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DOE OSTI2022

Reduction of Pertechnetate by Chemical and Photochemical Approaches and Incorporation of Tc(IV) into Titanium Dioxide

Technetium-99 is a prevalent fission product from nuclear waste. The long half-life (211,000 yr) and environmental mobility of pertechnetate (TcO 4 - ) render Tc particularly challenging to isolate and stabilize. Here, in this study, we present two approaches for development of potential wasteforms using titanium dioxide, TiO 2 . Approach 1 is a low temperature chemical synthesis of TiO 2 doped with Tc(IV) from TcO 4 - intended to mimic the Tc waste stream from the UREX family of separations and removes 98.5 % of the Tc, mainly present as edge-shared Tc(IV) pairs. Approach 2 utilizes TiO 2 to photocatalytically reduce TcO 4 - to Tc(IV) stabilized on the surface of or within the TiO 2 lattice. The %Tc removed from solution and adsorbed to TiO 2 is pH dependent, with the maximum Tc(IV) adsorbed at pH 3–4 as either TcO 2 or edge-sharing Tc(IV) octahedra. The Tc(IV)-TiO 2 composites materials formed by both approaches are suitable for consolidation into a dense wasteform by Hot Isostatic Pressing (HIPing).

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2024

The simultaneous removal of technetium and iodine from Hanford tank waste

The simultaneous removal of radionuclides technetium-99 and iodine-129 from an actual decontaminated Hanford tank waste sample (a mixture of decontaminated waste from tanks 241-AP-105 and 241-AP-107) was demonstrated for the first time in this work. A series of commercially available ion exchange resins were evaluated in batch contact tests in the tank waste, and all showed removal of both Tc and I. The highest Tc removal was observed for Purolite A530e while the highest iodine removal was observed for ResinTech SIR-110-MP. Batch tests in simulated tank waste with these two resins showed that the SIR-110-HP-MP had consistently higher K d for both pertechnetate and iodide and much higher K d than previous works on Tc removal from Hanford waste. As such, the SIR-110-MP was evaluated in a dual -column (lead/lag) test processing 5.2L of the tank waste mixture showing 60% breakthrough of Tc on the lead column and no significant breakthrough on the lag after 625 bed volumes (BV, 6 mL size) while significant iodine breakthrough (>50%) occurred after 28 BV. The limited iodine uptake was attributed to the column conditions generating mass transfer limitations. A fraction of the Tc and I was not captured by the resin (<10%) in either the batch tests or column tests. The iodine fraction was identified to be an iodide, likely organo-iodide. The fraction of the Tc was identified as a non-pertechnetate species, which is the first time non-pertechnetate has been identified in AP-105 and AP-107 tanks, although the exact species is still unknown.

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

Metallic Technetium Sequestration in Nickel Core/Shell Microstructure during Fe(OH)2 Transformation with Ni doping

This study investigates the impacts of Ni doping on technetium-99 (Tc) sequestration in aqueous solutions through transformation of Fe(OH)2(s) to iron spinel (magnetite) under alkaline conditions. Extensive solid characterization was performed for the mineral phases produced, as well as the Tc/Ni speciation and distribution within these phases. X-ray diffraction results show that iron spinel was the dominant mineral product without detectable Ni incorporation. The doped Ni ions mainly precipitated as fine Fe/Ni oxide/hydroxide particles, including strongly reduced nanometer?sized spheroidal Ni-rich and metallic Ni phases. High-resolution analytical scanning transmission electron microscopy using energy dispersive X-ray spectroscopy and electron energy loss spectroscopy on the produced solid samples (focused ion beam-prepared specimens) revealed three Tc distribution domains dominated by nanocrystals and, especially, a Tc-rich metallic phase. Instances of metallic Tc were specifically found in spheroidal, Ni-rich and metallic nanoparticles exhibiting a core/shell microstructure that suggests strong reduction and sequential precipitation of Ni-Tc-Ni. Mass balance analysis showed nearly 100% Tc removal from the 4.8 × 10-4 M Tc solutions. The finding of the metallic Tc encapsulation indicates that Tc sequestration through Ni-doped Fe(OH)2(s)?to?iron spinel transformation process likely provides an alternative treatment pathway for Tc removal and could be combined into further waste treatment approaches.

