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DOE OSTI2022

Idaho National Laboratory CY 2021 National Emission Standards for Hazardous Air Pollutants Analysis, Methodology and Results for Radionuclides

This report details calculations of potential dose at public receptor locations surrounding the Idaho National Laboratory (INL) Site boundary, and INL in-town facilities, from radionuclides reported to be in use and potentially emitted from INL facilities during calendar year (CY) 2021. All calculations were performed in accordance with the requirements in Code of Federal Regulations (CFR), Title 40, “Protection of the Environment,” Part 61, “National Emission Standards for Hazardous Air Pollutants (NESHAPs),” Subpart H, “National Emission Standards for Emissions of Radionuclides Other than Radon from Department of Energy Facilities” (40 CFR 61, Subpart H). Modeling methodology, model input parameters, and contribution to dose by facility, source, and radionuclide at the maximally exposed individual (MEI) location are also discussed. The information in this report supports the “National Emission Standards for Hazardous Air Pollutants - Calendar Year 2021 INL Report for Radionuclides” (INL 2022). In CY 2021, the estimated annual potential dose at the INL Site MEI location was 6.67E-02 mrem/yr, up slightly from the previous year, but far less than the regulatory standard of 10 mrem/yr (CFR 40 Part 61, Subpart H). Approximately 97% of the total dose to the INL Site MEI originated from MFC sources. Emissions from INL in-town facilities resulted in an estimated annual potential dose of 6.21E-03 mrem/yr to the MEI, down 40% from the CY 2020 estimated dose. Year-to-year variations in estimated annual dose can be attributed to adjustments in laboratory operations, changes to facility infrastructure, and variation in meteorological conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Idaho National Laboratory CY 2022 National Emission Standards for Hazardous Air Pollutants Analysis, Methodology and Results for Radionuclides

This report details calculations of potential dose at public receptor locations surrounding the Idaho National Laboratory (INL) Site boundary, and INL in-town facilities, from radionuclides reported to be in use and potentially emitted from INL facilities during calendar year (CY) 2022. All calculations were performed in accordance with the requirements in Code of Federal Regulations (CFR), Title 40, “Protection of the Environment,” Part 61, “National Emission Standards for Hazardous Air Pollutants (NESHAPs),” Subpart H, “National Emission Standards for Emissions of Radionuclides Other than Radon from Department of Energy Facilities” (40 CFR 61, Subpart H). Modeling methodology, model input parameters, and contribution to dose by facility, source, and radionuclide at the maximally exposed individual (MEI) location are also discussed. The information in this report supports the “National Emission Standards for Hazardous Air Pollutants – Calendar Year 2022 INL Report for Radionuclides” (DOE-ID 2023). In CY 2022, the estimated annual potential dose at the INL Site MEI location was 1.78E-02 mrem/yr, down from the previous year, and far less than the regulatory standard of 10 mrem/yr (CFR 40 Part 61, Subpart H). Approximately 87% of the total dose to the INL Site MEI originated from Materials and Fuels Complex sources. Emissions from INL in-town facilities resulted in an estimated annual potential dose of 4.03E-03 mrem/yr to the MEI, down 35% from the CY 2021 estimated dose. Year-to-year variations in estimated annual dose can be attributed to adjustments in laboratory operations, changes to facility infrastructure, and variation in meteorological conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Retention of LDR Inorganics in Solidified/Immobilized Waste

This report provides a summary of the laboratory data and statistical analysis used to calculate retention factors in solidified/stabilized Hanford tank waste for the 13 LDR inorganic species associated with Hanford tank waste (SRNL-STI-2020-00228). The data presented is a summary of the results of research performed to build a correlation between the untreated waste concentration and the TCLP response of a solidified/stabilized waste form. This report establishes the efficacy of two treatment technologies found in 40 CFR 268.42, CHRED (chemical reduction) and STABL (stabilization), that are key to the Sample-and-Send strategy.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Computational Materials Design for Ceramic Nuclear Waste Forms Using Machine Learning, First-Principles Calculations, and Kinetics Rate Theory

Ceramic waste forms are designed to immobilize radionuclides for permanent disposal in geological repositories. One of the principal criteria for the effective incorporation of waste elements is their compatibility with the host material. In terms of performance under environmental conditions, the resistance of the waste forms to degradation over long periods of time is a critical concern when they are exposed to natural environments. Due to their unique crystallographic features and behavior in nature environment as exemplified by their natural analogues, ceramic waste forms are capable of incorporating problematic nuclear waste elements while showing promising chemical durability in aqueous environments. Recent studies of apatite- and hollandite-structured waste forms demonstrated an approach that can predict the compositions of ceramic waste forms and their long-term dissolution rate by a combination of computational techniques including machine learning, first-principles thermodynamics calculations, and modeling using kinetic rate equations based on critical laboratory experiments. By integrating the predictions of elemental incorporation and degradation kinetics in a holistic framework, the approach could be promising for the design of advanced ceramic waste forms with optimized incorporation capacity and environmental degradation performance. Such an approach could provide a path for accelerated ceramic waste form development and performance prediction for problematic nuclear waste elements.

