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DOE OSTI2020

Selection of Well Locations for Saturated Zone Monitoring at the E-Area Low Level Waste Facility

Savannah River Site (SRS) has adopted a revised approach to performance monitoring of the E- Area Low Level Waste Facility (ELLWF) that adds a saturated zone component to the existing vadose zone monitoring program (Kubilius, 2019). The revised monitoring approach specifies that eight additional water table wells be installed near Slit Trench 1 (ST1) and Engineered Trench 2 (ET2). Six wells will be installed at locations characterized in 2017 and 2018. These include four wells near ET2 (ELF001D, ELF002D, ELF003D, and ELF004D) at locations identified by Kubilius (2019) and two wells north of ST1 (ELF005D and ELF006D) at locations finalized in this report. Additionally, two wells (ELF007D and ELF008D) are to be installed at locations south of ST1 based on characterization work completed in 2020 and documented in this report. This area is near well cluster BGX-2 between the Mixed Waste Management Facility (MWMF) and the ELLWF. The area is upgradient of ELLWF and, contamination in the water table aquifer is sourced from MWMF.

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DOE OSTI2021

VERSE-LC TCCR Model Calibration using Batch 1A, Batch 2, and Batch 3 Column Data

The objective of this work is to utilize on-line gamma monitoring data collected during Tank Closure Cesium Removal (TCCR) Column Batch 3 operations to calibrate the parameters (essentially, the correction/dilution factors (CF or DF)) used in the VERSE-LC models for the prediction of cesium breakthrough while processing Savannah River Site (SRS) Tank 10H dissolved saltcake High-Level Waste through a packed cylindrical bed of Crystalline Silicotitanate (CST). Calibrated VERSE-LC models could be applied to aid in evaluating future TCCR operations.

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DOE OSTI2022

LAW Primary Offgas Process (LOP) and LAW Secondary Offgas/Vessel Vent Process (LVP) System Design Description

This system design description (SDD) defines the technical, functional, and performance requirements of the Low-Activity Waste (LAW) Facility primary offgas process system (LOP) and the LAW Facility secondary offgas/vessel vent process system (LVP). This document details waste treatment requirements, environmental compliance requirements, and authorization basis requirements of the LOP and LVP systems as they are currently known and understood.

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DOE OSTI2023

Summary of SRNL Support to the DOE-ORP Enhanced Waste Glass Program for FY23

In fiscal year 2023 (FY23) Savannah River National Laboratory (SRNL) continued tasked work for the Office of River Protection (ORP) to expand glass compositional regions accessible for low-activity waste (LAW) and high-activity waste (HLW) vitrification processing. Experimental work continued in four primary technical areas focused on processing and performance of glasses relevant to the Hanford missions. The data and results from this work will be used to expand and validate the glass models being developed at Pacific Northwest National Laboratory (PNNL) for waste processing and acceptance. This report summarizes the activities and deliverables associated with work performed in FY23.

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DOE OSTI2026

A Brief Review of Technologies for Separating Tritium from the Aqueous Effluent of Reprocessing Plants

The scope of this letter report encompasses a review of technologies applicable to separating tritiated water from light and heavy water with a focus on wastewater arising from the reprocessing of UNF. Reports considering flowsheet options for tritium management are also reviewed. The main objective of this letter report is to support consideration of programmatic opportunities for the Materials Recovery and Waste Form Development (MRWFD) Campaign within the Department of Energy’s Office of Nuclear Energy.

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DOE OSTI2023

FY23 PFLOTRAN Development Support: GDSA PFLOTRAN

This report details the FY23 contributions to PFLOTRAN development by PNNL code developers (i.e., Dr. Glenn Hammond) under the Spent Fuel Waste Science and Technology Campaign within the US Department of Energy Office of Nuclear Energy to support the Geologic Disposal Safety Assessment (GDSA) Framework.

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DOE OSTI2024

Consolidated Interim Storage Advantages and Disadvantages from Prior Reports and Studies

The question of whether centralized storage of civilian spent nuclear fuel (SNF) should be part of the federal waste management system as an intermediate step before permanent disposal has been debated for more than four decades. Since the Nuclear Waste Policy Act (NWPA) (U.S. House of Representatives 1982) was enacted, the need for interim storage facilities (ISFs) as part of the federal waste management system has been analyzed and documented in numerous reports and studies. This paper summarizes the advantages and disadvantages as described in those reports of incorporating an ISF into the waste management system.

