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The laboratory creates a variety of wastes, each with its own health and safety concerns, regulatory requirements, and disposal pathways. However, the basic process, from preplanning to waste disposal, is always essentially the same.
This Fiscal Year (FY) 2020 annual summary report documents the continued adequacy of the performance assessment (PA), the composite analysis (CA) and associated operating disposal-authorization statement (ODAS) technical basis documents for the Remote Handled (RH) Low-Level Waste (LLW) Disposal Facility at Idaho National Laboratory. Annual review of the adequacy of the PA and CA for the RHLLW Disposal Facility ensures that conclusions of the analyses remain valid in accordance with requirements of Department of Energy (DOE) Order 435.1 Change 1, “Radioactive Waste Management.” In FY 2020, no significant operational changes or other activities occurred that would cause deviation from the assumptions in the PA and CA pertaining to disposal geometry, verification of waste characteristics, tracking disposal inventories against total limits, facility closure design, or institutional controls. Fifteen total waste-canister shipments were received at the RHLLW Disposal Facility, and 15 total waste canisters were emplaced.
Environmental monitoring data are collected at and around the Area 3 and Area 5 Radioactive Waste Management Sites (RWMSs) within the Nevada National Security Site (NNSS). This report summarizes the 2021 environmental data to provide an overall evaluation of RWMS performance and to support environmental compliance and performance assessment (PA) activities.
The plenary panel focused on government and industry world leaders speaking on the pressing issues facing radioactive waste management in 2020 and beyond. Panelists included an OECD/ NEA representative for the international community, a Huntington Ingalls Industries representative for the commercial sector, and the US DOE Senior Advisor to the Under Secretary of Science representing DOE Environmental Management. Panelists with presentations: WM2020 Speech (Michael Lempke); Optimization: A Better Path toward Managing Nuclear Decommissioning and Legacies (William Magwood); Update On the DOE Environmental Management Program (William White)
The US Department of Energy Office of Nuclear Energy (DOE-NE) Office of Spent Fuel and High-Level Waste Disposition is examining a set of system options and conducting supporting analyses to inform the development of an integrated waste management system, which may include one or more federal staging facilities (FSFs) for used nuclear fuel (UNF ) sited using a collaborative siting process. This paper focuses on the ongoing activities in two systems engineering and analysis work areas: (1) data and tools development, validation, and maintenance and (2) systems engineering execution. Within the first work area, the STANDARDS 5.0 UNF data and analysis tool, formerly known as UNF-ST&DARDS, is being developed as a foundational resource to assist in the management of UNF data. It has the key capability to model UNF throughout the entire back end of the fuel cycle. STANDARDS also includes several compatible analysis tools for the time-dependent characterization of UNF and related systems by interfacing with the SCALE code system for nuclear analysis and COBRA-SFS for thermal analysis. Also, within the data and tools area is the Next Generation System Analysis Model (NGSAM), which is an agent-based simulation software tool expressly designed to be capable of modeling the waste management system, including the transportation of UNF to and from a FSF. NGSAM has been developed to enable informed decision-making by providing the capability to analyze various potential system options for the management of UNF and high-level radioactive waste. Finally, in the systems engineering execution area, the team has begun to apply a disciplined systems engineering approach at the system level along with supporting analysis to guide the development of the FSF project requirements (including associated transportation infrastructure). Systems engineering principles and practices and their adaptation/application to design and development activities will ensure that the waste management system is effectively implemented as work proceeds. Other activities include investigating the implications of changes in various assumptions and parameters related to waste management systems, such as UNF acceptance rates, receipt logic, facility capacities and capabilities, use of standardized canisters, and different assumed facility operation start dates. Keywords: federal staging facility (FSF), used nuclear fuel (UNF), integrated waste management (IWM) system, Next Generation System Analysis Model (NGSAM), STANDARDS, systems engineering