Wang, Guohui↗
DOE OSTI2022

Reductive removal of pertechnetate and chromate by zero valent iron under variable ionic strength conditions

Radioactive technetium-99 (Tc) present in waste streams and subsurface plumes at legacy nuclear reprocessing sites worldwide poses potential risks to human health and the environment. This research comparatively evaluated efficiency of zero valent iron (ZVI) toward reductive removal of Tc(VII) in presence of Cr(VI) from NaCl and Na 2 SO 4 electrolyte solutions under aerobic conditions. In both electrolytes, anticorrosive Cr(VI) suppressed oxidation of ZVI at elevated concentrations resulting in the delay of initiation of Tc(VII) reduction to Tc(IV)). In the absence of Cr(VI), no such delay was observed in the analogous systems. At low ionic strength, retarded ZVI oxidation inhibited Tc(VII) reduction in part due to the competing thermodynamically more favorable reduction of Cr(VI) to Cr(III); this effect was particularly pronounced in the chloride medium at the ionic strength ? 0.1. Higher ionic strength favored reduction of both Tc(VII) and Cr(VI), which followed a second-order reaction rate in both electrolytes attributed to the more efficient iron oxidation as evident from x-ray diffraction and electron microscopy measurements. Magnetite was the primary iron oxide phase, and its higher fraction in the SO 4 2- solutions facilitated reductive removal of Tc(VII) and Cr(VI). In the Cl- matrix, Cr(VI) promoted further oxidation of magnetite as well as formation of chromite diminishing overall reductive capacity of this system and resulting in less effective removal of Tc(VII) compared to the SO 4 2- solutions.

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

Accumulation mechanisms for contaminants on weak-base hybrid ion exchange resins

Mechanism of hexavalent chromium removal (Cr(VI) as CrO 4 2- ) by the weak-base ion exchange (IX) resin ResinTech® SIR-700-HP (SIR-700) from simulated groundwater is assessed in the presence of radioactive contaminants iodine-129 (as IO 3 - ), uranium (U as uranyl UO 2 2+ ), and technetium-99 (as TcO 4 - ), and common environmental anions sulfate (SO 4 2- ) and chloride (Cl-). Batch tests using the acid sulfate form of SIR-700 demonstrated Cr(VI) and U(VI) removal exceeded 97%, except in the presence of high SO 4 2- concentrations (536 mg/L) where Cr(VI) and U(VI) removal decreased to ≥ 80%. However, Cr(VI) removal notably improved with co-mingled U(VI) that complexes with SO 4 2- at the protonated amine sites. These U–SO 4 2- complexes are integral to U(VI) removal, as confirmed by the decrease in U(VI) removal (<40%) when the acid chloride form of SIR-700 was used instead. Solid phase characterization revealed that CrO 4 2- is removed by IX with SO 4 2- complexes and/or reduced to amorphous Cr(III)(OH) 3 at secondary alcohol sites. Tc(VII)O 4 - and I(V)O 3 - also undergo chemical reduction, following a similar removal mechanism. Oxyanion removal preference is determined by the anion reduction potential (CrO 4 2 - >TcO 4 - >IO 3 - ), geometry, and charge density. For these reasons, 39% and 69% of TcO 4 - and 17% and 39% of IO 3 - are removed in the presence and absence of Cr(VI), respectively.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2023

Development of an accelerator-based neutron source to prototype Mo-99 production, part I: A liquid LBE windowless target

