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At least 235 records · Page 13

RCT Continuing Training: Radiological Emergency Response

Objectives: 1) List the revised initial response and supplemental actions while responding to a major injury inside of a radiological area; 2) List the initial response and supplemental actions while responding to a minor injury inside of a radiological area; 3) List the initial response and supplemental actions for responding to a Continuous Air Monitoring (CAM) alarm; 4) List the initial response and supplemental actions for responding to a radioactive spill; 5) List the initial response and supplemental actions of responding to an Area Radiation Monitor (ARM) alarm; 6) List the initial response and supplemental actions of an Electronic Personal Dosimeter (EPD) dose and dose rate alarm; 7) List the initial response and supplemental actions of responding to a Stationary Contamination Monitor (SCM) alarm; 8) Identify the methods to perform personnel decontamination.

61 RADIATION PROTECTION AND DOSIMETRY↗

Updated Economic Model for Estimation of GDP Losses in the MACCS Offsite Consequence Analysis Code RDEIM Model Report for MACCS v4.2

This report updates the Regional Disruption Economic Impact Model (RDEIM) GDP-based model described in Bixler et al. (2020) used in the MACCS accident consequence analysis code. MACCS is the U.S. Nuclear Regulatory Commission (NRC) used to perform probabilistic health and economic consequence assessments for atmospheric releases of radionuclides. It is also used by international organizations, both reactor owners and regulators. It is intended and most commonly used for hypothetical accidents that could potentially occur in the future rather than to evaluate past accidents or to provide emergency response during an ongoing accident. It is designed to support probabilistic risk and consequence analyses and is used by the NRC, U.S. nuclear licensees, the Department of Energy, and international vendors, licensees, and regulators. The update of the RDEIM model in version 4.2 expresses the national recovery calculation explicitly, rather than implicitly as in the previous version. The calculation of the total national GDP losses remains unchanged. However, anticipated gains from recovery are now allocated across all the GDP loss types – direct, indirect, and induced – whereas in version 4.1, all recovery gains were accounted for in the indirect loss type. To achieve this, we’ve introduced new methodology to streamline and simplify the calculation of all types of losses and recovery. In addition, RDEIM includes other kinds of losses, including tangible wealth. This includes loss of tangible assets (e.g., depreciation) and accident expenditures (e.g., decontamination). This document describes the updated RDEIM economic model and provides examples of loss and recovery calculation, results analysis, and presentation. Changes to the tangible cost calculation and accident expenditures are described in section 2.2. The updates to the RDEIM input-output (I-O) model are not expected to affect the final benchmark results Bixler et al. (2020), as the RDEIM calculation for the total national GDP losses remains unchanged. The reader is referred to the MACCS revision history for other cost modelling changes since version 4.0 that may affect the benchmark. RDEIM has its roots in a code developed by Sandia National Laboratories for the Department of Homeland Security to estimate short-term losses from natural and manmade accidents, called the Regional Economic Accounting analysis tool (REAcct). This model was adapted and modified for MACCS. It is based on I-O theory, which is widely used in economic modeling. It accounts for direct losses to a disrupted region affected by an accident, indirect losses to the national economy due to disruption of the supply chain, and induced losses from reduced spending by displaced workers. RDEIM differs from REAcct in in its treatment and estimation of indirect loss multipliers, elimination of double-counting associated with inter-industry trade in the affected area, and that it is intended to be used for extended periods that can occur from a major nuclear reactor accident, such as the one that occurred at the Fukushima Daiichi site in Japan. Most input-output models do not account for economic adaptation and recovery, and in this regard RDEIM differs from its parent, REAcct, because it allows for a user-definable national recovery period. Implementation of a recovery period was one of several recommendations made by an independent peer review panel to ensure that RDEIM is state-of-practice. For this and several other reasons, RDEIM differs from REAcct.

