Simulant Testing To Support SB8 Sludge Preparation

Scoping studies were completed for high sodium simulant SRAT/SME cycles to determine any impact to CPC processing. Two SRAT/SME cycles were performed with simulant having sodium supernate concentration of 1.9M at 130% and 100% of the Koopman Minimum Acid requirement. Both of these failed to meet DWPF processing objectives related to nitrite destruction and hydrogen generation. Another set of SRAT/SME cycles were performed with simulant having a sodium supernate concentration of 1.6M at 130%, 125%, 110%, and 100% of the Koopman Minimum Acid requirement. Only the run at 110% met DWPF processing objectives. Neither simulant had a stoichiometric factor window of 30% between nitrite destruction and excessive hydrogen generation. Based on the 2M-110 results it was anticipated that the 2.5M stoichiometric window for processing would likely be smaller than from 110-130%, since it appeared that it would be necessary to increase the KMA factor by at least 10% above the minimum calculated requirement to achieve nitrite destruction due to the high oxalate content. The 2.5M-130 run exceeded the DWPF hydrogen limits in both the SRAT and SME cycle. Therefore, testing of this wash endpoint was halted. This wash endpoint with this minimum acid requirement and mercury-noble metal concentration profile appears to be something DWPF should not process due to an overly narrow window of stoichiometry. The 2M case was potentially processable in DWPF, but modifications would likely be needed in DWPF such as occasionally accepting SRAT batches with undestroyed nitrite for further acid addition and reprocessing, running near the bottom of the as yet ill-defined window of allowable stoichiometric factors, potentially extending the SRAT cycle to burn off unreacted formic acid before transferring to the SME cycle, and eliminating formic acid additions in the frit slurry.

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

Results of Hg speciation testing on tanks 30, 32, and 37 depth samples

The Savannah River National Laboratory (SRNL) was tasked with preparing and shipping samples for Hg speciation by Eurofins Frontier Global Sciences, Inc. in Seattle, WA on behalf of the Savannah River Remediation (SRR) Mercury Task Team. The twelfth shipment of samples was designated to include 3H evaporator system Tank 30, 32, and 37 depth samples. The Tank 30 depth sample (HTF-30-15-70) was taken at 190 inches from the tank bottom and the Tank 32 depth sample (HTF-32-15-68) was taken at 89 inches from the tank bottom and both were shipped to SRNL on June 29, 2015 in an 80 mL stainless steel dip bottles. The Tank 37 surface sample (HTF-37-15-94) was taken around 253.4 inches from the tank bottom and shipped to SRNL on July 21, 2015 in an 80 mL stainless steel dip bottle. All samples were placed in the SRNL Shielded Cells and left unopened until intermediate dilutions were made on July 24, 2015 using 1.00 mL of sample diluted to 100.00 mL with deionized H 2 O. A 30 mL Teflon® bottle was rinsed twice with the diluted tank sample and then filled leaving as little headspace as possible. It was immediately removed from the Shielded Cells and transferred to refrigerated storage where it remained at 4 °C until final dilutions were made on October 20. A second portion of the cells diluted tank sample was poured into a shielded polyethylene bottle and transferred to Analytical Development for radiochemical analysis data needed for Hazardous Material Transportation calculations.

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

Preliminary Results from Electric Arc Furnace Off-Gas Enthalpy Modeling

This article describes electric arc furnace (EAF) off-gas enthalpy models developed at Oak Ridge National Laboratory (ORNL) to calculate overall heat availability (sensible and chemical enthalpy) and recoverable heat values (steam or power generation potential) for existing EAF operations and to test ORNL s new EAF waste heat recovery (WHR) concepts. ORNL s new EAF WHR concepts are: Regenerative Drop-out Box System and Fluidized Bed System. The two EAF off-gas enthalpy models described in this paper are: 1.Overall Waste Heat Recovery Model that calculates total heat availability in off-gases of existing EAF operations 2.Regenerative Drop-out Box System Model in which hot EAF off-gases alternately pass through one of two refractory heat sinks that store heat and then transfer it to another gaseous medium These models calculate the sensible and chemical enthalpy of EAF off-gases based on the off-gas chemical composition, temperature, and mass flow rate during tap to tap time, and variations in those parameters in terms of actual values over time. The models provide heat transfer analysis for the aforementioned concepts to confirm the overall system and major component sizing (preliminary) to assess the practicality of the systems. Real-time EAF off-gas composition (e.g., CO, CO2, H2, and H2O), volume flow, and temperature data from one EAF operation was used to test the validity and accuracy of the modeling work. The EAF off-gas data was used to calculate the sensible and chemical enthalpy of the EAF off-gases to generate steam and power. The article provides detailed results from the modeling work that are important to the success of ORNL s EAF WHR project. The EAF WHR project aims to develop and test new concepts and materials that allow cost-effective recovery of sensible and chemical heat from high-temperature gases discharged from EAFs.