In this article, Molybdenum-99 (Mo-99)’s decay product, technetium-99 (Tc-99 m), is one of the most critical isotopes for medical diagnostics. To provide U.S. domestic supply of Mo-99 without using high-enriched uranium (HEU), a subcritical uranium target assembly (UTA) is irradiated by an accelerator-based neutron source to create Mo-99 through fission. This study discusses the development of the accelerator-based neutron source. The high-energy electrons from the accelerator irradiate a liquid lead-bismuth eutectic (LBE) target to produce neutrons. Part I of this work focuses on numerical and experimental analysis towards the development of a liquid LBE windowless target. Unlike the existing windowless targets in literature, the current design creates a vertical free surface for a beam to irradiate. First, a hydrodynamic analysis of the LBE windowless target is performed. Simplified analytical calculations are assisted by 2D computational fluid dynamics (CFD) simulations to design the target, with the focus on eliminating recirculation zones and avoiding cavitation. With the optimized geometry, the experimental study is performed to investigate the flow hydrodynamics using liquid LBE. The experiments (1) compare pressure drop in the system to correlation predictions; (2) visualize the free surface liquid LBE flow from the beam view; (3) validate the LBE flow profile using temperature sensitive paint from the side view; and (4) validate the liquid LBE film thickness using gamma densitometer measurements. Second, the power handling capability of the designed windowless target is investigated. The divider plate in the current design is susceptible to overheating due to the thin LBE film in front. As LBE erosion and corrosion is likely to occur at an LBE velocity of 2.0 m/s and temperature above 500 °C, a power limit of 10 kW of beam power was established to prevent this corrosion from occurring, which is calculated by a Nusselt number correlation. The divider plate surface temperature at 10 kW agrees well with the 3D CFD simulation results. Part I demonstrates the fundamental physics in liquid LBE windowless target design and associated testing. A companion paper, Part II will demonstrate how to couple this windowless target into the Mo-99 production system, including an accelerator system operating under an ultra-high vacuum and the UTA cooled by water at room temperature.

43 PARTICLE ACCELERATORS↗
DOE OSTI2023

Elucidating Actinide–Pertechnetate and Actinide–Perrhenate Bonding via a Family of Th–TcO 4 and Th–ReO 4 Frameworks and Solutions

Technetium-99, a β-emitter produced from 235 U fission, poses a challenge for the nuclear industry due to co-extraction of pertechnetate (TcO 4 – ) with the actinides (An) during nuclear fuel reprocessing. Previous studies suggested that direct coordination of pertechnetate with An plays an important role in the coextraction process. However, few studies have provided direct evidence for An–TcO 4 – bonding in the solid state, and even fewer in solution. The present study describes synthesis and structural elucidation of a family of thorium(IV)-pertechnetate/perrhenate (ReO 4 – , nonradioactive surrogate) compounds, which is obtained by dissolution of thorium oxyhydroxide in perrhenic/pertechnic acid followed by crystallization, with or without heating. For reaction ratios of 3:1, 4:1, and 6:1 MO 4 – /Th(IV) (M = Tc, Re), the crystallized compounds reflect the same ratio, suggesting facile and flexible coordination. Furthermore, nine structures reveal 1-dimensional and 2-dimensional frameworks with varying topologies. While a multitude of compounds isolated from 4:1 (and 6:1) reaction solutions feature Th monomers linked by MO 4 – , the 3:1 reaction solution yielded the well-known dihydroxide-bridged thorium dimer, linked, and capped by MO 4 – . Density functional theory calculations on ReO 4 – /TcO 4 – isomorphs suggest similar bonding characteristics in the solid state, but experimental solution characterization noted differences. Specifically, small-angle X-ray scattering studies suggest the bonding of Th–TcO 4 – persists in solution, while Th–ReO 4 – bonding is less apparent.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2024

Advanced Porous Materials as Designer Platforms for Sequestering Radionuclide Pertechnetate