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DOE OSTI2023

Overview of System Integration Analysis Activities for Integrated Waste Management

Spent nuclear fuel (SNF) generated by the current fleet of commercial nuclear reactors is being stored at the reactor sites in spent fuel pools (SFPs) and in dry independent spent fuel storage installations (ISFSIs). The US Department of Energy Office of Nuclear Energy (DOE-NE) is developing an Integrated Waste Management Program (IWMP) comprising a suite of options and supporting analyses to enable future informed choices. The IWMP is organized into the following five major areas: 1) Consent-Based Siting, 2) IWM Facilities and Equipment Concepts and Development, 3) Transportation Capability Analysis and Support, 4) Information Technology Solutions and Support, and 5) System Integration Analysis and Support. This paper discusses the activities ongoing in the IWMP System Integration Analysis and Support area. Two main areas of research in system integration are data and tools development, as well as system analysis assessments. One of the tools being developed in the system integration area is the Used Nuclear Fuel-Storage, Transportation & Disposal Analysis Resource and Data System (UNF-ST&DARDS) tool. It is being developed as a foundational resource for DOE-NE to manage SNF data, along with several compatible analysis tools for time-dependent characterization of SNF and related systems. UNF-ST&DARDS has the unparalleled ability to track SNF through the entire back end of the fuel cycle—from the time the fuel is discharged from a reactor through its disposal in a geological repository. UNF ST&DARDS interfaces with the SCALE code system for nuclear analysis and COBRA-SFS for thermal analysis. Another main tool being developed is the Next Generation System Analysis Model (NGSAM). NGSAM is an agent-based simulation software tool expressly designed to be capable of modeling the waste management system. NGSAM has been developed to enable informed decision-making by providing the capability of analyzing various potential system options for the management of SNF and HLW. Using NGSAM, system architecture analyses are being conducted to support the future deployment of a comprehensive nuclear waste management system that considers all major back-end aspects of the nuclear fuel cycle (i.e., transportation, storage, and disposal). System analysis assessments may investigate the implications of various strategies such as different acceptance rates, acceptance queues, facility capacities and options, standardized canisters, and different assumed system operation start dates. Recently, some system analysis effort has begun to look at how the waste management system might operate for advanced reactor fuel cycles.

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DOE OSTI2020

Treatment of Problematic Reactive Metal Wastes Using GeoMelt ICV - WM2020 Conference Paper

Decommissioning of sodium-cooled reactors and fast reactor technologies has generated a number of reactive metal waste configurations that are problematic to treat and typically lack cost effective treatment methods and disposition options. As a result, Veolia Nuclear Solutions, under contract with Idaho National Laboratory (owned by the U.S. Department of Energy and managed and operated by Battelle Energy Alliance, LLC) demonstrated its GeoMelt® In-Container Vitrification (ICV)™ technology to safely convert sodium metal to a non-reactive vitrified oxide form. The demonstration project, supported by glass formulation and crucible testing, consisted of a series of ICV™ melts that processed elemental sodium into stable non-reactive glass. INL is currently implementing GeoMelt® technology as a means to safely and reliably convert radioactive reactive metal residues that contaminate sodium cooled reactor components into waste forms that comply with existing disposition pathways. Reactive metal wastes require treatment in order to remove the Resource Conservation and Recovery Act (RCRA) reactivity and ignitability characteristics to comply with land disposal restrictions. GeoMelt®, which is an alternative to other potential treatment approaches, provides a robust approach that chemically converts the reactive metals to an inert oxide while also immobilizing radionuclides in a vitrified waste form with durability equal to or better than vitrified nuclear fuel reprocessing wastes (very robust and inert waste forms). Most other treatment approaches generate hydrogen gas which is problematic. In 2016, Veolia Nuclear Solutions first demonstrated the effectiveness of the GeoMelt® ICV™ process in deactivating reactive sodium metal. Crucible, bench-scale, and engineering-scale demonstrations were conducted on several surrogate waste configurations with various ratios of sodium metal and glass-formers. Each ratio and configuration demonstrated complete deactivation of the surrogate sodium metal. Follow-on work in 2017 demonstrated the deactivation of reactive sodium by GeoMelt® ICV™ at a higher waste loading relative to previously demonstrated work performed in 2016; the higher waste loading optimized glass chemistry while enhancing the economical full-scale treatment of reactive metals. Additionally, follow-on demonstration testing in 2018 and 2019 focused on more complex shapes and other reactive-metals (mocked up Experimental Breeder Reactor II [EBR-II] subassembly, sodium filled heat exchanger, and a can containing sodium potassium alloy) which were all performed at engineering-scale. Since 2018, a 10-metric ton full-scale GeoMelt unit (GeoMelt® Richland) was designed, installed, and commissioned at Perma-Fix Northwest in Richland, Washington for the treatment of reactive metal wastes. As of September 2019, over 900 55-gallon drums containing a total of around 3,500 lb of sodium with low levels of radioactivity have been treated at GeoMelt ® Richland, with resulting glass monoliths disposed at the Nevada National Security Site (NNSS). A full-scale radiological demonstration melt on an actual EBR-II subassembly has also been performed using the full-scale melter in 2019. The GeoMelt® technology is a proven radioactive waste treatment technology capable of immobilizing radioactive wastes, including bulk rubble such as drums and other steel vessels usually without pretreatment. Utilizing the GeoMelt® technology to treat reactive metals eliminates pretreatment steps resulting from having to separate the reactive metal from steel containers or jackets as GeoMelt® can easily operate at temperatures sufficient to melt the steel and expose the reactive metal for treatment. Eliminating handling steps of reactive metals is a significant safety advantage since reactive metals are pyrophoric. The results generated as a part of the 2018-2019 demonstration program are presented in the paper.

