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Generic FEPs Catalogue and Salt Knowledge Archive

This report describes the development of a comprehensive catalogue of generic features, events, and processes (FEPs) that are potentially important for the post-closure performance of a repository for high-level radioactive waste (HLW) and spent nuclear fuel (SNF) in salt (halite) host rock. The FEPs and other supporting information have been entered into a “SaltFEP” Database. The generic salt repository FEPs include consideration of relevant FEPs from a number of U.S., Dutch, German, and international FEP lists and should be a suitable starting point for any repository program in salt host rock. The salt FEP catalogue and database employ a FEP classification matrix approach that is based on the concept that a FEP is typically a process or event acting upon or within a feature. The FEP matrix provides a two-dimensional structure consisting of a Features/Components axis that defines the “rows” and a Processes/Events axis that defines the “columns” of the matrix. The design of the FEP classification matrix is consistent with repository performance assessment – the Features/Components axis is organized vertically to generally correspond to the direction of potential radionuclide migration (from the waste to the biosphere) and the Processes/Events axis is designed to represent the common two-way couplings between thermal processes and other processes (such as thermal-mechanical or thermal-hydrologic processes). Related FEPs can be easily identified – related FEPs will typically be grouped in a single matrix cell or aligned along a common row (Feature/Component) or column (Process/Event). The online SaltFEP database can be downloaded from www.saltfep.org. It contains the FEP matrix, the FEPs, and the associated processes for each FEP. It provides a starting point to create and document site-specific individual FEPs. Furthermore, the FEP matrix is connected to the Salt Knowledge Archive (SKA), a database of about 20,000 references and documents representing the historical knowledge on radioactive disposal in salt. This work is the result of an ongoing collaboration between researchers in the U.S., the Netherlands, and Germany, and supports the NEA Salt Club Mandate. It builds upon prior work which is documented.

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The User Guide for the ComPro Database

An extensive database of glass composition and durability data has been compiled at Savannah River National Laboratory to support the development of nuclear waste glasses. This database is referred to as the Glass Composition-Properties Database (ComPro). The ComPro Database, Revision 3, contains 14,134 total rows of data and 125 columns of composition, durability, as defined by the Product Consistency Test, and other fabrication and characterization information, if available, for each glass. Of the 14,134 total rows, 8,484 rows have been classified as “Model” data and 5,650 rows have been classified as “Non-Model” data. An integral supplement to the ComPro database is the User Guide. The User Guide was developed as a tool to aid the End User in a more effective use of the ComPro database. The User Guide provides a road-map of the specific datasets that comprise the ComPro database (both “Model” and “Non-Model” data) as well as a technical basis for the terminology and definitions the End User will encounter. In this report, a general description of the format and information contained in the User Guide is provided. In addition, specific terminology used in the User Guide is also discussed.

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Data Qualification Report: SRNL Glass Composition-Properties (ComPro) Database

The Savannah River National Laboratory Glass Composition-Properties (ComPro) database is an extensive database containing pertinent composition and durability data to support the accelerated clean-up mission at the Defense Waste Processing Facility. The activities described in this data qualification report were performed to support the information contained in the database. There were two objectives of the original data qualification process. The first objective was to review supporting documentation to determine if DOE/RW-0333P Quality Assurance Requirements and Description had been implemented during the original work. If the DOE/RW-0333P Quality Assurance Requirements and Description had not been directly implemented during the original work, the second objective was to determine if the controls that were used were adequate to meet the intent of the DOE/RW-0333P Quality Assurance Requirements and Description. The results of these two objectives and the activities performed to support these decisions are described in this document. An assessment of each dataset was made to determine if the data were RW-0333P Compliant, RW-0333P Equivalent or Non-RW-0333P Compliant. The original data qualification was performed in accordance with E7, Conduct of Engineering Manual, Procedure 3.70, Revision 4, Qualification of Data. The specific method that was used was Equivalent Controls as described in E7, 3.70. Revision 2 of this document adds supporting information for the RW-0333P Compliant datasets added to Revision 3 of the database.

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LANL Waste Management

All waste at Los Alamos National Laboratory (LANL or the Laboratory) is subject to one or more types of regulations and must be managed properly. This document describes LANL requirements for waste generated and managed by waste generators and Treatment and Storage Facilities (TSFs) to ensure compliance with legal mandates and Laboratory requirements as necessary to protect human health, safety, and the environment. This document has been revised as part of a process in which the Laboratory systematically plans, documents, executes, and evaluates its management of regulated waste streams.

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Acceptable Knowledge Technologist Procedure

Process acceptable knowledge (AK) is a waste stream characterization method that can be used to meet all or part of the waste analysis requirements appropriate for the waste. The method may include documented knowledge of process (KOP), additional characterization data, and/or facility records of analysis. LANL will use process AK to ensure chemical compatibility; assign matrix parameter categories and Environmental Protection Agency (EPA) hazardous waste numbers (HWNs) commonly referred to as EPA codes to waste streams; and determine the waste material parameters and radionuclides present in waste streams.