29 ENERGY PLANNING, POLICY, AND ECONOMY↗

Characterization of Salt Waste Processing Facility August 2022 Tank 201 sample

Two Salt Waste Processing Facility (SWPF) Tank 201 samples were collected from the replacement of the decontaminated salt solution (DSS) coalescer at different times; one in December of 2021 and the other in August 2022. The December 2021 sample was collected after nitric and water flushing of the DSS Coalescer. The Aug 2022 sample was collected after water flushing only. The December 2021 Tank 201 solids were previously characterized and documented. SWPF requested the characterization of the August 2022 Tank 201 sample, and the comparison of the characterization results for these two Tank 201 solids collected at different dates.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Mo-99 Concentration and processing by Solvent Extraction and Ion Exchange

Normally derived from its transient-equilibrium parent molybdenum-99 ( 99 Mo) (t½ = 66 h), the short-lived isotope technetium-99m ( 99m Tc) (t ½ = 6.0 h) is the most widely used medical isotope in diagnostic imaging today. The National Nuclear Security Administration’s (NNSA’s) Material Management and Minimization (M3) program—established under the auspices of the American Medical Isotope Production Act—has been tasked with facilitating the work of domestic 99 Mo suppliers that do not use highly enriched uranium (HEU). Superconducting electron linear accelerators that employ high-Z converter targets can generate bremsstrahlung photons and neutron fluxes that can induce photonuclear reactions and uranium (U) fission. Argonne, in collaboration with industrial partners, has been developing the process chemistry for superconducting linear accelerator (LINAC)-irradiated triuranium octoxide (U 3 O 8 ) targets to produce 99 Mo. In short, the process involves dissolving the irradiated targets in nitric acid, which simultaneously releases valuable fission products (xenon and iodine). After adjusting the acid concentration, producers process the feed using tri-n-butyl phosphate (TBP), then mix the raffinate derived from this extraction cycle with a phosphinic acid such as Cyanex 272 or di(2-ethylhexyl) phosphoric acid (HDEHP), which is selective for Mo (molybdenum liquid liquid extraction [MoLLE]). In the next step, a concentration column comprising an anion exchange platform is used to decontaminate the remaining fission products, generating a pure 99 Mo stream in sodium hydroxide (NaOH)/sodium chloride (NaCl).

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

Cesium Removal from SY-101 Tank Waste Using Crystalline Silicotitanate

The newly operational Tank Side Cesium Removal (TSCR) system removes radioactive cesium-137 ( 137 Cs) and solids from tank waste supernate in the 200 east area of the Hanford site. Efforts to expand the removal capabilities to the 200 west area are underway by a system anticipated to be called the West Area Risk Management (WARM) system. Laboratory-scale ion exchange processing using expected WARM unit operations were conducted to contribute toward Washington River Protection Solutions (WRPS) establishing accurate process flowsheets for the individual feed campaigns planned for the west area supernate pretreatment. This report describes the small-scale ion exchange testing with 8.0 L of filtered supernate from tank 241-SY-101 (referred to as SY-101) at 16 °C (62 °F) to demonstrate processing conditions that would be prototypic of what the WARM system may experience. One of the waste acceptance criteria (WAC) for the Waste Treatment Plant (WTP) Low-Activity Waste Facility is that the waste must contain less than 3.18×10 -5 Ci 137 Cs per mole of Na. For the SY-101 tank waste to meet this criterion, only 0.44% of the influent 137 Cs concentration may be delivered to the WTP; this requires a Cs decontamination factor of 227. Testing with SY-101 matched current TSCR prototypic operations where a lead-lag configuration is used until the lag column reached the WAC limit. Neither the lead nor lag columns reached the WAC, so a polish column was never utilized during this test. Feed was processed at 1.9 bed volumes (BVs) per hour; the flowrate, in terms of contact time with the crystalline silicotitate (CST) bed, matched the current flowrate at TSCR (an expected flowrate to be used at WARM). The lead column only reached 0.09% Cs breakthrough after processing ~1363 BVs of feed. Cesium breakthrough from the lag column was not observed during the entire processing. Table S.1 and Figure S.1 summarize the observed column performance and relevant Cs loading characteristics. Table S.1 and Figure S.1 also summarize the measured SY-101 Cs load performance.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Review of Mercury Sequestration in Cementitious Waste Forms