Technetium-99 ( 99 Tc), predominantly present as pertechnetate ( 99 TcO 4 – ), is a challenging contaminant in nuclear waste from artificial nuclear fission. The selective removal of 99 TcO 4 – from nuclear waste and contaminated groundwater is complex due to (i) the acidic and intricate nature of high-level liquid wastes; (ii) the highly alkaline environment in low-activity level tank wastes, such as those at Hanford, and in high-level wastes at locations like Savannah River; and (iii) the potential for 99 TcO 4 – to leak into groundwater, risking severe water pollution due to its high mobility. This Review focuses on recent developments in advanced porous materials, including metal–organic frameworks (MOFs), covalent organic frameworks (COFs), and their amorphous counterparts, porous organic polymers (POPs). These materials have demonstrated exceptional effectiveness in adsorbing 99 TcO 4 – and similar oxyanions. We comprehensively review the adsorption mechanisms of these anions with the adsorbents, employing macroscopic batch/column experiments, microscopic spectroscopic analyses, and theoretical calculations. In conclusion, we present our perspectives on potential future research directions, aiming to overcome current challenges and explore new opportunities in this area. Our goal is to encourage further research into the development of advanced porous materials for efficient 99 TcO 4 – management.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2024

Soil Desiccation Treatability Testing at BC Waste Disposal Cribs

During Hanford’s production period, low-level waste products generated from chemical processing of uranium fuel rods were discharged directly to the ground through a system of cribs and trenches located in the 200-BC-1 Operable Unit (OU). The site consists of 6 cribs and 20 trenches that received more than 117,000 m3 of radioactive liquid waste discharged to the soil. These unlined infiltration galleries held volumes of liquid waste while it seeped into the ground, with the understanding that the 100 m (330 ft) thick vadose zone in the area would effectively capture the effluent and prevent groundwater impacts. A conceptual model showing the operation of cribs and trenches is shown in Figure 1. Data show effluent from the 26 cribs and trenches containing about 410 curies of Technetium-99 (Tc-99) is primarily located between 30 m and 70 m (98 ft and 230 ft) depth (Corbin et al., 2005; Ward et al., 2004). Despite no evidence indicating that the contamination has reached the groundwater at BC cribs and trenches, the mobility of Tc-99 had been demonstrated in laboratory tests and was recognized as a threat to groundwater at the site. Using data from numerical models, laboratory analyses, field investigations, and information on historical discharges, the EPA and Ecology identified Tc-99 and U contamination of the vadose zone as a remediation priority. The U.S. DOE was notified by EPA and Ecology regarding risks associated with Tc-99 contamination in a letter requesting development of a strategy for improved methods to understand the nature and extent of vadose zone contamination, specifically Tc-99, and to develop remedial options for addressing such contamination. To develop the appropriate technology for characterizing, remediating, and monitoring the deep vadose zone Tc-99 contamination, the U.S. DOE worked with the EPA and Ecology to create a Treatability Test Plan under a Remedial Investigation/Feasibility Study (RI/FS) for the Hanford 200 Areas. Under this RI/FS, it was determined that a treatability test for soil desiccation should be carried out as it was identified as a promising in-situ treatment technology for mitigating risks posed by Tc-99 contamination to the groundwater table. The BC Cribs and Trenches site was identified as a representative site for Tc-99 and U contamination and selected for the soil desiccation treatability test. In this chapter, we summarize the overlying regulatory framework of RI/FS and treatability tests and illustrate how development and experimentation supported the evaluation of selected remedies. We briefly discuss the RI/FS for the 200 Areas of the Hanford Site and focus on the soil desiccation treatability testing performed at the BC cribs and trenches site under the Deep Vadose Zone Treatability Test Plan for the Hanford Central Plateau (DVZ-TT). The DVZ-TT is one component of the remedial investigation/feasibility study for the Hanford 200 Areas and represents the underlying regulatory framework that drives site operations towards records of decision and site closure.