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Refinement of Pitting Factor Basis to Support the Corrosion Control Program (Interim Report)

At Savannah River Site (SRS), High-Level Waste is stored in below-grade tanks constructed of carbon steel. This waste is composed of sludge, salt cake, and/or supernate. In part, preparation of this waste for future processing involves dissolution of the salt cake layer. The salt dissolution process can create conditions that leave the carbon steel tanks susceptible to a number of corrosion processes. The salt to be dissolved contains high concentrations of nitrate, that once released, create an environment that may be conducive to pitting corrosion or stress corrosion cracking (SCC) of carbon steel. Additionally, during the salt dissolution process, in the absence of mixing, stratification of the supernatant liquid may occur. This can result in less dense, more dilute waste layers occurring higher in the tank. In these more dilute waste layers, the susceptibility to localized corrosion could potentially differ from that of the more concentrated salt solutions evaluated in previous testing, as the amount of inhibiting and aggressive species, not just the ratios, can affect susceptibility. Evaluation of the susceptibility to localized corrosion in these more dilute waste chemistries could provide insight into the amount of inhibitors required to effectively mitigate pitting corrosion in stagnant salt dissolution environments, as well as other tank farm operations involving dilute waste streams.

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Characterization of Infrequent Samples from the Concentration, Storage, and Transfer Facility: Leak Detection Box (LDB) Drain Cell Sample: May 15, 2023, Sample

Savannah River Mission Completion Engineering (SRMC-E) requested that the Savannah River National Laboratory (SRNL) analyze the Concentration, Storage, and Transfer Facility (CSTF) samples from the following Tank Farm areas: the sump encasement, catch tank, drain cell, and waste tank annulus. In general, these CSTF samples will be analyzed on an infrequent basis and analyses will include detection for total beta/gamma activities, total alpha activity, free hydroxide, and pH measurement. SRMC-E corresponded with SRNL to obtain the final results through memo SRNL-L3120-2023-00013. This report presents characterization from results for the leak detection box (LDB) May 15, 2023, drain cell sample. The sample was clear and colorless with no visible particulates. The results are measurements for total gamma, total alpha, total beta, free hydroxide, pH, and density. These analyses were performed in triplicate. A summary of the average analytical results for the LDB sample includes the following. The measured pH for the LDB May 15, 2023 "as-received" drain cell sample range was 6.60-6.89, and the free hydroxide concentration as expected, based upon the pH, was <0.02 M. The average measured density of the "as-received" drain cell sample determined at 22 °C was 1.022±0.001 g/mL. the average measured total alpha activity for the LDB May 15, 2023, sample was 3.04E+02±1.13E+01. This value is less than 4.83E+03 dpm/mL, which is the procedural limit for non-waste determination. The total empirical activity of the beta and gamma emitting (represented by by the sum of total beta and B-137 m activities) averaged to 1.83+04±1.07E+03 dpm/mL. This value is less than 8.69E+05 dpm/mL, which is the procedural limit for non-waste determination.

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Systems Engineering and Analysis in Support of a US Federal Staging Facility for UNF

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

Joseph, Robert↗

Implementation of the Glycolate Destruction Process with Sodium Permanganate in the Defense Waste Processing Facility (DWPF) - 20141