Based on the higher interest in Advanced Reactor (AR) deployment (e.g., ARDP ) for potential new fuel cycles, the Spent Fuel & Waste Science and Technology (SFWST) Program has begun to evaluate the possible implications of long-term management and final disposition of potential Advance Reactor spent nuclear fuels (SNF) that would be generated in potential advanced reactors. Safely managing and dispositioning the potential future AR SNF, and any other associated radioactive wastes, is the primary focus of this initial preliminary assessment of those. This paper summarizes three primary tasks the Spent Fuel & Waste Science and Technology (SFWST) Program is executing (or collaborating on) related to the back end of the nuclear fuel cycle (BENFC) for potential future advanced reactors: 1. Advanced Reactors Spent Fuel and Waste Streams: Strategies for the BENFC This set of activities define a high-level strategy for how to systematically approach, identify, and close research and development (R&D) gaps/potential issues associated with long-term management and final disposition of AR SNF and other possible AR waste streams. This task involves summarizing advanced reactor concepts, their likely SNF and other waste forms, and identifying previous experience with similar materials, for example from DOE-managed SNF, with closely related characteristics to the potential future AR SNF. Technical R&D gaps between the breadth of detailed understanding for safe storage, transportation and disposal of the existing light water reactor SNF fuel cycle (e.g., see NASEM, 2022) and potential future fuel cycles based on advanced reactors would then be identified. 2. Characterization and Packaging Options of Advanced Reactor SNF These activities evaluate characteristics and packaging options for advanced reactor spent fuel forms. The fuel forms are categorized into three types: (1) tri-structural isotropic (TRISO), (2) metallic, and (3) fuel salt. Emphasis is given to TRISO and metallic SNF and additional waste streams from such AR as driven by the near-term anticipated operation of the Xe-100 and the Natrium reactors as advanced-reactor demonstrations1. Preliminary information for the spent-fuel salt discharged from molten-salt reactors (MSRs) is examined to provide a baseline for future efforts. All calculations and assumptions in this work are based on publicly available information. The following characteristics are calculated or estimated for use in the preliminary assessments: SNF volume and mass, radiation/activity levels through time, thermal conditions through time, potential radionuclide source terms, chemical interactions and evolutions, disposal inventories, and waste-form lifetime. Using those characteristics, calculations to determine the applicability of existing canister designs were performed. These evaluations included geometric (e.g., dimension, volume) and mass/weight considerations, known operational approaches and loading procedures, physical and chemical considerations/conditions for storage environments, as-loaded radiation, thermal, and criticality analyses to identify constraints for storage, transportation, and disposal. 3. Back-End Management of Advanced Reactors (BEMAR) The DOE NE-8 organization has defined an Integrated Project Team to evaluate the Back End Management of Advanced Reactors (BEMAR) (includes DOE staff from a range of organizations (e.g., NE-81, NE-82, OCED) and national laboratory technical staff within the DOE NE-81 and NE-82 programs). This BEMAR group works directly with advance reactors developers to assess for the DOE the technical feasibility of storage, transportation, and disposal of AR SNF based on the characteristics provided by the developers to DOE (much of which is proprietary). The BEMAR is also tasked to develop rough-order-of-magnitude cost estimates to compare the waste management system for individual advanced reactors to existing light-water reactor management practices. To accomplish this, the BEMAR group is implementing a Systems Engineering approach.
This panel focused on DOE and Non-US Technology Development Projects discussing how they approach tech development to maximize return on investment (ROI). Successful examples were discussed. After the individual panelist's presentations, the three panelists were joined by Kurt Gerdes in discussing the genesis of this panel and why it is important. Panelists with presentations: Ensuring a Strong Return on Investment for DOE EM Technology Development (Anthony Banford) DOE Technology Development Projects that Provided a Strong Return on Investment (ROI (Mark Lesinski) Australian Nuclear Science and Technology Organisation (Gerry Triani) Decommissioning Situation of Nuclear Power Plant in Japan (Teruyuki Hirai)
The U.S. Department of Energy (DOE) operates a low-level radioactive waste (LLW) disposal site at Material Disposal Area G, in Los Alamos, New Mexico, USA. Area G has been the primary LLW disposal site for Los Alamos National Laboratory (LANL) since the 1960's. In addition to LLW, Area G is host to a variety of other wastes, the disposition of which must be determined before closure of the site. A probabilistic Radiological Risk Assessment (RRA) for Area G is used in order to support decision making regarding some wastes that are not addressed in the extant Area G Performance Assessment (PA) and Composite Analysis (CA). Between 1979 and 1987, 33 special shafts were augered into the Bandelier Tuff at Area G. This volcanic tuff is present across Pajarito Plateau on the eastern slopes of the Jemez Mountains, and