Mangel, Adam R.↗
DOE OSTI2021

Annual Status Report (FY 2020): Performance Assessment for the Disposal of Low Level Waste in the 200 East Area Burial Grounds

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 requirements of DOE O 435.1 Chg 1 and performance objectives defined in companion documents DOE M 435.1-1 Chg 1and 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 Chg 1 is expected. A new PA study was initiated in fiscal year (FY) 2019 for evaluation of active disposal sites within the 200 East and 200 West Areas (Trench 94 in 200 East; Trenches 31 and 34 in 200 West) due to extended time elapsing between the current annual status report and the original PA for the active disposal sites. The new PA for the active disposal sites is expected to be completed in FY 2021. Within the active burial grounds in the 200 East Area, low-level waste and mixed low-level waste will continue to be disposed of in the dedicated U.S. Navy reactor compartment trench at the 218-E-12B Burial Ground (Trench 94). During this reporting period (FY 2020, from October 1, 2019, through September 30, 2020), two reactor compartments were disposed in Trench 94. Results from sorption experiments are summarized for this reporting period to quantify the efficacy of concrete waste forms in retaining key radionuclides (e.g., technetium-99 and iodine-129). The test durations ranged from 1 to 3 months. Continued groundwater monitoring of the 200 East Area LLBGs indicates no groundwater contamination due to LLBG waste. Current assumptions about future land use at the Hanford Site are consistent with PA analysis1 assumptions of a post-closure facility that will not be degraded by human activity. The LLBGs are located in an area identified for waste management and containment of residual contamination. This area will remain after final environmental remediation and the proposed shrinkage of Hanford Site boundaries to small sections within the 200 East and 200 West Areas in the Central Plateau (DOE/EIS-0391). The current closure plan for the LLBGs (DOE/RL-2000-707) estimates that the 200 East LLBGs will be closed in the 2050 timeframe. The Disposal Authorization Statement, other technical basis documents, and the radioactive waste management basis are of continued adequacy to meet the performance objectives of DOE O 435.1 Chg 1. Overall, there are no substantive changes to primary PA assumptions nor the PA analysis conclusion; therefore, compliance with DOE O 435.1 Chg 1 and the Disposal Authorization Statement is maintained.

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

Vitrification of Hanford Tank Wastes for Condensate Recycle and Feed Composition Changeover Testing