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

Nevada National Security Site V3XA Spherical Explosive Confinement Vessel Awareness Package

This report summarizes the plans and activities that will facilitate shipment of the V3XA experimental spheres (a.k.a., Nevada Spheres) from the Nevada National Security Site (NNSS) to the Idaho National Laboratory’s (INL) Advanced Mixed Waste Treatment Project (AMWTP). The spheres require shipment to INL to be segmented, characterized, and repackaged for ultimate disposal at the Waste Isolation Pilot Plant (WIPP). The NNSS does not currently have the facilities or safety basis to perform these activities. The INL has the facilities, the INL M&O contractor owns an 8-120B cask, and AMWTP has a WIPP Waste Acceptance Criteria approved characterization program. The NNSS has shipped other legacy transuranic waste to INL for WIPP certified characterization in the past.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

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) [1] 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 presentation summarizes the advantages and disadvantages, as described in those reports, of incorporating an ISF into the waste management system.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Technology Applicable to Remote Hazardous Constituents and Waste Stream Grouping

Originally, Rocky Flats transuranic (TRU) waste was managed from waste streams that were composed of similar chemical constituents and were generated from identical processes. Early waste processing at Idaho National Laboratory (INL) maintained the Rocky Flats record until INL began generating their own item description codes (IDCs). Ultimately TRU waste management at INL was driven to combine wastes from all IDCs that were assigned by Rocky Flats to a summary category group regardless of whether the chemical constituents in the waste were known or not. At this point, the Waste Isolation Pilot Plant (WIPP) certified program requirements to process the TRU waste according to waste streams was nearly eliminated. In April 2018, operations in the Sludge Repackage Project (SRP) area were halted by a radiological materials release accident in which drums containing waste experienced a loss of containment after over-pressurization occurred. Four SRP drums, repackaged in the Accelerated Retrieval Project (ARP) V area, had an adverse, delayed chemical reaction event occur when prohibited materials from one container were mixed with chemically-incompatible materials from other containers in the repackaging process. This incident was the direct result of processing unknowns under a protocol where the components of the waste were added to incompatible materials resulting in a run-away reaction. After a Difficult Waste Team (DWT) and Los Alamos National Laboratory – Carlsbad Operations (LANL-CO) subject matter expert (SME) visit in September 2019, it was determined by the team that all of the work observed on the September visit was performed upstream of the certified program and that the Visual Examination (VE) and other methods typically used by the certified program were being used for Acceptable Knowledge (AK) augmentation. This augmentation was needed in order to determine what the constituents in the waste might be, and is based primarily on collections of VE results. The waste at INL that is being processed in what is considered to be the Priority 2 (yellow) and Priority 3 (red) lists either needs AK to be completed before certified characterization can occur or is waiting to be processed because the waste has been mixed with unknown wastes resulting in a lack of information that could have impact of the ability to certify and ship the waste to the WIPP.

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

Antifoam Development for Eliminating Flammability Hazards and Decreasing Cycle Time in the Defense Waste Processing Facility