Decontaminated salt solution from Tank 50 at the Savannah River Site (SRS) is known to contain dissolved mercury species which are present as hydroxides and oxides. These species are reported in analytical measurements as the organic cation [CH 3 Hg] + and the inorganic cation Hg 2+ . The salt solution is mixed with cementitious reagents, blast-furnace slag (BFS) and thermally beneficiated Class F fly ash in the Saltstone Processing Facility and pumped into engineered Saltstone Disposal Units, (SDUs), to produce a solidified waste form. The solidified waste form, saltstone, encapsulates and chemically stabilizes the mercury and other contaminants, sequestering them from release into the environment. With the discovery of methylmercury in the salt solution, testing was performed to determine the extent to which saltstone prepared with methylmercury (MeHg) retains mercury in the Toxicity Characteristic Leaching Procedure (TCLP). Results from this testing indicated that the amount of mercury extracted remained below the regulatory limit, and that the saltstone formulation with a greater proportion of BFS performed slightly better at retaining Hg in the solid.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Benchmark Analysis for the Optical Dilatometry Method Using Silicon Carbide Temperature Monitors

The main objective of this project was to conduct a benchmark analysis for the optical dilatometry method by using NSUF’s SiC temperature monitors: two (2) SiC temperature monitors provided by NSUF’s BSU-8242 experiment and two (2) SiC temperature monitors provided by NSUF’s GE-Hitachi experiment. Per the BSU-8242 experiment, KGT-3597 and KGT-3591 had a design temperature of 400°C and an exposure of 3 dpa. Per the GE-Hitachi experiment, KGT-3341 and KGT-3336 had a design temperature of 290°C +/- 50°C and an exposure of 0.5–1 dpa. The KGT-3336 SiC monitor was split into two pieces during the decontamination process, making the dilatometry method the only way to analyze both those pieces. The BSU-8242 monitors revealed peak irradiation temperatures under the design temperature, and the GE-Hitachi monitors revealed peak irradiation temperatures within the design temperature range. The optical dilatometry method measured the peak irradiation temperature of BSU-8242 KGT-3597 to be 330°C, while the resistivity method measured the peak irradiation temperature of BSU-8242 KGT-3591 to be 320°C +/- 20°C. The optical dilatometry method measured the peak irradiation temperatures of the pieces of GE-Hitachi KGT-3336 to be 260°C (for the larger piece) and 220°C (for the smaller piece), while the resistivity method measured the peak irradiation temperature of GE-Hitachi KGT-3341 to be 300°C, with an accuracy range of -50°C to +20°C. Both methods of SiC temperature monitor analysis produced very similar peak irradiation temperatures for each pair of SiC passive monitors from the two experiments, BSU-8242 and GE-Hitachi. The results show dilatometry method to be a reliable and less time-intensive process for determining irradiation temperatures from passive SiC thermometry.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

Impacts of Fast Critical Assembly Fuel Discards on Liquid Waste Processes

The Savannah River Nuclear Solutions (SRNS) Fast Critical Assembly (FCA) mission is reestablishing the electrolytic dissolver for processing of Pu and Pu-U materials clad in stainless steel (SS). H-Canyon is planning to dissolve and neutralize FCA fuel without recovering the special nuclear material (i.e., Pu) prior to discarding to the Concentration, Storage, and Transfer Facilities (CSTF) operated by the Savannah River Mission Completion (SRMC) Liquid Waste (LW) Organization. The FCA discards will be combined with sludge in the CSTF after Low Temperature Aluminum Dissolution (LTAD) if needed. The combined waste will be washed, concentrated, and vitrified at the Defense Waste Processing Facility (DWPF). The high level waste canisters produced will be temporarily stored in the Glass Waste Storage Buildings prior to transfer to a future federal repository. Decants from LTAD and sludge washing will be combined with DWPF recycle and dissolved salt cake to prepare salt batches for processing in the Salt Waste Processing Facility. The resulting decontaminated salt solution will be processed in the Saltstone Production Facility.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Integrated Off-Gas System Tests on the DM1200 Melter with RPP-WTP LAW Sub-Envelope Bl Simulants, VSL-03R3851-1, Rev. 0 (Oct 2003)

The report documents melter and off-gas performance results obtained on the DM1200 melter during processing of LAW Sub-Envelope B1 feed. The principal objectives of the DM1200 melter testing were to demonstrate suitable processing and production rates for LAW B1 feed; characterize the glass product for elemental composition; measure elemental emission rates and decontamination factors (DFs) across the melter and off-gas treatment components; and provide additional data on the performance of the thermal catalytic oxidation and selective catalytic reduction (TCO/SCR) unit efficiency in treating organic emissions. The feed was spiked with organic compounds for part of the testing duration. All data were collected for use in engineering and system design purposes. No sampling and analysis to collect environmental data was performed during these tests.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Investigation of Anion-Exchange Method for Purification of Fission-Produced 99 Mo

The LEU-Modified Cintichem (LMC) process is a small-scale (50- to 100-mL) method for purifying fission-produced 99 Mo. Traditionally, the process utilizes custom-made glassware and requires significant manual operations with hot cell manipulators. As an alternative to the LMC process, there is great interest in developing a 99 Mo purification method that could be automated to minimize remote handling while achieving the same purity specifications for the 99 Mo product. This report explores an ion-exchange approach to purifying fission-produced 99 Mo and discusses the decontamination factors for various fission products that are present after initial recovery of Mo on a titania column from a sulfate-based and uranium-containing solution.