The Savannah River Site (SRS) seeks to replace formic acid with glycolic acid as a chemical reductant in the Defense Waste Processing Facility (DWPF), where borosilicate glass is mixed with high-level radioactive waste, melted at high temperatures, then poured into stainless steel canisters for safe storage. Prior research and development have demonstrated the feasibility and advantages of the new nitric-glycolic acid flowsheet over the current nitric-formic acid flowsheet to chemically adjust the radioactive sludge slurry waste in the Sludge Receipt and Adjustment Tank (SRAT) prior to vitrification. Use of glycolic acid reduces hydrogen and ammonia generation during the reduction process, thus reducing flammability hazards and purge requirements in the DWPF processing vessels. In addition, glycolic acid will support the new Salt Waste Processing Facility (SWPF) by allowing for receipt of higher volumes of waste due to reduction of flammable gases, increased boil-up rates in the vessels, and easing the processing of future sludge batches containing high levels of mercury. However, during DWPF operations, trace amounts of glycolate are anticipated to be entrained in the recycle stream and sent to the Concentration, Storage, and Transfer Facilities (CSTF) via the Recycle Collection Tank (RCT). The RCT receives condensate from the Slurry Mix Evaporator Condensate Tank (SMECT) and Off-gas Condensate Tank (OGCT) that may contain unreacted glycolate. A literature review found the potential for hydrogen generation due to thermolysis of glycolate in caustic CSTF conditions. Savannah River National Laboratory (SRNL) conducted several tests on simulant and radioactive waste from various tanks at SRS and confirmed that glycolate added to waste at concentrations of approximately 1000 mg/L and higher generates hydrogen through thermolysis. There are potential operating scenarios in the future where glycolate concentration could exceed 1000 mg/L in the 2H Evaporator Drop Tank, the highest accumulation point for DWPF organics received via recycle. These findings presented a need to develop another process to mitigate the impact of glycolic acid to the CSTF. Several DWPF compatible options were evaluated for their ability to destroy glycolate in the DWPF recycle stream under various processing conditions, thereby mitigating its introduction to the CSTF. Downselection testing identified sodium permanganate as the most promising and simplest option to oxidize glycolate. It has been demonstrated to be effective in various RCT simulant compositions, operating temperatures, oxidant addition rates, solution pH, and initial glycolate concentrations. A feasibility study identified existing tanks within the facility that could be used for receiving and storing the commercially available sodium permanganate, as well as feeding it to the RCT for processing. The use of existing equipment as well as the ability to purchase the pre-mixed chemical reduces implementation and operational complexity. The downstream impacts of the sodium permanganate reaction were also evaluated and demonstrated to have minimal impact on other chemical species present in the RCT and CSTF, as well as on the materials of construction of the RCT and CSTF vessels and associated piping and instrumentation. Future work is in progress to validate the glycolate destruction process with testing in actual radioactive RCT waste before implementation in DWPF. (authors)

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Segmented Gamma Scanner for Radioactive Waste Assessment: A User Guide

Radioactive waste assessment is important for ensuring nuclear material security at various types of facilities, such as enrichment, fuel fabrication, and reprocessing plants. The waste generated at such nuclear facilities is stored in standard containers and is required to be characterized for material-accounting purposes. The segmented gamma scanner system is a popular, nondestructive analysis measurement system used for characterizing nuclear material, including radioactive waste. This document provides guidance on how to achieve effective performance from a segmented gamma scanner system for accurately quantifying fission products, activation products, and transuranic wastes.

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332TFAD04 Tank Closure Plan

Enclosed is the 332TFAD04 Aboveground Tank Closure/Removal Plan for a tank regulated under Spill Prevention, Controls, and Countermeasures (SPCC) and Aboveground Petroleum Storage Act (APSA) regulations. This tank held diesel fuel for an emergency generator at Building 332. The generator is being replaced with a new unit that contains a belly tank, therefore this aboveground tank will no longer be in use and is being removed. The tank will be SPCC permanently closed on June 2, 2020.

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Section 106 Recordation, Interpretation, and Documentation for the Demolition of Buildings 9201-5 and 9204-4, Y-12 National Security Complex, Oak Ridge, Tennessee

In 2019 and 2020, Cultural Resource Analysts, Inc. (CRA), began preparing recordation packages for 18 World War II and Cold War Era buildings slated for demolition at the request of Consolidated Nuclear Security, LLC (CNS), on behalf of the U.S. Department of Energy (DOE), National Nuclear Security Administration (NNSA). The recordations for these 18 buildings, which are located within what is now known as the Y-12 National Security Complex (Y-12) in Oak Ridge, Tennessee, were prepared and submitted in two separate packages. The first report (EC-NP-004) was prepared as a mitigation measure to address adverse effects that will result from the proposed demolition of 16 support and ancillary buildings. The current report is the second of the two packages and addresses the last two of the 18 buildings slated for demolition (9201-5 and 9204-4), which are two of the large process buildings located at Y-12. All 18 of these buildings have previously been determined eligible for listing in the National Register of Historic Places (NRHP). The two buildings, Building 9201-5 and Building 9204-4, are located south of Bear Creek Road and west of the main Y-12 entrance gate. Both buildings were constructed as uranium enrichment facilities in support of the Manhattan Project, a top-secret World War II mission to develop the world’s first nuclear weapons. Building 9201-5, completed in October 1944, is the Alpha-5 process facility and Building 9204-4, completed in Fall 1945, is the Beta-4 process facility. Over the years, these facilities were repurposed for a number of different functions.

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Request for Exemption and Permit Modification for S-6008: Sewer Diversion Complex

Lawrence Livermore National Laboratory (LLNL) operates its Sewer Diversion Complex (SDC) under two relevant permits: discharges to air are controlled by Condition 5533 of LLNL’s BAAQMD permit for Source 6008, and discharges to the municipal sewer system are controlled by LLNL’s Wastewater Discharge Permit (Permit #1250). LLNL intends to modify the SDC by changing the currently permitted tanks to new models. Generally, this change will increase the volume of the tanks, update the abatement carbon system to suit the larger volume, and switch to a set of taller but thinner tanks.

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