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At least 127 records · Page 7

Off-design operation and performance of pumped thermal energy storage

In this article, we describe off-design models and control strategies for a Pumped Thermal Energy Storage (PTES) system that uses liquid thermal energy storage: specifically molten salt for hot storage and methanol for cold storage. Off-design conditions arise when load-following, or due to variations in storage tank temperatures or ambient temperatures. We propose a control strategy that uses inventory control to manage the mass flow rate in the thermodynamic cycles, which facilitates load following. We also propose a control strategy for the storage fluid mass flow rates, which are varied to ensure the molten salt is maintained at its design temperature. This maximizes efficiency and minimizes problems with salt freezing or degradation. The cold storage fluid mass flow rate is varied so that the cold tanks have the same state-of-charge as the hot tanks. This leads to variations in cold fluid temperature, but these variations are shown to be acceptably small (e.g. 7.5% increase), and this control method is shown to be simpler and more efficient than an alternative strategy where tanks become unbalanced. The ambient temperature and storage tank temperatures are moved ±50 °C from the design values and the impact on power, duration, and tank temperatures is quantified. Results demonstrate that the proposed control strategy is stable and self-correcting - that is, storage temperatures converge on stable values after two-to-three charge-discharge cycles. When inputs return to design values, the system returns to its design point after two charge-discharge cycles. We also demonstrate that inventory control enables delivery of the target power output even when off-design conditions exist that would normally reduce the power output.

25 ENERGY STORAGE↗

Failure Analysis–Informed Risk Assessment Framework for Geological Carbon Storage Using Numerical Simulation and Machine Learning

Geological carbon storage (GCS) is recognized as a critical technology for achieving large-scale reductions in anthropogenic carbon dioxide (CO 2 ) emissions. Ensuring long-term containment and safety requires robust risk assessment frameworks that account for geological uncertainty and identify potential failure scenarios. Among various indicators, the area of review (AoR) serves as a key metric for evaluating storage performance, regulatory compliance, and monitoring design, as it delineates the spatial extent impacted by pressure buildup and plume migration. However, conventional AoR-based risk assessments typically perturb parameters within narrow uncertainty bounds, potentially overlooking rare but high-impact events arising from extreme geological conditions. In this study, we present a failure analysis–informed risk assessment framework for large-scale GCS projects to improve site prescreening and monitoring design. A suite of 300 numerical simulations was generated using stochastic geological models that vary five key parameters: net-to-gross ratio, anisotropy azimuth, porosity multiplier, permeability multiplier, and vertical-to-horizontal permeability ratio. Among these, 200 realizations represent normal geological uncertainty, while 100 additional cases explore extreme yet plausible conditions for failure-case analysis. The AoR was simulated and computed from pressure and CO 2 saturation fields, where the baseline AoR boundary, representing the extent predicted under typical geological uncertainty, was defined as the union of 200 normal-range simulations, and failure was identified when extreme-range cases exceeded this baseline. Results show that incorporating broader parameter uncertainty produces significantly larger AoR extents, underscoring the potential underestimation of risk under conventional uncertainty ranges. Furthermore, spatial probability maps derived from failure-induced AoR exceedance identify regions requiring enhanced monitoring attention. Various machine learning (ML)–based classifiers were developed to predict failure occurrence from geological parameters, with the random forest model achieving the highest performance (F1-score of 0.986). Consistent findings from correlation coefficient, feature importance, and Sobol sensitivity analyses reveal that low net-to-gross ratios and permeability multipliers are the dominant risk drivers, reflecting reduced reservoir connectivity and limited pressure dissipation. Altogether, these results provide a novel framework for risk-informed site prescreening and monitoring design that explicitly considers rare but high-impact geological scenarios in GCS projects.