07 ISOTOPE AND RADIATION SOURCES↗

Technical Assessment of Off-Gas System Technologies for Potential Use in Molten Salt Reactors

Molten salt reactors (MSRs) are a class of Generation IV advanced reactor technologies aimed to enhance and improve the safety, fuel utilization and cost-effectiveness of nuclear power generation. MSR concepts are based on using a molten salt mixture as a primary nuclear reactor coolant, while the fuel can be either directly dissolved in the coolant (i.e., fluid-fueled) or can be in a separate solid form (i.e., salt-cooled solid-fueled). MSRs can be operated in either thermal or fast neutron spectra, as simple fissile convertors, or breeder reactors (e.g., utilizing both fissile and fertile fuels), as well as accelerator-driven sub-critical reactors or transuranic element burners. Operation of MSRs will necessitate management and treatment of the off-gas streams released to the headspace of the reactor where, in some designs, a cover gas can be circulated to remove certain fission products and maintain an inert atmosphere. An inert MSR cover gas swept over or sparged into the salt will confine radionuclides emerging from the free surface, including radioactive noble and non-noble gases, aerosols, and volatile species. The elements and compounds of these species will have a wide range of thermochemical and physical properties, some of which will decay to different products in the off-gas stream. Therefore, system components used to manage and treat MSR off-gas streams will need to be robust under dynamic reactor and chemistry conditions to support adequate reactor performance by managing the composition and species in the salt coolant and cover gas space. Off-gas system components would perform as a radionuclide boundary to the reactor vessel and a primary means for preventing radionuclide release; therefore, they are of high safety significance for MSR operations. This report reviews various technologies and materials for potential implementation in the design of an MSR off-gas system per a general conceptual framework. The reviewed technologies operate based on scrubbing, capture, delay, or separation of species in the off-gas stream. These include molten hydroxide scrubbers, solid sorbents, delay off-gas systems, particle traps, and cryogenic distillation. Since aspects of an off-gas system will be design-specific, the approach is to provide general discussions on attributes of these technologies and materials, including target off-gas stream species, operational parameters and conditions of importance to decontamination factors, implementation maturity, system component monitoring, potential off-normal conditions and considerations for waste products. The information aims to support the U.S. Nuclear Regulatory Commission in their safety evaluations for MSR designs.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Engineering Assessment And Certification of Integrity Of the DWTF Waste Treatment Tanks At Lawrence Livermore National Laboratory

This Engineering Assessment and Certification of Integrity of the Building 695 (B-695) Decontamination and Waste Treatment Facility (DWTF) Waste Treatment Tanks has been prepared using the guidelines of CCR §66270.16(a) “Specific Part B Information Requirements for Tank Systems”, and the CCR §66264.192(b) requirements for the written assessment of tank systems that store hazardous waste and have secondary containment. The regulations require that this assessment be completed and certified by an independent, qualified, California-registered Professional Engineer. This document will be kept on file by the Lawrence Livermore National Laboratory (LLNL) Environmental Functional Area and is applicable for a maximum of five (5) years from the date of issue.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

U.S. Efforts in Support of Examinations at Fukushima Daiichi - September 2024 Meeting Notes