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DOE OSTI2020

Aging of Concrete for the Evaluation of Repair Materials to Protect the Walls of the HCAEX Tunnel at Savannah River - 20301

H-Canyon, located in the Savannah River site, is a unique facility for chemical reprocessing of plutonium, highly-enriched uranium, and other radioactive materials. The exhaust gases of the H-Canyon are sent through the H-Canyon Exhaust (HCAEX) tunnel for contamination removal. Robotic inspections of the tunnel revealed significant degradation of the reinforced concrete structure that was associated with acid attack, and could compromise the structural stability of the tunnel. Thus, the identification and evaluation of potential repair materials that could be applied on the degraded walls to mitigate and prevent further degradation is of significant interest to the Department of Energy and the Savannah River representatives. This research effort has been divided into two phases: 1) Development and evaluation of aged concrete under accelerated aging conditions (which is the focus of this paper) and 2) Evaluation of potential repair materials applied on aged and non-aged concrete under simulated aggressive conditions. In order to develop and evaluate concrete samples exposed to accelerated aging conditions in simulated aggressive environments, a literature review of the HCAEX tunnel was conducted that included 1) characterization and extent of the concrete damage, 2) environmental conditions inside the tunnel and 3) primary deterioration mechanisms. In addition, potential coatings and/or repair materials for degraded concrete surfaces exposed to aggressive environments (primarily acidic) were selected from the literature review and the most common testing and measurements for evaluating acid attack phenomena, erosion, etc. were reviewed. From the literature review findings, a preliminary bench-scale test plan for the concrete aging was developed, including accelerated aging tests with aggressive conditions (acid fumes, humidity, etc.). Concrete samples were exposed to the aging accelerated conditions (e.g. immersion in acid solutions) and visual inspection, mass loss and pH change (acid solution) were recorded over time. Correlations between the visual inspection, mass loss and pH changes results and the aging time or the aging conditions were developed. Specimens submitted to the highest acid concentration showed the fastest and most intense degradation. The type of coarse aggregate (limestone) used for the concrete seemed to be the cause of the fastest aging observed, compared to the cement paste. The research findings created the foundation for the ongoing investigation, in which new concrete samples with a mix design similar to the HCAEX tunnel will be tested and will serve as the substrate for testing the selected coatings and /or repair materials. In this paper, the literature review and preliminary results from the aging tests are provided. (authors)

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DOE OSTI2020

Los Alamos National Laboratory Hazardous Waste Facility Permit Community Relations Plan