During the vitrification of Hanford Site nuclear waste at the Waste Treatment and Immobilization Plant (WTP), which is a part of the safe and efficient retrieval, treatment, and disposal mission of the U.S. Department of Energy Office of River Protection, the offgas condensate generated from the waste-to-glass conversion is currently planned to be concentrated by evaporation in the Effluent Management Facility (EMF). This concentrated condensate can then be recycled back to the incoming waste and vitrified. To test the recycle process, an apparatus was designed and built to mimic the EMF evaporator and was then used to concentrate a volume of condensate produced during the vitrification of a sample of Hanford tank 241-AP-107 (referred to herein as AP-107) waste in a continuous laboratory-scale melter (CLSM). The concentrated condensate was added to an additional sample of AP-107 waste, to mimic one round of the recycle process, and the combined solution was vitrified, producing a second round of recycle condensate. In the current study, the EMF test apparatus was used to concentrate the second-round recycle condensate under evaporation conditions (at 45 °C and 1.4 psia) designed to emulate EMF operation. The condensate was successfully concentrated by a factor of ~10 while retaining over 95 % of the technetium-99 ( 99 Tc), Cs, and I inventories in the concentrate. Another portion of AP-107 waste was retrieved by Washington River Protection Solutions, LLC (WRPS) and transferred to Pacific Northwest National Laboratory (PNNL), where it was pretreated and then combined with the second-round recycle AP-107 condensate concentrate and glass-forming chemicals (GFCs) to form the two-time recycle AP-107 melter feed, approximating a second round to the recycling action to be performed at the WTP. A portion of AP-105 waste was also retrieved by WRPS and provided to PNNL for pretreatment and combining with GFCs to form AP-105 melter feed. The two-time recycle AP-107 and AP-105 melter feeds were processed consecutively in the CLSM. The CLSM run proceeded for 13.63 hours, producing 9.70 kg of glass for an average glass production rate of 1464 kg m 2 d -1 during the two-time recycle AP-107 feed charging and 1568 kg m 2 d -1 during the AP-105 feed charging. The rate during AP-107 charging was essentially equivalent to the rate when processing no-recycle AP-107 feed and lower than that achieved when processing one-time recycle AP-107 feed. However, all rates were within the potential range of variability when processing melter feeds with similar composition in the CLSM. Likewise, the rate during AP-105 charging was higher than the previous rate processing AP-105, but within the potential CLSM range. The cold-cap characteristics changed from the typically thin AP-107 cold cap to a foamy-edged cold cap as previously seen with AP-105 shortly after transitioning to the AP-105 melter feed. The glass produced during the CLSM run was within 10 % of its target composition for the primary glass components. The CaO and Li 2 O targets varied by more than 1 wt% between the two-time recycle AP-107 and AP-105 glass targets and it took about 2 turnovers of the CLSM glass inventory to reach a relative chemical steady state in the glass for CaO and Li 2 O after the melter feed inputs were switched. The 99 Tc and total cesium content in the melter feeds were maintained at concentrations expected to be experienced at the WTP. During the CLSM run, while processing the two-time recycle AP-107 melter feed at a relative chemical steady state, the 99 Tc/Cs ratio was 10, and 34% of 99 Tc and 74% of Cs were retained in the glass. These values were higher than those measured in the CLSM run with one-time recycle AP-107 melter feed. After the transition to processing the AP-105 melter feed, when the production reached a relative chemical steady state, the 99 Tc/Cs ratio was 77 while 44% of 99 Tc was retained in the glass. The C's retention during this time frame reached 200% due to the excess C's in the glass after the target content decreased to 15% of its initial level in the two-time recycle AP-107 melter feed to the lower target in the Ap-105 melter feed. While iodine was below inductively coupled plasma mass spectrometry analytical reporting limits in the melter feed and glass samples, it was detected in quantities above the analytical reporting limits in the liquid and filter samples collected from the CLSM offgas treatment system. The behavior of iodine in the CLSM offgas treatment system followed a similar pattern to those of 99 Tc and Cs.

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

Hanford Site Composite Analysis: LLBG-200-W B Vadose Zone Model

The objectives of the vadose modeling for the updated Hanford Site Composite Analysis (CA) are to simulate the flow and transport of water and radionuclide releases from the surface to the water table and to provide radionuclide transfer rates to the CA saturated zone model (CP-57037, Model Package Report: Plateau to River Groundwater Model, Version 8.3). Water additions include natural recharge and water discharged to the ground as a result of industrial processes associated with Hanford Site operations. Contaminant sources include radionuclides in water discharged to the ground during operations and radionuclides disposed “dry” in solid waste burial grounds or other means. The following 16 radionuclides were selected for this modeling effort: carbon-14 (C-14), chlorine-36 (Cl-36), tritium (H-3), iodine-129 (I-129), neptunium-237 (Np-237), rhenium-187 (Re-187), strontium-90 (Sr-90), technetium-99 (Tc-99), uranium-232 (U-232), uranium-233 (U-233), uranium-234 (U-234), uranium-235 (U-235), uranium-236 (U-236), uranium-238 (U-238), radium-226 (Ra-226), and thorium-230 (Th-230). The simulation time starts in 1943 and ends at 12070, which is 10,000 years after assumed Hanford Site closure in 2070. The parallel version of the Subsurface Transport Over Multiple Phases (STOMP 1 ) simulator, officially named the exascale Subsurface Transport Over Multiple Phases (eSTOMP), is used to simulate flow and transport for the vadose models. The documentation for the STOMP code is comprehensive. The theoretical and numerical approaches applied in the STOMP code are documented in a published theory guide (PNNL-12030, STOMP Subsurface Transport Over Multiple Phases Version 2.0 Theory Guide). The code has undergone a rigorous verification procedure against analytical solutions, laboratory-scale experiments, and field-scale demonstrations. The application guide (PNNL-11216, STOMP Subsurface Transport Over Multiple Phases Application Guide) provides instructive examples in the application of the code to classical groundwater problems. The user’s guide (PNNL-15782, STOMP: Subsurface Transport Over Multiple Phases Version 4.0: User’s Guide) describes the general use, input file formatting, compilation, and execution of the code. The primary output of the vadose zone modeling is radionuclide transfer rates to the groundwater for input into the saturated zone model. The rates will be summed over the 100 by 100 m saturated zone model grid cells that fall within the vadose zone model source domain.