Electric Arc Furnace↗
DOE OSTI2019

Aluminum Clad SNF: Technical Considerations and Challenges for Extended (>50 years) Dry Storage - 1900y

The U.S. Department of Energy (DOE) manages a significant inventory (>13 MTHM) of aluminum clad spent nuclear fuel (ASNF) using a combination of wet and dry storage at three primary sites; Hanford, Idaho National Laboratory (INL), and Savannah River Site (SRS). The majority of this ASNF is generated from U.S. and international research test reactors (RTR), examples include Advanced Test Reactor (ATR) at INL and High-Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). The Obama Administration suspension of the Yucca Mountain repository, issuance of the January 2013 DOE 'Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste,' delays in implementing the 'regionalization' SNF approach recommended in the 1995 'Department of Energy Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs Final Environmental Impact Statement,' and limited ASNF processing capability at SRS H-Canyon means that interim storage has become the de-facto strategy for managing ASNF. The 1995 Idaho Settlement Agreement requires the DOE to transfer all SNF stored at INL to interim dry storage prior to December 31, 2023 and wet storage capacity at the SRS is nearing capacity. This situation was recognized by the DOE Spent Nuclear Fuel Working Group (SNFWG) and at its June 2016 meeting formed a ASNF sub-working group comprised subject matter experts to evaluate this area and provide a report of findings. The ASNF sub-working group met many times over several month period to review the engineering and technical basis of existing and planned DOE ASNF dry storage facilities and additional studies on the behavior of ASNF under different conditions, including during reactor service. The ASNF sub-working group focused its review on RTR SNF. Based on these reviews, the ASNF sub-working group identified five knowledge gaps and technical data needs that should be addressed to ensure viability and safety of extended, greater than 50 years, sealed or vented interim dry storage of ASNF. These five areas include; behavior/chemistry of aluminum oxyhydroxide layer in dry storage, radiolytic gas generation properties of aluminum oxyhydroxide layers, effect of episodic breathing and radiolytic generation of corrosive gases during storage, performance of ASNF in existing dry storage systems, and effects of high temperature (>1000 C) drying on chemistry and behavior of aluminum oxyhydroxide layers. The ASNF sub-working group also recommended development of an action plan to identify technical and engineering activities needed to address the knowledge gaps and technical data needs. Results of the ASNF sub-working group are documented in 2017 report 'Aluminum Clad Spent Nuclear Fuel: Technical Considerations and Challenges for Extended (>50 years) Dry Storage,' DOE-ID/RPT-1575. The INL assembled a team comprised of twenty INL and Savannah River National Laboratory SNF management and subject matter experts to develop this action plan. Over the course of 2 months the team worked to develop the action plan. The plan identified and recommended six technical and engineering tasks needed to address the previously identified knowledge gaps and technical data needs. DOE Environmental Management Office of Technology Development is providing funding to the INL to execute these tasks. This paper will present and discuss the results of the ASNF sub-working group, tasks identified in the action plan, and experimental activities underway at INL and SRNL to address the action plan. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2014

Single Stage Contactor Testing Of The Next Generation Solvent Blend

The Modular Caustic Side Solvent Extraction (CSSX) Unit (MCU) facility at the Savannah River Site (SRS) is actively pursuing the transition from the current BOBCalixC6 based solvent to the Next Generation Solvent (NGS)-MCU solvent to increase the cesium decontamination factor. To support this integration of NGS into the MCU facility the Savannah River National Laboratory (SRNL) performed testing of a blend of the NGS (MaxCalix based solvent) with the current solvent (BOBCalixC6 based solvent) for the removal of cesium (Cs) from the liquid salt waste stream. This testing utilized a blend of BOBCalixC6 based solvent and the NGS with the new extractant, MaxCalix, as well as a new suppressor, tris(3,7dimethyloctyl) guanidine. Single stage tests were conducted using the full size V-05 and V-10 liquid-to-liquid centrifugal contactors installed at SRNL. These tests were designed to determine the mass transfer and hydraulic characteristics with the NGS solvent blended with the projected heel of the BOBCalixC6 based solvent that will exist in MCU at time of transition. The test program evaluated the amount of organic carryover and the droplet size of the organic carryover phases using several analytical methods. The results indicate that hydraulically, the NGS solvent performed hydraulically similar to the current solvent which was expected. For the organic carryover 93% of the solvent is predicted to be recovered from the stripping operation and 96% from the extraction operation. As for the mass transfer, the NGS solvent significantly improved the cesium DF by at least an order of magnitude when extrapolating the One-stage results to actual Seven-stage extraction operation with a stage efficiency of 95%.

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

Emergency response and solutions for damaged fuel assembly handling events - Lessons learned from a 2013 fuel positioning issue

In 2013, during the refueling of a French 1300 MWe reactor, a fuel assembly (FX3599) was damaged by a fuel assembly positioning tool in the reactor vessel. The upper part of a corner rod of the fuel assembly was twisted and the top end grid was torn off. Because of the protruding fuel rod, the assembly could not be easily transferred for repair to the fuel storage pool via the transfer tube. This unexpected and rarely encountered situation was assessed jointly by EDF and AREVA NP. The chosen option was to directly perform the field service inside the reactor building in cells generally used for intermediate storage during refueling. This operation, performed by EDF and AREVA, involved removing the twisted rod and positioning it in a special-purpose canister by means of a gripper-type tool for removal (KIPCI gripper) and a gripper for holding the rod in place. The special-purpose canister was designed, ordered and supplied by a sub-supplier within one week. The operation went through 3 phases. First a preparation step in the fuel building: The stripper was disassembled and technical checks carried out to ensure correct operation. The gripper tightening force was checked in real-life conditions with the gripper suspended from auxiliary crane. The fuel rod removal was simulated several times by trained operators. Second, the equipment was installed in the reactor building. And third, the operation took place in the reactor cavity: the special-purpose canister was opened, the radioactive waste canister placed in the special-purpose canister, the twisted fuel rod was then grasped by the KIPCI gripper and removed from the fuel assembly. The twisted fuel rod was positioned above the canister and dropped into it. Once the fuel rod had been removed, the canister and fuel assembly FX3599 were transferred to the fuel building and refueling resumed. The canister containing the fuel rod is currently in interim storage in the reactor pool until appropriate action is taken for it to be shipped and eventually reprocessed. EDF has performed improvement in its handling procedures and in its handling tools to avoid a similar new event.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