Information obtained from Fukushima Daiichi Nuclear Power Station (Daiichi) is required to inform future Decontamination and Decommissioning (D&D) activities, improving the ability of the Tokyo Electric Power Company Holdings, Incorporated (TEPCO Holdings) to characterize potential hazards and to ensure the safety of workers involved with cleanup activities. This information also has important implications for the safety and operation of U.S. Commercial nuclear power plants. A collaborative U.S. and Japanese effort was initiated in 2014 by the Department of Energy Office of Nuclear Energy to identify Daiichi examination needs and evaluate recent Daiichi examination data to address these needs. This document summarizes information presented at and findings, action items, and recommendations by U.S. and Japanese experts in reactor safety and plant operations during the September 2024 Forensics Effort meeting. Significant safety insights were obtained in several areas: system and component performance, radionuclide surveys and sampling, debris end-state location, combustible gas effects, and plant operations and maintenance. In addition to reducing uncertainties and knowledge gaps in severe accident modeling progression, these insights continue to be used to assess whether additional updates are needed in guidance for severe accident prevention, mitigation, and emergency planning. Furthermore, Daiichi-related activities, such as code modeling improvements and analysis, testing, and new technology deployment efforts, have the potential to offer additional safety and economic benefits to the operating fleet and new light water reactor (LWR) and non-LWR designs.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Overview on Current Activities of Conduction-Cooled SRF Accelerators and their Applications

When Nb3Sn was reintroduced to the SRF community as an alternative to pure niobium, one key motivation has been to reduce the cryogenic requirements of new and existing accelerators by shifting from 2 K to 4 K operation. Meanwhile, a variety of implementations beyond research machines are being explored. The combination of Nb3Sn with conventional cryocoolers, enabling cryogen-free operation, has paved the way for the development of compact, standalone systems suitable for applications far beyond research, such as enhancing the durability of synthetics via crosslinking or sterilizing food and medical equipment, as well as environmental cleanup when it comes to decontaminating liquid and solid waste material. So, while fundamental R&D continues to refine Nb3Sn resonators, exploring improvements such as replacing the niobium substrate with copper, parallel research efforts are investigating how the increased beam power provided by SRF could expand the commercial use of electron beams. This presentation aims to deliver a comprehensive overview of ongoing research efforts to harness the benefits of SRF through Nb3Sn and conduction cooling.

Vennekate, John [Thomas Jefferson National Acceler↗

Tank 48H Phase 1 Initial Testing Using Sodium Permanganate to Decompose Tetraphenylborate: Simulant Studies with Shaker Oven and 2-L Vessel

Tank 48H currently holds legacy material including organic tetraphenylborate (TPB) compounds from the operation of the In-Tank Precipitation process. The large quantity of TPB is not compatible with the waste treatment facilities at SRS and must be removed or undergo treatment to oxidize the organic compounds before the tank can be returned to routine Tank Farm service. Tank 48H currently holds approximately 270,000 gallons of legacy material comprised of decontaminated salt solution, approximately 20,000 kilograms of TPB solids, 3,400 kilograms of sludge solids, and 1,800 kilograms of monosodium titanate (MST).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Radiological Data Assessment Guidance for Emergency Response

This document provides guidance on how to apply data quality practices to measurements and information collected during the response and recovery to a radiological release. The process of applying data quality practices to measurement data is called data assessment. All data quality practices applied during a radiological incident response must balance the rigorous and time-consuming process normally applied to, for example, site decommissioning and decontamination, with the time- and resource-constraints present during emergency response. The guidance in this document presents the application of data quality practices for data assessment in a graded approach, where the recommended rigor increases as the response continues through its phases and as time and resource constraints relax.

54 ENVIRONMENTAL SCIENCES↗

Evaluation of New Additions to OLI Software in Predicting Mercuric and Mercurous Species in Liquid Waste Operations