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

Hanford Site Composite Analysis: Comparison of One- and Three-Dimensional Simulation of Vadose Zone Transport and Transfer to Groundwater for Selected Waste Sites

The purpose of this document is to compare 1D and 3D technetium-99 (Tc-99) VZ transport and transfer rates to groundwater for five selected waste sites (two tanks, two cribs, and one trench) to demonstrate the need to consider 3D flow and transport mechanisms and commingling of contaminant plumes originating from adjacent waste sites. The waste sites selected for the comparison evaluation are the 241-A-105 Tank in the A Farms Area model, the 216-B-18 Crib and the 216-B-25 Trench in the BC Cribs and Trenches model, the 216-B-49 Crib in the B Complex model, and the 241-SX-115 Tank in the S Farms Area model. The selected waste sites all represent different disposal or leak conditions, with and without potential impacts due to release-fluid commingling, occurring at the Hanford Site Central Plateau.

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

Predictive Contaminant Transport Simulation with the P2R Model for the Composite Analysis Inventory Sensitivity Case

The Plateau to River (P2R) Model is a groundwater flow and contaminant fate and transport (F&T) simulation model used to support remedial activities conducted by the Central Plateau Cleanup Company (CPCCo) at the Hanford Site in Washington State. The model simulates contaminants of concern within the saturated zone of the uppermost aquifer beneath the Central Plateau and downgradient to the Columbia River. The overall objective of the saturated zone modeling effort is to provide a basis for making informed remedial action decisions based on descriptions of current and expected future contaminant concentrations in groundwater at decision points within and downgradient of the Central Plateau of the Hanford Site. Specifically, the purpose of this environmental calculation file (ECF) is to describe an inventory sensitivity case of the CA base case. The inventory sensitivity case implements a change in the activity contribution from the vadose zone in the A Trenches Area model (for tritium (H-3) and iodine-129 (I-129)), BC Cribs and Trenches model (for I-129 and technetium-99 (Tc-99)), and the PUREX Area model (for H-3 and I-129). All other simulated inventories are identical to the CA base case. The simulation of F&T of contaminants reported in this case will support dose predictions as part of the Hanford Site CA.

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

Predictive Contaminant Transport Simulation with the P2R Model for the Composite Analysis Recharge Sensitivity Case

The Plateau-to-River (P2R) model is a groundwater flow and contaminant fate and transport (F&T) simulation model used to support remedial activities conducted by the Central Plateau Cleanup Company (CPCCo) at the Hanford Site in south-central Washington State. The overall objective of the saturated zone modeling effort is to provide a basis for making informed remedial action decisions based on descriptions of current and expected future contaminant concentrations in groundwater at decision points within and downgradient of the Central Plateau of the Hanford Site. Specifically, the purpose of this environmental calculation file (ECF) is to describe a recharge sensitivity case of the CA base case. The recharge sensitivity case implements a change in the activity contribution from the vadose zone in the A Trenches Area model (for tritium [H-3] and iodine-129 [I-129]), BC Cribs and Trenches model (for I-129 and technetium-99 [Tc-99]), or the Plutonium Uranium Extraction (PUREX) Area model (for H-3 and I-129). All other simulated inventories are identical to the CA base case. The simulation of fate and transport of contaminants reported in this case will support dose predictions as part of the updated Hanford Site CA.