25 ENERGY STORAGE↗

University of Missouri Research Reactor (MURR) LEU Fuel Fluid-Structure Interaction Analysis

The University of Missouri Research Reactor (MURR ® ) is one of five U.S. high performance research reactors (USHPRR), plus one critical facility, that are actively collaborating with the National Nuclear Security Administration (NNSA) Material Management and Minimization (M 3 ) Reactor Conversion Program to convert from highly enriched uranium (HEU, ≥ 20 wt.% U-235) to low-enriched uranium (LEU, < 20 wt.% U-235) fuel. A new type of LEU fuel with very high density, based on an alloy of uranium and 10 weight percent molybdenum (U-10Mo), is expected to allow the conversion to LEU of USHPRR that have been found unable to be converted with previously qualified uranium silicide-aluminum (U 3 Si 2 -Al) dispersion fuel. MURR has been working with the USHPRR Reactor Conversion (RC) Pillar at Argonne National Laboratory to perform fuel element design and fuel cycle performance analyses, steady-state thermal hydraulics safety analyses, and accident safety analyses in preparation for the conversion of MURR and to support a preliminary Safety Analysis Report (SAR) for conversion to LEU fuel. In this work, Fluid-structure interaction (FSI) analysis at the fuel element level, as compared to the fuel plate level of the previous work. is performed which models all components of the MURR LEU fuel element, including fuel plates and the supporting structures e.g., side plates, end fittings, and combs. Therefore, the effect of supporting structures on the coolant flow distribution, the fuel plate deflection, and the resulting coolant channel gap reduction can be evaluated. In addition to the nominal element geometry and flow rate, the tolerances in the geometry dimensions of coolant channel and plate thickness, the effect of a comb on plate deflection, and the uncertainty of the flow rate per element are considered in this work. The effect of comb on plate displacement is quantified through two bounding cases: the case assuming a perfect bond between the comb and plates and the case neglecting the comb effect. Note that in this analysis, the FSI has been decoupled from the other structural effects caused by irradiation (e.g., swelling and irradiation creep). Assessment of combined effects is planned for a later stage of this project.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Numerical modeling of Joule heated ceramic melter

Computational fluid dynamics and heat transfer models of waste glass melters are being developed to support the vitrification campaign at the Waste Treatment and Immobilization Plant (WTP) as part of the United States nuclear waste stewardship and environmental protection programs. The WTP mission aims to achieve the stable and durable processing of vitrified waste in the safest, most efficient, and economical manner possible. Hanford’s tank waste is far more complex and varied than the waste treated by other vitrification facilities and both startup and long-term operational challenges are likely to be even greater than prior experience has shown. There are thousands of different waste compositions, which makes processing the waste a complex and challenging task. Models of test melters, pilot-scale, and full-scale melters have been developed to provide understanding and insight into the various physical and chemical phenomena occurring during vitrification. So the results of these modeling activities will be used to support the WTP operations to process legacy nuclear waste into a stable form for disposal.

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Integrated Traction Drive Thermal Management: Keystone Project 3

The presentation was developed for the U.S. Department of Energy (DOE) Vehicle Technologies Program 2020 Annual Merit Review and Peer Evaluation Meeting. It reports on progress and accomplishments of DOE funded Integrated Traction Drive Thermal Management project. Highlighted are NREL's research accomplishments during FY 2020. Challenges associated with thermal management of integrated traction drives for electric vehicles arise due to the limited packaging space, coolant distribution, electrical connections, and ambient temperatures. For this project at NREL, research focuses on enabling thermal management technologies to support a modular, compact, self-contained integrated electric drive unit with motor and power electronics. Integration approaches, based on packaging location and coolant options for power electronics integrated with candidate motor technologies to reduce cables and connectors, are reviewed and compared using finite element analysis (FEA) tools. Automatic transmission fluid (ATF) jet impingement cooling characterization is carried out experimentally (measuring heat transfer coefficients) and using computational fluid dynamics (CFD) modeling.