varies widely in its consistency, from weakly indurated non-welded layers to welded layers that uphold the mesa cliffs of the Plateau. These mesas are home to LANL, Area G, and the townsites of Los Alamos and White Rock, with residences about 1400 m from Area G. The 33 Shafts were lined with steel casing, and contain remote-handled (RH) transuranic wastes (TRU) resulting from experiments and analysis performed in special glove boxes at the Chemistry and Metallurgy Research (CMR) facility at LANL. Some of these wastes originated as used nuclear fuel. The purpose of the Area G RRA is to evaluate the potential future risk to humans and the environment from the RH TRU in the 33 Shafts in the context of the risk associated with the surrounding wastes at Area G. The analysis is responsive to expectations outlined in DOE Order 458.1, Radiation Protection of the Public and the Environment, and is informed by the Manual and Guidance accompanying DOE O 435.1, Radioactive Waste Management. Because the waste meets the definition of TRU, the regulatory context necessarily takes into consideration the regulation governing the disposal of TRU from the U.S. Environmental Protection Agency (EPA): 40 CFR 191, Environmental Radiation Protection Standards for Management and Disposal of Spent Nuclear Fuel, High-Level and Transuranic Radioactive Wastes. Given the broader regulatory context for the RRA, the analysis is subject to different assumptions from those made in the existing DOE O 435.1 PA and CA, such as allowing for future occupation of the site. The analysis begins with a comprehensive evaluation of features, events, processes, and exposure scenarios (FEPS) for Area G and the wastes it contains. These FEPSs are screened to eliminate from further consideration those of extremely low probability and/or consequence, and a conceptual site model (CSM) is subsequently developed. The scope and structure of the Area G RRA Model is informed by this CSM, and the Area G RRA Model is developed using the GoldSim systems analysis modeling platform. This paper presents the initial version of a defensible, transparent, and reasonably realistic model, which is based on the state of knowledge of the wastes, the site, and the FEPSs that govern contaminant transport from wastes into the environment and subsequent exposures to humans and other biota. Probabilistic model input distributions represent uncertainties inherent in the real and modeled systems. The results of the Area G RRA Model inform decisions regarding the disposition of the RH TRU in the 33 Shafts. (authors)
This report describes the FY24 status of international collaboration regarding geologic disposal research in the Office of Spent Fuel and Waste Science and Technology (SFWST) in the Department of Energy’s (DOE) Office of Nuclear Energy (NE). The mission of SFWST is to provide confidence in the safe long-term management of the nation’s spent nuclear fuel and high-level radioactive waste by reducing uncertainty and advancing technology for extended storage, transportation, and geologic disposal.
This panel focused on young professionals and covered views on radioactive waste management from young persons' perspective from all around the world. An informal panel enabled the future leaders of our industry to share ideas and experiences, and of course, develop contacts within their peer group on a truly international level. The panel was led by young professionals who actively encouraged participation from all those attending. Panelists with presentations: Young Professionals in Nuclear Science and Engineering An International Journey (Mario Gomez Fernandez); Young Professionals in Nuclear Science and Engineering (Mellissa Komninakis); Young Professionals in Nuclear Science and Engineering - an International Perspective (Hannah Paterson, Carwyn Chamberlain)
This report demonstrates the capabilities of the net-zero microgrid (NZM) Xendee platform for modeling an SR module with electricity, heat extraction and thermal storage in microgrids configurations. The model effectively captures the most important technical and economic considerations for SR technology specific analysis: cost and operational characteristics of SR technology and financial costs and incentives. The model can analyze multiple scenarios to establish metrics for cost-competitive and zero-carbon microgrids connected to the grid or completely isolated. The model is fully integrated within the Xendee platform for modeling and analysis of clean energy microgrids with storage and generation from renewable energy sources. The model captures the capabilities, constraints, and nuances of SR by incorporating parameters related to plant economics, design efficiency and performance, plant operation and component and fuel lifespan. The cost and operational parameters modeled in the SR module are specific to the technology selected for integration in the microgrid. Cost parameters recognize advanced nuclear technology for modular production and installation based on economies of scale from factory manufacture and related commissioning, and cost reduction through technology maturation—first-of-a-kind (FOAK) and nth-of-a-Kind (NOAK). The cost