The Savannah River National Laboratory (SRNL) was requested to develop a new antifoam control method for the Defense Waste Processing Facility’s (DWPF) Chemical Process Cell (CPC). SRNL completed testing of both chemical and nonchemical foam controls. The nonchemical foam controls were either ineffective (or worse, created more foam) or impractical (a water spray can control foam, but excessive water is needed). As a result, the focus of this study was on finding a superwetter or commercial antifoam for controlling foam. Thirty potential antifoams were tested as part of this study. A series of tests were developed to help screen out ineffective alternatives including: 1. Spreading testing of superspreaders, 2. Foam column testing with physical simulants, 3. Boiling testing with physical and chemical simulants, 4. Days-only Sludge Receipt and Adjustment Tank (SRAT) process simulations with sludge (containing noble metals and mercury), Precipitate Reactor Feed Tank (PRFT), and Slurry Mix Evaporator Feed Tank (SEFT) simulants in the RC1 Reaction Calorimeter (purchased for antifoam testing), and 5. Around-the-clock SRAT and Slurry Mix Evaporator (SME) process simulations with sludge(containing noble metals and mercury), PRFT, and SEFT simulants in the RC1 Reaction Calorimeter. Evonik Surfynol® MD20, a commercially available defoamer, was relatively effective in controlling foam, while remaining chemically stable in SRAT and SME processing across the pH range of 4 to 13. No degradation products were detected in the offgas, in the condensate or in the SRAT and SME products. In nitric-glycolic acid flowsheet testing, 250 mg/kg Evonik Surfynol® MD20 was needed for foam control compared to 1,625 mg/kg for Antifoam 747, DWPF’s current antifoam. In nitric-formic acid flowsheet testing, 1,125 mg/kg of Evonik Surfynol® MD20 was needed to control foam throughout the SRAT and SME cycles. The commercially available superspreader Momentive™ Y-17112 was even more effective than Evonik Surfynol® MD20 as both a defoamer and an antifoam. Not only was the foam destroyed upon addition but also was less persistent between additions. It was the most effective antifoam in testing using both the nitric-glycolic acid flowsheet and the nitric-formic acid flowsheet. In nitric-glycolic acid flowsheet testing, only 100 mg/kg Momentive™ Y-17112 was needed to control foam throughout the SRAT and SME cycles. In nitric-formic acid flowsheet testing, 300 mg/kg Momentive™ Y-17112 was needed to control foam throughout the SRAT and SME cycles. Momentive™ Y-17112 is also resistant to hydrolysis as demonstrated by its chemical stability in SRAT and SME processing across the pH range of 4 to 13 and lack of degradation products in offgas or condensate. Both candidates were effective as potential replacements for Antifoam 747, with Y-17112 demonstrating superior foam control. During nitric-glycolic flowsheet testing 50% less antifoam was needed when using Momentive™ Y-17112 compared to MD20. During nitric-formic flowsheet testing 75% less antifoam was needed when using Momentive™ Y-17112 compared to MD20. Foam remediated with Momentive™ Y-17112 was less persistent throughout testing. In addition, no degradation products were detected in the offgas, in the condensate or in the SRAT and SME products. Based on this testing, Momentive™ Y-17112 is clearly superior to Evonik Surfynol® MD20 and Antifoam 747, especially for the nitric-formic acid flowsheet processing; it is recommended that Momentive™ Y-17112 replace Antifoam 747 in DWPF. An antifoam addition strategy is recommended for both the nitric-glycolic acid flowsheet and the nitric-formic acid flowsheet. Implementation of Momentive™ Y-17112 is expected to decrease SRAT and SME boiling times by up to 50%, eliminate the issues resulting from Antifoam 747 degradation products, and minimize foamovers. To validate the effectiveness of these defoaming agents, SRNL recommends irradiation of a SRAT or SME product simulant containing fresh antifoam. The goal of this testing is to determine whether the irradiation causes decomposition of the antifoam that would make it less effective or produce new species in the offgas or slurry. This testing began in April 2020. An evaluation should be completed to determine the thermolytic hydrogen and methane generation rate in downstream equipment, including the High-Level Waste evaporators.

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

What's the big deal with TA-55's trash?

There's a beast of hazardous waste, and we're learning how to tame it. Plutonium and production activities are ongoing at TA-55, which means the beast of hazardous waste is always looming around the corner. The Laboratory must dispose of this delicate type of waste properly and swiftly. Learning from previous snags and pitfalls in an incredibly complex process, TA-55 leadership and staff have worked together to greatly improve and streamline waste processes. The result: As of the last week of fiscal year 2020, 42 shipments containing 1,275 containers of transuranic (TRU) waste were sent to the Waste Isolation Pilot Plant (WIPP) in southern New Mexico, the nation’s only repository for defense-generated TRU waste. In FY 2019, we shipped less than half that amount.

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

Annual Update Pantex Plant Site Treatment Plan/Compliance Plan for Development of Mixed Waste Treatment Capacities and Technologies

This updated STP identifies how DOE proposes to obtain commercial treatment or as appropriate, develop technologies for the site's mixed waste in inventory that does not currently have a developed disposition path. This updated STP also provides the methodology to be implemented if mixed waste is generate or identified that requires development of a disposition path.