Speciation of mercury during the pretreatment steps of tank waste processing is critical to successful mercury removal prior to vitrification during Liquid Waste Operations (LWO) at SRS. OLI software has been used to predict mercury speciation and activity throughout LWO. The OLI software operates based on a thermodynamic framework called the Mixed Solvent Electrolyte (MSE) framework. The MSE framework allows prediction in theoretically infinitely dilute to concentrated mixtures (e.g., purely solute solutions). Before modification to the MSE framework databanks, certain critical mercury species were missing in the MSE databank, and some thermodynamic data needed to be updated for the OLI software to accurately predict mercury chemical species in SRS waste tanks. To better reflect streams across LWO, new mercury species were integrated into the MSE database. To evaluate the changes to the OLI MSE framework per the Technical Task Request (TTR) and the Task Technical and Quality Assurance Plan (TTQAP), waste stream compositions from Tanks 38, 43, and Tank 50 decontaminated salt solution (DSS) were used as model inputs. Models were developed and executed using both the old and new databases. Compositional analyses from caustic Tank 50 DSS and caustic Tanks 38 and 43 were used as the input streams. These streams represent the most comprehensive chemical data sets where both mercury and tank constituents were measured together. Results for Tank 50 DSS predict HgO as the predominant species in both databases. Both methyl and dimethyl Hg species are present when the new database is ‘on’ and are not predicted with the new database turned ‘off’. The new database predicts a greater amount of HgO and a greater fraction of it in the solid phase. Pourbaix diagrams (potential vs. pH) generated for each Tank 50 DSS were identical regardless of which database was used. Elemental Hg and HgO were predicted in the water stable region under basic conditions. Tanks 38 and 43 follow similar trends as the Tank 50 DSS models. Unlike Tanks 38 and 50 DSS, the Tank 43 Pourbaix plot shows a region of stability for an aqueous HgOHCO3 - species between approximately pH 7-11. In all streams, when MeHg+ is included in the inputs, the new database predicts aqueous MeHgOH as the dominant species. If elemental or dimethyl mercury is in the waste stream, the new database model predicts they are unchanged and remain in those states and quantities. Additionally, the total mercury values are reported for both the measured input data and the OLI output data for all considered tanks. The summary indicates that the percentage error between the measured and calculated values is less than 1% in all cases The reconciliations and generation of the Pourbaix diagrams for Tank 50 DSS took approximately ten times longer with the new database ‘on’. In addition, over the course of that time, models with the new database ‘on’ were more likely to crash or display an error. Some modest performance improvements were noted when modeling with an i7 processor versus an i5. An example error is found in Appendix A. Furthermore, Appendix B provides V&V for two chemical systems analyzed with the OLI software, results were satisfactory. It is recommended to utilize the new databases (i.e., HCO.ddb and SR-Hg.ddb) in future Savannah River Mission Completion applications of OLI to represent pseudo steady-state. Furthermore, the integration and utilization of the new databases (i.e., HCO.ddb and SR-Hg.ddb) in modeling applications (e.g., Aspen) is also recommended.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Evaluation of Technologies to Mitigate the Presence of Gaseous Elemental Mercury in Waste Disposal Containers

A study was conducted to evaluate sorbent technologies that can mitigate the presence of elemental mercury (Hg⁰) in waste containers for mercury-contaminated debris (MCD). Decontamination and demolition (D&D) activities at the Y-12 National Security Complex (Y-12) and other U.S. Department of Energy (DOE) Oak Ridge Reservation (ORR) facilities generate MCD requiring offsite disposal. The debris is packaged in appropriate waste containers and may be temporarily stored onsite prior to transport for treatment and/or disposal. During transportation of loads that had no visible liquid Hg at the point of origin, temperature changes can cause Hg⁰ to evaporate, condense, and form droplets on container walls. Furthermore, vibration during transportation could cause beads of Hg to be released from the debris, container walls, and ceiling, resulting in pools of liquid Hg⁰ on the container floor. Waste acceptance criteria (WAC) limitations for commercial disposal facilities, the Nevada National Security Site (NNSS), and ORR mixed low-level waste landfills prohibit the presence of any free liquids in containers identified as a solid waste form. Potential solutions to mitigate the presence of residual liquids that could be formed through vapor condensation include the use of sorbents or similar materials to capture and stabilize volatile Hg⁰ vapors and thus ensure compliance with landfill WAC requirements. This report summarizes data from small-scale laboratory experiments conducted to evaluate sorbent materials for Hg⁰ vapor suppression and sorption of liquid Hg⁰ that could form under relevant transportation and disposal conditions. A series of experiments was conducted to evaluate commercial sorbent materials and their effectiveness for Hg⁰ sorption across a temperature range from 19.4°C to 60°C. The impact of residual moisture on sorption was investigated under relevant conditions, and leaching tests were performed to assess the stability of Hg⁰ captured sorbent materials. The results provide estimates for sorbent quantities needed for a given Hg⁰ mass loading based on experimental results exposing sorbents to gaseous and liquid Hg⁰ at various mass ratios. Overall, sorbents that were most effective for Hg⁰ vapor suppression were brominated activated carbons, mackinawite-based sorbents coated on vermiculite, and sulfur-modified granular activated carbon. Elevated temperatures and moisture conditions did not result in significant increases of Hg⁰ headspace concentrations, and the materials also demonstrated high sorption capacities for the sorption of liquid Hg⁰.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