parameters include installation, operations and maintenance (O&M), fuel refueling cycle, and reactor life. Installation cost reflects economies of scale due to unit sizing at scale and colocation. O&M economies of scale for both fixed- and variable-cost fuel life-cycle costs are incurred at every refueling interval, with separate front- and back-end fuel costs, as well as waste-handling and disposition costs. This report investigates key characteristics of different SR technologies suitable for microgrid applications, including design principles, sizing, coolant properties, temperature ratings, fuel structures, and life-cycle considerations. This also includes fuel technologies applicable to these SR systems, alongside strategies for nuclear-waste and spent-fuel management and approaches to address safety, security, and proliferation challenges. Four primary groups of SR technologies are examined: water-cooled, liquid-metal-cooled, high-temperature gas-cooled, and molten-salt-cooled systems. In this report, an initial guideline for technology selection is established, aligning the characteristics of the technologies with the requirements of microgrids. The selection of technology in a microgrid is influenced by various factors, including financial capacity, location and accessibility, demand type and characteristics, reliability and resilience requirements, area constraints, and the lifespan of the microgrid. The types of electrical and non-electrical applications within the microgrid also play a significant role in technology selection. The characteristics of SRs, such as their smaller size, modularity, transportability, long refueling interval, improved safety features, ability to operate in autonomous or semi-autonomous mode, and provision of high-grade heat, are particularly appealing for microgrids. Furthermore, a list of considerations for implementing SRs in microgrids is outlined. The SR model is created to be continuously improved with the acquisition of actual data on investment and operational costs, experience with supply chains, production at scale, and field deployments. In the near term, performance data on applications in microgrids will become available from lessons learned from laboratory tests, such as those planned for the Microreactor Applications Research Validation and Evaluation Project (MARVEL), led by Idaho National Laboratory (INL). The SR model incorporates scenario data and known SR design specifications, enabling technoeconomic analysis for SR deployment in microgrids. It specifically considers the distinctive attributes of SRs as generators in technoeconomic studies. SRs can be modeled and analyzed with generation from renewable-energy sources, energy storage, and flexible loads over a range of functionality and applications. This offers a comprehensive tool for feasibility studies, scenario development, and sensitivity analysis for “what-if” consideration of any range of assumptions about SRs in microgrids and other aggregations of distributed-energy resources, including virtual power plants.
Technetium is a problematic radioisotope for used nuclear fuel (UNF) and subsequent waste management owing to its high environmental mobility and coextraction in reprocessing technologies as the pertechnetate anion (TcO 4 - ). Consequently, several strategies are under development to control the transport of this radioisotope. A proposed approach is to use diaminoguanidine (DAG) for TcO 4 - and transuranic ion redox control. Although the initial DAG molecule is ultimately consumed in the redox process, its susceptibility to radiolysis is currently unknown under envisioned UNF reprocessing conditions, which is a critical knowledge gap for evaluating its overall suitability for this role. To this end, we report the impacts of steady-state gamma irradiation on the rate of DAG radiolysis in water, aqueous 2.0 M nitric acid (HNO 3 ), and in a biphasic solvent system composed of aqueous 2.0 M HNO 3 in contact with 1.5 M N,N-di-(2-ethylhexyl)isobutyramide (DEHiBA) dissolved in n-dodecane. In addition, we report chemical kinetics for the reaction of DAG with key transients arising from electron pulse radiolysis, specifically the hydrated electron (e aq - ), hydrogen atom (H ˙ ), and hydroxyl ( ˙ OH) and nitrate (NO 3 ˙ ) radicals. The DAG molecule exhibited significant reactivity with the ˙ OH and NO 3 ˙ radicals, indicating that oxidation would be the predominant degradation pathway in radiation environments. This is consistent with its role as a reducing agent. Steady-state gamma irradiations demonstrated that DAG is readily degraded within a few hundred kilogray, the rate of which was found to increase upon going from water to HNO 3 containing solutions and solvents systems. This was attributed to a thermal reaction between DAG and the predominant HNO 3 radiolysis product, nitrous acid (HNO 2 ), k(DAG + HNO 2 ) = 5480 ± 85 M -1 s -1 . Although no evidence was found for the radiolysis of DAG altering the radiation chemistry of the contacted DEHiBA/n-dodecane phase in the investigated biphasic system, the utility of DAG as a redox control reagent will likely be limited by significant competition with its degradation by HNO 2 .