47 OTHER INSTRUMENTATION↗

Implementation of a High-Fidelity Interface Resolving Method in Nek5000

The development and utilization of computational fluid dynamics (CFD) models for large, high-temperature electric melters in the Waste Treatment and Immobilization Plant (WTP) in eastern Washington State have proven to be valuable for various purposes. These models allow for a better understanding of the physio-chemical processes occurring within the melter vessels and can contribute to improving operational efficiency, throughput, and addressing operational issues related to vitrification. The CFD models employed for these melter vessels incorporate multiphase fluid flow and heat transfer simulations in different regions, including the plenum, cold cap, and molten glass regions. As the tank waste and glass formers are introduced into the melter, a reacting batch layer known as the cold cap forms on top of the molten glass. To enhance the melt rate, forced convection bubblers located at the bottom of the melters generate convection currents that help homogenize the molten glass and provide heat to the cold cap. As the bubbles rise through the highly viscous glass, they adopt a spherical-cap shape [1]. Meanwhile, the conversion of the batch to glass generates significant amounts of gases (such as water vapor, carbon dioxide, sulfur dioxide, and NOx) due to thermal decomposition [2]. These gases become trapped between the cold cap and molten glass, forming a foam layer [3]. For modeling multiphase flow in CFD and heat transfer simulations of waste glass melters at different scales, efforts are underway to augment the capabilities of the Nek5000 [4] and NekRS [5] open-source codes [6]. Nek5000/NekRS is a scalable and efficient spectral element code that has been successfully applied to a wide range of fluid dynamics problems. By leveraging the Nek5000/NekRS software, it becomes possible to model the melter systems more affordably and with lower computational requirements compared to currently utilized commercial CFD software. The specific objectives of this ongoing effort include: 1. Implementation of a level set method in Nek5000/NekRS: The level set method is a numerical technique commonly employed in CFD simulations to track and represent the interface between different phases or materials accurately. By incorporating this method into Nek5000/NekRS, the ability to simulate multiphase flows in waste glass melters at a high level of fidelity can be achieved. 2. Demonstration of capability for air bubbling through molten glass: As part of the development process, a specific case of air bubbling through molten glass will be simulated using the augmented Nek5000/NekRS code. This demonstration aims to showcase the ability of the software to accurately capture and analyze the complex phenomena involved in the multiphase flow within waste glass melters. By achieving these objectives, the improved Nek5000/NekRS code will offer a powerful computational tool for simulating and analyzing waste glass melter systems, enabling better understanding, optimization, and troubleshooting of these vitrification processes. The ability to accurately model and simulate multiphase flows has broad relevance across many industries and scientific domains, and the improved functionality can contribute to advancements in various fields beyond waste glass melter simulations.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Effect of cold cap coverage and emissivity on the plenum temperature in a pilot-scale waste vitrification melter

Integrated computational fluid dynamics (CFD) models are being developed to model the complex physics occurring within the high-level waste melter for vitrification of legacy tank waste at the Hanford site. Here, we present a validation of the integrated CFD model by using data from two experimental runs in a pilot-scale melter. While the model uses several simplifying assumptions (such as constant heat sinks from a cooling jacket and inleakage of ambient air, steady state feed-to-batch conversion heat, and a cold cap model with a simplified shape), it closely predicts the molten glass (1150°C and 1175°C) and plenum temperatures (550°C) obtained from thermocouples during two pilot-scale tests, with an average cold cap coverage of 80%. Additional simulations were performed to explore the sensitivity of the predicted plenum temperatures to variations in cold cap coverage (fraction of melt surface covered by the glass batch) and batch emissivity. The plenum temperature was found to be in the range of 606°C when cold cap coverage decreased from 95% to 70%. Cold cap emissivity had a smaller effect, increasing the plenum temperature by as much as 179°C when cold cap emissivity increased from 0.2 to 0.8. Maintaining a high cold cap coverage without overfeeding is important for a sustained melter operation with high glass throughput. This work provides a tool for achieving that goal in terms of correlating the plenum temperature with the cold cap coverage.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Actively controlled coolant tank to increase thermal storage capacity of hybrid electric vehicles