The requirements of the Hazardous Waste Facility Permit are fulfilled through execution of LANL's fourth strategic goal: “Enabling mission delivery through next-generation facility, infrastructure, and operational excellence.” With input from our stakeholders, we will continue to fulfill those requirements and to improve our waste operations. In 2015, the Laboratory purpose statement focused waste management efforts by directing delivery of mission success through operational effectiveness and scientific excellence. This Plan advances that purpose through community involvement. In the process of achieving our national security mission, Los Alamos National Laboratory generates some hazardous and mixed waste. Hazardous waste is solid waste that is dangerous or potentially harmful to human health or the environment. Hazardous wastes can be liquids, solids, gases, or sludges. They can be discarded as commercial products, such as cleaning fluids or pesticides, or as the byproducts of operations. Specific substances are listed in 40 Code of Federal Regulations (CFR), Part 261: 40 CFR Part 261. Hazardous waste management activities are regulated by the U. S. Environmental Protection Agency and the New Mexico Environment Department (NMED) pursuant to New Mexico Hazardous Waste Act (HWA; Chapter 74, Article 4 NMSA 1978) and regulations under the Act. In 1989, NMED issued the Hazardous Waste Facility Permit (EPA ID Number NM0890010515-1) that established standards for the Laboratory to manage, store, and treat hazardous wastes on-site and to undertake the closure, post closure care, and cleanup as necessary, of permitted waste management units. On November 30, 2010, NMED renewed the Hazardous Waste Facility Permit (the Permit). In June 2020 the Permittees submitted a permit renewal application to the NMED for their review and approval. The Hazardous Waste Facility Permit Community Relations Plan (CRP) describes the scope of public involvement in the activities of the Permit. The CRP is specifically designed to facilitate the community outreach, engagement, and relations activities concerning the Laboratory’s Hazardous Waste Facility Permit and coordinates with but does not include public involvement for other Laboratory programs, initiatives, or environmental activities.

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DOE OSTI2021

Management of New Mexico Special Waste

New Mexico “Special Waste” (NMSW) is a non-hazardous solid waste that has unique handling, transportation, and/or disposal requirements to assure protection of the environment and the public health, welfare, and safety.

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DOE OSTI2023

Performance Assessment for the Disposal of Low-Level Waste in the 200 East Area Burial Grounds (Annual Status Report FY 2022)

This annual review provides the projected dose estimates of radionuclide inventories disposed in the 200 East Area Low-Level Waste Burial Grounds (LLBGs) since September 26, 1988. These estimates are calculated using the original dose methodology developed in the performance assessment (PA) analysis (WHC-SD-WM-TI-730). The estimates are compared with the performance objectives defined in U.S. Department of Energy requirements (DOE O 435.1 and its companion documents DOE M 435.1-1 and DOE-STD-5002-2017). All performance objectives are currently satisfied, and operational waste acceptance criteria (HNF-EP-0063) and waste acceptance practices continue to be sufficient to maintain compliance with performance objectives. Inventory estimates and associated dose estimates from future waste disposal actions are unchanged from previous years’ evaluations that indicate potential impacts well below performance objectives; therefore, future compliance with DOE O 435.1 is expected.

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DOE OSTI2022

Performance Assessment for the Disposal of Low-Level Waste in the 200 West Area Burial Grounds (Annual Status Report FY 2021)

This annual review provides the projected dose estimates of radionuclide inventories disposed in the active 200 West Area Low-Level Waste Burial Grounds (LLBGs) since September 26, 1988. These estimates are calculated using the original dose methodology developed in the performance assessment (PA) analysis (WHC-EP-0645). The estimates are compared with the performance objectives defined in U.S. Department of Energy requirements (DOE O 435.1 and its companion documents DOE M 435.1-1 and DOE-STD-5002-2017). All performance objectives are currently satisfied, and operational waste acceptance criteria (HNF-EP-0063). and waste acceptance practices continue to be sufficient to maintain compliance with performance objectives. Inventory estimates and associated dose estimates from future waste disposal actions are unchanged from previous years’ evaluations that indicate potential impacts well below performance objectives; therefore, future compliance with DOE O 435.1 is expected.

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DOE OSTI2025

Guiding Principles for Geochemical/Thermodynamic Model Development and Validation in Nuclear Waste Disposal: A Close Examination of Recent Thermodynamic Models for H + —Nd 3+ —NO 3 - (—Oxalate) Systems