Abstract The “U‐phase,” a sodium‐containing (alumino‐ferrite‐monosubstituent) AFm phase, has been observed to form in sodium‐enriched highly alkaline cementitious systems, for example, of relevance to nuclear waste, and saline industrial brine management. But, minimal information is available of the U‐phase's (e.g., solubility or thermodynamic properties) due to its limited stability and its tendency to transform into ettringite or monosulfoaluminate. Herein, the U‐phase was systematically synthesized at four temperatures (5, 15, 25, and 50°C) and fully characterized in terms of its thermochemical properties. The average composition of the synthesized U‐phase (4CaO·Al 2 O 3 ·1.85SO 3 ·0.85Na 2 O·12H 2 O) deviates slightly from typical disclosures in the literature. The solubility product of the U‐phase formation was measured from conditions of oversaturation. The measured thermodynamic data accurately predicted experimental observations of U‐phase formation in cementitious environments. In general, it was noted that the U‐phase forms and persists (i.e., remains stable) at pH > 13.7 and [Na + ] concentrations superior to 1 mol/L.
The effects of ionizing radiation are ubiquitous throughout all aspects of a nuclear fuel cycle. The complexity and intensity of these effects are greatest during reactor operations and in the management of used nuclear fuel and waste. Radiation-induced processes typically promote the chemical transformation of molecules and materials with the formation of potentially detrimental degradation products and corresponding changes in physical and chemical properties, which ultimately impact the effectiveness and longevity of nuclear technologies. Consequently, a molecular-level understanding of radiation effects over multiple time, distance, and material domains is essential for the innovation and deployment of next generation nuclear technologies. Attaining this knowledge necessitates a firm grasp of radiation-induced reaction kinetics, for which pulsed electron radiolysis is the methodology of choice. Presented here are several recent studies from our group that demonstrate the critical role of pulsed electron radiolysis techniques in the advancement of our understanding of radiation-induced chemistry under advanced nuclear fuel cycle conditions. Research topics include late actinide redox chemistry, radiation robustness of used nuclear fuel reprocessing complexants, and the behavior of metal cations in high temperature molten salt media.
Effective management and disposal of legacy nuclear waste are essential for ensuring safe work environments and minimizing environmental impacts. Monitoring airborne actinide contamination is particularly critical due to the high internal dose potential of alpha-emitting radionuclides. Traditional continuous air monitoring systems (CAMs) used in the industry are limited in the volume of air they can sample, potentially leading to inaccurate radiation detection over large areas. For example, in 2018, elevated levels of airborne Plutonium-239 were detected beyond the controlled areas of the Hanford Plutonium Finishing Plant, highlighting the potential risks to both plant workers and nearby residents. To address these challenges, high-efficiency particulate absorbing (HEPA) air purifiers can enhance air flow by up to 1.5 orders of magnitude, thereby increasing monitoring efficiency and providing a cost-effective solution for large-area surveillance. To quantify the activities of alpha-emitting radionuclides on HEPA filters, the Savannah River National Laboratory is developing an advanced alpha-imaging detection system. This system includes scintillating materials combined with a digital scientific camera. A significant concern in operating such a large-area airborne monitoring system is the handling of HEPA filters, which may be contaminated with radioactive particles. To mitigate these hazards, it is crucial to ensure that any contamination is securely fixed onto the filters. Efforts have been made to optimize the sensitivity of scintillator-epoxy composites and apply them to HEPA filters. These materials were characterized using fluoroscence spectroscopy. These techniques confirmed the purity of the raw materials, the dispersion of scintillators in the epoxy matrix, and the stability of their optical and structural properties post-modification. The optimal scintillator-epoxy composite was selected for use on alpha-spiked HEPA filters to evaluate the efficiency of the sprayer. HEPA filters, embedded with alpha particles collected by an air purifier deployed in an airborne radiation area, have been tested to assess detection efficiency. Future work will focus on employing multiple imaging sensors simultaneously to enhance sensitivity across different regions of the HEPA filter.
Not Available
This is the licensing presentation for the Waste Management Symposia 2023. It covers licensing activities, LMP, and manufacturing licenses.