An engine and cabin thermal management system for use with a vehicle having an engine, a cabin heating system configured to thermally heat a cabin of the vehicle, a coolant system operably coupled to the engine and to the cabin heating system to thermally manage a temperature of the engine and a temperature of the cabin. The coolant system having one or more coolant thermal storage units fluidly coupled with a radiator and heater core of the coolant system forming a coolant loop. The system further having a control system configured to monitor and maintain at least a predetermined coolant temperature at the cabin heating system even during a coolant temperature decrease at the engine stops.

Amini, Mohammad Reza↗

Modeling of SRS Aluminum-clad Spent Nuclear Fuel in Standard DOE Sealed Canisters

At the Savannah River Site (SRS), approximately 7 MTHM of aluminum-clad spent nuclear fuel is currently in wet storage at L-basin. A significant portion of this fuel is slated to be dried and sealed in DOE standard canisters for long-term storage. This report will focus on a few fuel types, which should bound all other fuel stored at SRS. These consist of the reference fuel assembly, the MURR fuel and the HFIR fuel. These bound a hypothetical maximum scenario for MTR box fuel, a maximum for decay heat in an assembly, and a maximum for the aluminum surface area, respectively. The geometry for the packages were created from DOE specifications for storage of each of the fuels. As long as the minimum time from reactor discharge to sealed storage is 3 years, then the highest temperatures which occur within the sealed canisters do not exceed 100° C for all scenarios considered here. In the geometry considered, while the HFIR fuel has a large surface area, the packaging configuration leaves large void areas, such that the aluminum surface to free volume ratio is about 30% less than an ATR or MURR packaging configuration.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

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↗

Modeling of ATR fuel in DOE Standard Canisters with Helium Backfilled Condition

Road-ready and final disposition packaging configurations for the advanced test reactor (ATR) fuel dictates storage within helium backfilled sealed DOE standard canisters. These sealed canisters are intended for extened (>50 year) dry storage). The typical packaging configuration for the 15-foot DOE canisters places 10 ATR elements within a Type 1a basket, and three baskets are loaded within each DOE canister. During in-reactor operations and cooling pond storage conditions, oxyhydroxide layers form on the surface of the aluminum clad fuel. These layers produce hydrogen gas over time due to the fuel’s radiation field. As part of the packing procedure, the ATR fuel should be dried to remove any residual physio-/chemi- sorbed water from the surface. As testing to the effectiveness of the drying procedure is still underway, this modeling will include results at fully saturated and fully dried conditions. In previous modeling efforts, the G-value for the production of hydrogen from the oxyhydroxide layers was assumed to be in argon environments as measured by Task 2 - Oxyhydroxide Layer Radiolytic Gas Generation Resolution. Previous experimental testing showed differences in the hydrogen generated based on the gaseous environment. In the associated experimental work, Task 2 - Oxyhydroxide Layer Radiolytic Gas Generation Resolution, additional tests were completed in a helium environment, and updated G-values for the radiolytic production of hydrogen from the oxyhydroxide layers were provided. These values are 28% and 58% higher than values for argon. In addition, a change to the modeling of the oxyhydroxide radiolysis has been made from previous reports. This change assumes the dependency of the dose rate on the overall reaction rate is applied to the total weight of the sample, rather than to just the weight of the oxyhydroxide layer. This decreases the dependency of the radiolytic reaction on the thickness of the oxyhydroxide layer. For a nominal scenario of stored ATR fuel, assuming the chemi-/physio- sorbed water have been fully removed, the internal canister pressure increases to 1.61 atm over a 50 year period, with a hydrogen mole percentage of 21%. As in previous modeling, any oxygen present is in negligible amounts (<1 ppt). If a small amount residual air is present, nitric acid can form up to 1300 ppm. For a scenario with high fuel decay heat, the model shows internal pressure increasing to 2.1 atm, with 39.3 mole percentage of hydrogen. In a scenario where significant chemi-/physio- sorbed water is present within the corrosion layer, the nominal scenario shows a pressure increase to 2.54 atm, with 21.1 mole percent hydrogen. The high decay heat case shows a pressure increase to 3.18 atm with 39.9 mole percent hydrogen.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Modeling Summary of ASNF in DOE Sealed Standard Canisters