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

Removable, Sequestration Coatings for Mitigating Hazardous Contaminants Related to Deactivation and Decommissioning Activities

This project is a collaborative effort between InnoSense LLC (ISL) and Oak Ridge National Laboratory (ORNL) aimed at mitigating the mercury (Hg) contamination problem at the Y-12 National Security Complex (NSC). The report summarizes the research and technology development work of the Phase IIB Project, led by Dr. Uma Sampathkumaran at ISL for the period of July 31, 2017–October 31, 2021. The Y-12 site cleanup includes technologies for safe and cost-effective facility demolition, soil remediation, and on-site treatment and disposal of mercury bearing wastes. One of the technology demonstration goals was a reactive strippable coating for targeted pre-demolition and de-contamination to limit Hg mobilization and the volume of debris requiring treatment for disposal. Toward this, ISL has developed a water-based reactive strippable coating (Trap & See-Hg™) to sequester elemental Hg(0) and mercuric Hg(II) species in the coating. The coating provides multiple benefits including (1) Hg vapor suppression, (2) sealing in the hazardous contaminants, (3) visually locating the contaminant through color changes, and (4) remediating the contaminated surfaces upon coating removal. Through collaboration between ORNL and ISL in Phase II work, the team successfully demonstrated the Trap & See-Hg technology to (1) visually locate the presence of Hg on concrete, soil and drywall surfaces, (2) effectively sequester all mercury species (elemental, soluble and insoluble Hg) from lab-generated and ex-situ soil and brick/rubble samples collected from the Y 12 site with 80 to 95% efficiency to capture/reduce Hg vapor concentration, (3) be applied by manual and robotic spray applicators and cure at room ambient, (4) remain a cohesive and peelable coating upon drying, and (5) sequester Hg0 from walls in the alpha-2 building basement at Y-12 during a small scale in-situ field trial. In Phase IIB, ISL’s effort was focused on refining the coating formulations with additives containing passive and reactive sorbents, colorimetric indicator, and thixotropic agents to achieve the desired performance. The paints were then evaluated at ORNL for applicability to metal and concrete surfaces of different sizes and shapes, and soil rubble, Hg-sorption capacity and achieving sag-free coating on surfaces when applied by a manual spray applicator. Ambient cure of the coating resulted in a dry thickness of 16–20 mil. The coatings were removed from smooth or rough surfaces with low to moderate force. Both lab-generated Hg-contaminated samples and soil and brick/rubble samples containing a range of Hg concentrations (collected from the Y-12 site by UCOR) were used in this project. Optimized formulations were applied to these samples and evaluated for ability to suppress Hg vapors, sequester Hg 0 and Hg(II) species and color mapping of mercury species on the test surface. The ORNL team initially identified three to four potential sites at Y-12 and ORNL for potential demonstration. However, the pandemic closed any possible avenues for on-site demonstration of the robotic spray applicator. The metal and concrete samples with low to high Hg content provided by UCOR from an ongoing demolition were subsequently used for ex-situ sample characterization and analysis at ORNL over a period of 25 days. The coatings suppressed Hg vapors with an efficiency of 95–99% from heavily contaminated metal and 60% from concrete substrate. Analysis of total Hg in the peeled paints indicates sorption capacities ranging from 12 µg Hg/g of paint from concrete to 1.2 mg Hg/g of paint from the metal substrates while peeled paints from soil samples could capture a total Hg from 2.3 mg Hg/g to 66 mg/g paint. Toxicity characteristic leaching procedure (TCLP) tests recovered total mercury leached at 0.15±0.0008 µg/L for concrete to 126.81±0.67 µg/L and 377.25±5.67 µg/L for the two metal substrates, to allow for informed decisions on downstream disposal considerations. Together, this Phase IIB demonstrated the potential value of the Trap & See-Hg for Hg uptake, vapor suppression, and color mapping, and may thus be used to enhance worker safety during site cleanup, facility demolition, and downstream waste disposal at Y-12 NSC.

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

MARSAME Radiological Release Report for Diamond Drive Widening and Roundabout Construction Projects

Environmental Protection and Compliance, Environmental Stewardship (EPC-ES) has determined that only a portion of the soil associated with the Diamond Drive Widening/Roundabout Projects (Figure 1) meets the criteria for unrestricted release to the public under Department of Energy (DOE) Order 458.1, Radiation Protection for the Public and the Environment (DOE, 2020). The remaining soil did not meet the criteria for unrestricted release and must be managed appropriately. These conclusions are based on the known history of the area combined with soil sample data collected in 2022; the findings are consistent with DOE Order 458.1 and Los Alamos National Laboratory (LANL) Functional Series Document EPC-ES-FSD-004, Environmental Radiation Protection (LANL, 2021). Sampling and data analysis, as described in this report, were sufficient to meet measurement quality objectives (MQOs) under the Multi-Agency Radiation Survey and Assessment of Materials and Equipment (MARSAME) manual (NUREG, 2009) and LANL procedures (LANL, 2020). Final approvals for waste disposition will come from LANL’s Waste Management Program.