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

Vadose Zone Model for B-63 Area for Composite Analysis (Rev. 1)

The objectives of the vadose modeling for the updated Hanford Site composite analysis (CA) are to simulate the flow and transport of water and radionuclide releases from the surface to the water table and to provide radionuclide transfer rates for the plateau to river (P2R) model, version 8.3 (CP-57037, Model Package Report: Plateau to River Groundwater Model, Version 8.3). Water additions include natural recharge and water discharged to the ground as a result of industrial processes associated with Hanford Site operations. Contaminant sources include radionuclides in water discharged to the ground during operations and radionuclides disposed “dry” in solid waste burial grounds or other means. The following 16 radionuclides were selected for this modeling effort; carbon-14 (C-14), chlorine-36 (Cl-36), tritium (H-3), iodine-129 (I-129), neptunium-237 (Np-237), rhenium-187 (Re-187), strontium-90 (Sr-90), technetium-99 (Tc-99), uranium-232 (U-232), uranium-233 (U-233), uranium-234 (U-234), uranium-235 (U-235), uranium-236 (U-236), uranium-238 (U-238), radium-226 (Ra-226), and thorium-230 (Th-230). The simulation time starts in 1943 and ends at 12070, which is 10,000 years after assumed Hanford Site closure in 2070.

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

Vadose Zone Model for LLBG-200W A for Composite Analysis

The objectives of the vadose modeling for the updated Hanford Site composite analysis (CA) are to simulate the flow and transport of water and radionuclide releases from the surface to the water table and to provide radionuclide transfer rates for the plateau to river (P2R) model, version 8.3 (CP-57037, Model Package Report: Plateau to River Groundwater Model, Version 8.3). Water additions include natural recharge and water discharged to the ground as a result of industrial processes associated with Hanford Site operations. Contaminant sources include radionuclides in water discharged to the ground during operations and radionuclides disposed “dry” in solid waste burial grounds or other means. The following 16 radionuclides were selected for this modeling effort: carbon-14 (C-14), chlorine-36 (Cl-36), tritium (H-3), iodine-129 (I-129), neptunium-237 (Np-237), rhenium-187 (Re-187), strontium-90 (Sr-90), technetium-99 (Tc-99), uranium-232 (U-232), uranium-233 (U-233), uranium-234 (U-234), uranium-235 (U-235), uranium-236 (U-236), uranium-238 (U-238), radium-226 (Ra-226), and thorium-230 (Th-230). The simulation time starts in 1943 and ends at 12070, which is 10,000 years after assumed Hanford Site closure in 2070.

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

Vadose Zone Model for LLBG-200W A for Composite Analysis

The objectives of the vadose modeling for the updated Hanford Site composite analysis (CA) are to simulate the flow and transport of water and radionuclide releases from the surface to the water table and to provide radionuclide transfer rates for the plateau to river (P2R) model, version 8.3 (CP-57037, Model Package Report: Plateau to River Groundwater Model, Version 8.3). Water additions include natural recharge and water discharged to the ground as a results of industrial processes associated with Hanford Site operations. Contaminant sources include radionuclides in water discharged to the ground during operations and radionuclides disposed "dry" in solid waste burial grounds or other means. The following 16 radionuclides were selected for this modeling effort: carbon-14 (C-14), chlorine-36 (Cl-36), tritium (H-3), iodine-129 (I-129), neptunium-237 (Np-237), rhenium-187 (Re-187), strontium-90 (Sr-90), technetium-99 (tc-99), uranium-232 (U-232), uranium-233 (U-233), uranium-234 (U-234), uranium-235 (U-235), uranium-236 (U-236), uranium-238 (U-238), radium-226 (Ra-226), and thorium-230 (Th-230). The simulation time starts in 1943 and ends at 12070, which is 10,000 years after assumed Hanford Site closure in 2070.

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