A pathway for road-ready and final disposition packaging configurations for the aluminum-clad spent nuclear fuel (ANSF) fuel dictates storage within helium backfilled sealed DOE standard canisters. The typical packaging configuration for the 15-foot DOE canisters places 10 advance test reactor (ATR) elements within a Type 1a basket, and three baskets are loaded within each DOE canister. During in-reactor operations and cooling pond storage conditions, oxyhydroxide layers form on the surface of the aluminum clad fuel. These layers produce hydrogen gas over time due to the fuel’s radiation field. As part of the packing procedure, the ATR fuel should be dried to remove any residual physio-/chemi- sorbed water bound to the surface. The results of this modeling will include results at fully saturated and fully dried conditions. In addition, fuels that are currently stored at the Savannah River Site were also studied for their potential for hydrogen and pressure build up. These two additional fuels modeled were the Missouri University Research Reactor (MURR) fuel, which is packaged in the same configuration as the ATR, but with a 10-foot-tall DOE standard canister. This was selected due to its relatively high decay heat compared to other DOE-managed ASNF. The second additional fuel studied with the modeling effort was the High Flux Isotope Reactor (HFIR) fuel. This fuel is modeled as two separate DOE canisters with the inner and outer annulus split for storage. The HFIR was selected for study due to its high aluminum cladding surface area. In the associated experimental work, Task 2 - Oxyhydroxide Layer Radiolytic Gas Generation Resolution, additional tests were completed in a helium environment, and updated G-values for the radiolytic production of hydrogen from the oxyhydroxide layers were provided. These values were 2.92 ×10 -4 µmol/J at 50% relative humidity and 4.12 ×10 -4 µmol/J at 100% relative humidity. These values are lower than the value for Argon that was used in prior modeling results. In addition, prior modeling results have been completed with the G-value applied to just the mass of the corrosion layer, and this has been updated to apply the G-value for hydrogen generation to the full mass of the fuel. These two effects combine to show much smaller pressure and hydrogen build up for the sealed canister model. For a nominal scenario of stored ATR fuel, after 50 years the model results give a 1.36 atm total pressure, 7% mole percent hydrogen, for the upper decay heat, 1.51 atm total pressure, 16% mole percent hydrogen, and for upper decay heat with undried fuel 2.6 atm total pressure, 15% mole percent hydrogen. For the MURR nominal case, the model results give 1.34 atm and 6% hydrogen, for upper decay heat this gives 1.41 atm total pressure with 10.8 % hydrogen, and for upper decay heat with undried fuel, this gives 2.38 atm total pressure with 9.9% hydrogen. The nominal scenario for HFIR fuel gives 1.39 atm total pressure with 9.9% hydrogen, the upper decay heat case gives 1.43 atm with 12.1% hydrogen, and the upper decay heat with undried fuel gives 2.17 atm total pressure with 11.9% hydrogen. These results confirm the ATR scenario bounds the other intact ASNF modeled here. No case modeled yields significant oxygen, and the lower decay heat cases for all fuels modeled have hydrogen concentrations that are under the 4% flammability limit after 50 years of storage. In addition, the modeled pressures for all cases are all significantly below the 500-psi limit for the DOE standard sealed canister.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Chemical Modeling of ATR Fuel in DOE Standard Canisters with Borated Stainless Steel Corrosion