Development of a defensible source-term model (STM), usually a thermodynamical model for radionuclide solubility calculations, is critical to a performance assessment (PA) of a geologic repository for nuclear waste disposal. Such a model is generally subjected to rigorous regulatory scrutiny. In this article, we highlight key guiding principles for STM model development and validation in nuclear waste management. We illustrate these principles by closely examining three recently developed thermodynamic models with the Pitzer formulism for aqueous H + —Nd 3+ —NO 3 - (—oxalate) systems in a reverse alphabetical order of the authors: the XW model developed by Xiong and Wang, the OWC model developed by Oakes et al., and the GLC model developed by Guignot et al., among which the XW model deals with trace activity coefficients for Nd(III), while the OWC and GLC models are for concentrated Nd(NO 3 ) 3 electrolyte solutions. The principles highlighted include the following: (1) Principle 1. Validation against independent experimental data: A model should be validated against experimental data or field observations that have not been used in the original model parameterization. We tested the XW model against multiple independent experimental data sets including electromotive force (EMF), solubility, water vapor, and water activity measurements. The results show that the XW model is accurate and valid for its intended use for predicting trace activity coefficients and therefore Nd solubility in repository environments. (2) Principle 2. Testing for relevant and sensitive variables: Solution pH is such a variable for an STM and easily acquirable. All three models are checked for their ability to predict pH conditions in Nd(NO 3 ) 3 electrolyte solutions. The OWC model fails to provide a reasonable estimate for solution pH conditions, thus casting serious doubt on its validity for a source-term calculation. In contrast, both the XW and GLC models predict close-to-neutral pH values, in agreement with experimental measurements. (3) Principle 3. Honoring physical constraints: Upon close examination, it is found that the Nd(III)-NO 3 association schema in the OWC model suffers from two shortcomings. Firstly, its second stepwise stability constant for Nd(NO 3 ) 2+ (log K 2 ) is much higher than the first stepwise stability constant for NdNO 3 2+ (log K 1 ), thus violating the general rule of (log K 2 –log K 1 ) < 0, or $\frac{K1}{K2}$>1. Secondly, the OWC model predicts abnormally high activity coefficients for Nd(NO 3 ) 2 + (up to ~900) as the concentration increases. (4) Principle 4. Minimizing degrees of freedom for model fitting: The OWC model with nine fitted parameters is compared with the GLC model with five fitted parameters, as both models apply to the concentrated region for Nd(NO 3 ) 3 electrolyte solutions. The latter appears superior to the former because the latter can fit osmotic coefficient data equally well with fewer model parameters. The work presented here thus illustrates the salient points of geochemical model development, selection, and validation in nuclear waste management.

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DOE OSTI2020

Investigation and Analysis of Dolomite Dissolution in Variable Strength Systems Relevant to the WIPP

The Waste Isolation Pilot Plant (WIPP) was opened in 1999 as a solution to the long-term disposal of the nation's nuclear defense-generated transuranic (TRU) waste. The salt beds were recommended as the ideal location because they are free from flowing fresh water, easily mined, impermeable, and geologically stable. However, for performance assessment modeling, release scenarios must be considered. The most likely release scenario was determined to be human intrusion leading to direct and/or long-term brine release (US DOE 1995, US DOE 1996, Perkins et al., 1999). The focus of this research is the Rustler formation because it is located above the WIPP and is the most transmissive layer (Perkins et al., 1999). Objective: To determine the effect of chelating ligands ( e.g. EDTA) on dissolution of dolomite in high ionic strength systems. Do brines impact dolomite dissolution and how is it affected by EDTA binding with cations? By varying the ionic strengths, different brine conditions can be simulated similar to those in and around the Culebra member and evaluate trends of dissolved Ca and Mg in the aqueous phase from the quantified data. Conclusions: <2% of dolomite dissolved over one week at pH 8.5. Greater dissolution occurred at higher ionic strength and in the presence of NaCl with approximately 0.4 vs. 0.9% of Mg removed from dolomite in 0.1 and 1.0 M Na. Further, 0.3 vs. 0.9% of Mg was removed from dolomite in 1.0 M IS CaCl{sub 2} versus NaCl. EDTA increased dissolution of dolomite at low ionic strength due to its strong complexation affinity for Ca/Mg with slightly greater removal of Ca than Mg. Significant differences were not observed at high ionic strength due to the presence of large concentrations of competing cations (Na, Ca, Mg). Slightly higher concentrations of Ca were removed from dolomite than Mg after one week, suggesting either incongruent dissolution or release through ion exchange of Ca from trace mineral impurities. Further, when EDTA was present, it may have increased Ca removal due to its greater complexation for Ca than Mg. Environmental Implications: High ionic strength brines and low ionic strength solutions with organic ligands increase dissolution of dolomite. Due to the long half-lives of radionuclides in the TRU waste, it may be important to consider the long term impact of dolomite dissolution for potential release scenarios.

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