Corrective Action Unit (CAU) 577, “Area 5 Chromium Containing Waste Disposal Cells,” includes five low-level waste cells at the Nevada National Security Site Area 5 Radioactive Waste Management Site where buried waste received from Nuclear Fuel Services, Inc., was subsequently determined to contain chromium that exceeded the toxicity characteristic leaching procedure regulatory limit, which would require the waste to carry hazardous waste code D007. CAU 577 was created to satisfy the requirements of the Settlement Agreement (SA) executed between the Nevada Division of Environmental Protection (NDEP) and the U.S. Department of Energy, National Nuclear Security Administration Nevada Field Office (NNSA/NFO) on April 25, 2019 (NDEP 2019). The SA required that the chromium-containing waste received from NFS would be addressed following the closure process laid out in the Federal Facility Agreement and Consent Order (FFACO). The FFACO process ensures proper closure of the chromium-containing waste and documentation of that closure through FFACO-type documents. The following three Corrective Action Sites (CASs) were closed, and their closure was documented in the Closure Report for Corrective Action Unit 577: Area 5 Chromium Containing Waste Disposal Cells, Nevada National Security Site, Nevada, DOE/EMNV--0030, dated September 2021 (U.S. Department of Energy [DOE] Environmental Management [EM] Nevada Program 2021a): • CAS 05-21-02, Waste Disposal Cell 12 • CAS 05-21-03, Waste Disposal Cell 15 • CAS 05-21-04, Waste Disposal Cell 17 Closure of the following CAS was previously documented in the Addendum to the Closure Report for Corrective Action Unit 577: Area 5 Chromium Containing Waste Disposal Cells, Nevada National Security Site, Nevada, DOE/EMNV--0030-ADD, dated September 2022 (DOE EM Nevada Program 2022): • CAS 05-21-05, Waste Disposal Cell 20 This second addendum to the Closure Report documents the closure activities that have occurred for CAS 05-21-06, Waste Disposal Cell 21. This is the last CAS in CAU 577. The final waste shipment was placed in the waste disposal cell on November 21, 2022. Following this, closure activities began on November 28, 2022, and were conducted according to the Corrective Action Decision Document/Corrective Action Plan (CADD/CAP) for CAU 577 (DOE EM Nevada Program 2021b). The following closure activities were performed: • Constructing an engineered evapotranspiration cover • Installing two subsidence monuments and vadose zone monitoring equipment • Seeding the cover with a mixture of native plant species • Installing four concrete monuments on the corners of the cover and placing two use restriction (UR) warning signs on each monument These activities fulfill applicable federal and state regulations for closure of CAS 05-21-06 and minimize potential future exposure pathways to buried waste. Completed closure activities are also consistent with closure of the nine historical Resource Conservation and Recovery Act (RCRA) units included in Section 10.2.2 of the RCRA Permit that governs hazardous waste management activities at the Nevada National Security Site (Permit NEV HW0101) (NDEP 2023). UR documentation for this CAS is included in Appendix B of this report. The post-closure plan is presented in detail in the CADD/CAP for CAU 577 (DOE EM Nevada Program 2021b), and the requirements are summarized in Section 5.2 of this document. In accordance with paragraph 5D of the SA, a request to incorporate the requirements for post-closure monitoring of CAU 577 was included with the permit application for RCRA Permit NEV HW0101 that was submitted in January 2022 (NNSA/NFO 2022). The request included the post-closure requirements for the three CASs that had been closed at the time of submittal of the application as well as requirements that would be implemented upon future approval of closure of the remaining two CASs. All CAU 577 post-closure monitoring requirements have been captured in the April 4, 2023, Revision 7 of the RCRA Permit (NDEP 2023). As the RCRA Permit NEV HW0101 has been revised since the submittal of the original CAU 577 Closure Report (DOE EM Nevada Program 2021a) and Addendum 1 (DOE EM Nevada Program 2022), the post-closure requirements in this Addendum 2 report do not align with the previous documents. Specific changes resulting from the issuance of Revision 7 of the RCRA Permit are discussed in Section 5.2 of this report. The requirements in this report are consistent with the current permit (NDEP 2023) and supersede all requirements listed in the CAU 577 Closure Report and Addendum 1. All CAU 577 post-closure requirements should be conducted in accordance with the version of the RCRA Permit that is current at the time of the activities being performed. The DOE EM Nevada Program is requesting a Notice of Completion from NDEP for closure of CAU 577. Although CAU 577 is not a legacy site, the FFACO process is being followed to ensure proper closure of the chromium-containing waste. Therefore, transfer of CAU 577 from Appendix III of the FFACO to Appendix IV, Closed Corrective Action Units, is requested, as all closure activities have been completed.