Road-ready and final disposition packaging configurations for the ATR fuel dictates storage within helium backfilled sealed DOE standard canisters. The packaging configuration for the DOE canisters places 10 ATR elements within a Type 1a basket, and three baskets are loaded within each DOE canister. In order to provide a criticality control mechanism, these stainless baskets are proposed to be built with 1-2 weight percentage of boron. This report seeks to identify effects of corrosion of this basket material over time. The primary reaction found through surveying literature was the oxidation of the iron in steel by water to produce some hydrogen. This reaction was built into the previous chemical model of the DOE canister to assess additional hydrogen and pressure build up over time. Literature for non-borated stainless steel, and carbon steel has shown that irradiation increases the rate of corrosion, and it is assumed a similar effect would occur with borated stainless steel. Due to the wide range of results in literature, and the variation of the corrosion with boron content, a wide range of corrosion rates was tested in the model from 80 nm/yr up to 1800 nm/yr.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Innovative mercury treatment technology options for the liquid waste system at the Savannah River Site: scoping studies

The Savannah River Site (SRS) Liquid Waste System (LWS) contains liquids, salts and sludges that are currently being processed into final wasteforms for disposition, specifically, waste glass for sludges, solids and liquids containing high levels of radioactivity, and saltstone grout for low activity decontaminated liquid solutions. The LWS also contains approximately 60,000 kg of mercury present in the following physical and chemical forms, specifically: a) ionic inorganic mercury, organomercury (e.g., methylmercury), and other minor components found in LWS fluids b) mercury solids such as oxides, hydroxides, amalgams sulfides and sorbed mercury, c) accumulations of dense liquid elemental mercury, and d) vapor phase elemental and organomercury mercury found in tank headspace gas and in evaporators. An effective and proactive management strategy of the mercury present in the LWS is needed to support processing of LWS wastes into glass and saltstone. Sustainable processing of the LAWS to completion requires mercury removal from the LWS at a rate of approximately 2,900 kg/yr. This removal can be accomplished through existing mercury treatment systems or newly implemented LWS “purge points”. The chemical speciation of mercury has emerged as the key factor that controls mercury behavior in the LWS. For example, past studies demonstrated that mercury speciation is adversely impacting the performance of existing removal systems so that significant levels of mercury are recycled from the Defense Waste Processing Facility (DWPF) back to the tank farm. Consequently, mercury concentrations have slowly increased in the LWS tank fluids over time. The presence of organo-mercury has also been identified as the cause of increased mercury leaching from saltstone. In response to these challenges, the Department of Energy (DOE) Office of Environmental Management (EM-TD) Technology Development Program has supported a series of scoping studies predicated on manipulating or controlling mercury speciation and mercury behaviors within the constraints of LWS waste chemistry and safety conditions. The intent of these studies was to rapidly triage potential technology options and develop a technically based go / no go recommendation for further work. This composite report presents the results of three scoping studies: 1) advanced photooxidation processes, 2) chemical reduction, and 3) mercury getters.

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NRAP-Open-IAM Multisegmented Wellbore Reduced-Order Model

Geologic carbon storage is one of the promising strategies to mitigate climate change by reducing the emission of carbon dioxide to the atmosphere. As part of the National Risk Assessment Partnership (NRAP), a systems-level stochastic analysis tool called the open source integrated assessment model, NRAP-Open-IAM, has been developed to estimate and manage the risk of containment loss at a geological carbon sequestration site. NRAP-Open-IAM contains several wellbore leakage model components that estimate the fluid leak rate that may occur through compromised legacy wells due to the increase in pressure resulting from CO 2 injection activities. Coupled to a reservoir component model, these components estimate the leakage of CO 2 and/or brine from a storage reservoir to overlying aquifer layers and the atmosphere through legacy wells. This report presents the theoretical framework and quality testing of the multisegmented wellbore reduced-order model. The model allows for segmenting of the legacy wells passing through the overlying stratigraphy into several intervals to simulate a site’s specific stratigraphic and hydrogeologic properties. For quality assurance, the analytical model is validated against numerical reservoir flow simulations for single and multiple aquifer(s) models. The results indicate that the model accurately predicts the transport of two-phase fluids (brine and injected CO 2 ) through the well over time. A detailed description of the model helps users to understand the model and provides a basis for future improvements.