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

Updated Waste Management System Analysis Tool Requirements and Enhancements

Draft presentation for presentation to Nuclear Waste Technical Review Board Winter/Spring Meeting March 2nd NOTE: FYI this presentation is almost the same as the presentation that will be uploaded soon entitled 'Next Generation System Analysis Model: Recently Added Features and Future Plans' for the Waste Management 2022 conference.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Reduced-Order CFD Modeling to Support Waste Loading Optimization in Hanford WTP Vitrification

The U.S. DOE Hanford Site stores over 56 million gallons of radioactive liquid tank waste that must be treated and immobilized for long-term disposal The Waste Treatment and Immobilization Plant (WTP) will vitrify this waste by feeding it into Joule-heated melters, where it is incorporated into a stable borosilicate glass Computational fluid dynamics (CFD) simulations of glass melters can provide insight into the maximum achievable waste loading under varying melter operating conditions Fully resolved VOF multiphase simulations were used as the reference model to capture bubble-driven convection in the melter, including bubble formation, rise behavior, and induced glass melt circulation Effective bubble column diameter and rise velocity were extracted from the resolved simulations, compared with empirical correlations, and refit across relevant viscosity and gas flow rate conditions Explicit gas–liquid interface tracking was replaced with a single-phase momentum source term model, enabling faster steady-state CFD simulations while preserving the dominant hydrodynamic effects of bubbling New empirical correlations were developed for effective bubble column diameter and bubble rise velocity by fitting resolved simulation data across expected melter viscosity and gas flow rate ranges, providing improved inputs for the momentum source term model compared with existing literature correlations The momentum source term model reduced fluid-domain mesh size by 89% and achieved an 8.4× computational speedup relative to resolved bubbling simulations The validated momentum source term approach enables prediction of process-relevant heat transfer behavior in the integrated melter model, including heat transfer from the molten glass to the cold cap, plenum, refractory walls, and surrounding structural regions under varying melter operating conditions

12 - MGMT OF RADIOACTIVE AND NON-RADIOACTIVE WASTE↗
DOE OSTI2026

Tank 11H Low Temperature Aluminum Dissolution and Inhalation Dose Potential Analyses at Savannah River Site – 26018

Currently, there is approximately 34 million gallons of high-level radioactive tank waste in the Tank Farm at the Savannah River Site (SRS). The ultimate goal of operations at the Tank Farm is to remove the high level waste (HLW) from the tanks followed by stabilization of the waste through vitrification of the HLW into glass or grouting the decontaminated waste into saltstone. After bulk removal of the HLW consisting of sludge, saltcake, and supernatant, further efforts are made to reduce the residual waste present in the tank in order to declare preliminary cease waste removal (PCWR) signifying completion of HLW removal. These reduction efforts can include tank washing to remove soluble salts and radioisotopes and dissolution of solids including aluminum. Aluminum in the form of gibbsite and boehmite is relatively insoluble in water. Through addition of aqueous sodium hydroxide, the aluminum can be dissolved at mild temperatures. In order for the waste tank to meet closure mode requirements of the Concentration, Storage, and Transfer Facilities (CSTF), which includes the Tank Farm, Documented Safety Analysis (DSA), a component of the safety basis, the inhalation dose potential (IDP) and the radiolytic hydrogen generation rate of the stored waste must be demonstrated to be lower than their respective designated limits. These parameters are calculated from measured radiochemical analyses of isotopes that emit a high amount of radioactivity including Cs-137, Sr-90, Pu-238, Pu-239, Pu-240, Pu-241, Am-241, and Cm-244. Following the low temperature aluminum dissolution (LTAD) process, Tank 11H slurry samples were pulled from the tank and sent to Savannah River National Laboratory (SRNL) to measure the extent of aluminum dissolution, hydroxide concentration, densities of slurry and supernatant, weight percent solids analyses, and radionuclide activities. The analyses of the composite sample found that approximately 90% of the total aluminum in the slurry was dissolved, indicating successful reduction of the insoluble aluminum in the waste tank. Additionally, the weight percent insoluble solids (slurry basis) measurement of the composite sample was found to be approximately 1%, demonstrating that minimal solids still remain in the tank. Finally, the radiochemical analyses of the composite sample determined that the waste contents of the tank met the IDP and radiolytic hydrogen generation rate requirements of the CSTF DSA. These measurements have shown that the LTAD process in Tank 11H was successful in waste reduction efforts and a positive step towards declaring PCWR and tank closure at SRS.

Dekarske, John [Savannah River National Laboratory↗