58 GEOSCIENCES↗

ISR Restoration Methods and Applicability to Groundwater Remediation at Abandoned Uranium Mills - 20342

There are numerous abandoned uranium mining and mill sites in the southwestern US. Surface remediation was undertaken in the 1980's and 90's to cover mill tailings and minimize infiltration of meteoric precipitation into the subsurface These actions have reduced the availability of tailings contaminants but many of these sites have persistent groundwater contamination, primarily due to the disposal of tailings and process fluids in unlined impoundments during the mill's operational life. Active in situ recovery (ISR) uranium mining and wellfield restoration practices are examined to determine the factors that contribute to the success of groundwater restoration. Hydrologic characteristics, geochemistry, ore body quality, regulatory requirements and the aboveground components of uranium production and water treatment are considered. While there is no such thing as a typical abandoned mill site, the potential use of ISR restoration methods is evaluated for viability as a remedial alternative to mine and mill sites that have residual contaminated groundwater plumes. (authors)

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Computationally-Aided Design of a Small-Scale Radioactive Waste Glass Melter

To provide mission support to the Hanford Waste Immobilization and Treatment Plant for the vitrification of legacy nuclear tank waste, a reduced-scale vitrification pilot system is being designed to process simulated and actual radioactive tank wastes. The tank waste will be separated into high-level waste (HLW) and low-activity waste (LAW) fractions, where the majority by mass (~90%) is LAW and by activity (>95%) is HLW. The first tank waste to be processed at the WTP will be LAW. Since Hanford tank waste and the resultant melter feed compositions are known to vary widely, pilot-scale operations are essential to identify potential problems, provide needed data, confirm assumptions, and determine impacts to full-scale melters and off-gas systems from these different feeds. In addition, the reduced-scale melter system must be capable of quickly providing results to operations without the typically high costs of radioactive operations and provide an engineering platform in close proximity to local operations staff. The objective is to produce similar process conditions to those encountered in the full-scale LAW melters and minimize the volume of radioactive waste necessary for the evaluations. To this end, melter design features are being explored to increase production without significantly increasing surface area, melter glass volumes or compromising data quality. In support of these objectives, a set of computational fluid dynamic models have been developed to provide insight into the flow patterns within these small melters and evaluate design features to increase production without significantly increasing glass inventory. Computed velocities were evaluated at the interface between the cold cap and the molten glass pool and within the bulk glass pool. A method was developed to correlate results to an estimated melt rate and down-select design features that produce the highest gains.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Dielectric Fluids for the Direct Forced Convection Cooling of Power Electronics

The future of electrification of vehicles and other systems will require the creation of high-power density power electronics with low junction-to-fluid thermal resistance cooling solutions. One way to create this solution is to move high-heat-flux liquid cooling (single- or two-phase) as close to the power electronics components as possible. One novel approach involves submersion in dielectric fluids as the cooling solution. We first provide the range of fluid properties and develop a figure of merit (FOM) to aid in dielectric fluid selection. Next, we perform computational fluid dynamics/heat transfer (CFD/HT) modeling using single-phase cooling (submerged jet impingement on an enhanced surface) to validate the dielectric fluid FOM. Results of the study show that the developed FOM is a good representation of the performance of the fluids when compared to the results of the CFD/HT analysis. Both FOM and the CFD/HT analysis show that based on pure thermohydraulic considerations, several commercially available fluids present higher performance, on the order of 5% of water. Finally, the FOM can be used to quickly assess the thermohydraulic performance of a dielectric fluid, as well as the secondary application-specific properties such as boiling point, saturation pressure, flash point, and global warming potential, thereby allowing for fluid candidates to be readily compared.

computational fluid dynamics↗