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DOE OSTI2020

Electrolyte compositions comprising ionic liquids and metal hydride batteries comprising same

Electrolyte compositions comprising a) an ionic liquid and b) a protic acid and/or an organic solvent are suitable for use in electrochemical cells, e.g. metal hydride batteries. The electrolyte compositions may replace the currently employed 30% by weight aqueous KOH. Suitable protic acids include carboxylic acids, mineral acids, sulfonic acids and the like. Suitable organic solvents include organic carbonates, ethers, glymes, ortho esters, polyalkylene glycols, esters, lactones, glycols, formates, sulfones, sulfoxides, amides, alcohols, ketones, nitro solvents, nitrile solvents and combinations thereof. Present batteries may achieve a nominal open-circuit voltage of >1.2 V (volts) and up to about 6 V. The electrolyte compositions allow enlargement of the electrochemical window, thus allowing the use of further cathode active materials. Further cathode active materials include transition metals and their oxides, hydroxides and fluorides; for example, Sc, Ti, V, Cr, Mn, Fe, Co, Ni, Cu, Zn, Y, Zr, Nb, Mo, Tc, Ru, Rh, Pd, Ag, Cd, Lu, Hf, Ta, W, Re, Os, Ir, Pt and Au and their oxides, hydroxides, oxide/hydroxides and fluorides.

25 ENERGY STORAGE↗

DOE OSTI2020

Growth and Properties of Vapor Diffused Nb3Sn Coating for Superconducting Radiofrequency Accelerator Cavity Applications

The desire to reduce construction and operating costs of future SRF accelerators motivates the search for higher-performing alternative materials. Nb3Sn (Tc ~ 18.3 K and Hsh ~ 425 mT) is the front runner. The tin vapor diffusion process is currently the technique of choice to produce promising Nb3Sn-coated cavities. Understanding Nb3Sn nucleation and growth in this process is essential to progress. Samples representing different stages of Nb3Sn formation have been produced and studied to elucidate the effects of nucleation, growth, process conditions, and impurities. Nb3Sn films with thickness from a few hundred nm up to ~15 µm were grown and characterized. The microscopic examination of samples suggests the mechanisms of thin film nucleation and growth. Broadly, nucleation deposits tin as a thin surface phase and, under some conditions, a few hundred nanometer sized particles as well. Conditions that impair nucleation promote the formation of defects, such as ?patches?, in subsequent coating growth. Analysis of coated samples is consistent with the model of Nb3Sn grown in which tin diffuses via grain boundaries to Nb3Sn-Nb interface, where the growth Nb3Sn into the niobium bulk takes place. Similar scaling laws are found for grain growth and layer thickness. Non-parabolic layer growth is consistent with significant grain growth, which reduces the number of Sn-transport channels. Examination of patchy region in Nb3Sn coating revealed it to be large single crystalline grains, pointing to impeded Nb3Sn layer growth due to low grain boundary density, resulting in a significantly thin coating in those areas. Examination of RF loss regions from a coated cavity, identified with a thermometry mapping system showed patchy regions and carbonus defects were associated with strong local field-dependent surface resistance. RF measurements of coated cavities were combined with material characterization of witness samples and coated-cavity cutouts to improve the coating process. Understanding obtained and applied to cavity coatings, resulted in single-cell Nb3Sn cavities with a quality factor of ~2 ×1010 up to 15 MV/m accelerating gradient at 4 K, without "Wuppertal" Q-slope. We have also produced Nb3Sn-coated CEBAF 5-cell cavities with accelerating gradients useful for accelerator cryomodules. This dissertation will discuss the genesis of the Nb3Sn coating in a typical tin vapor diffusion process, effects of different process parameters, and its consequences to the coating of single-cell and multi-cell SRF cavities.

Pudasaini, Uttar↗

DOE OSTI2021

AGR-5/6/7 Irradiation Test Final As-run Report

This document presents the as-run analysis of the Advanced Gas Reactor (AGR) 5/6/7 irradiation experiment. AGR-5/6/7 is the last of a series of experiments conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) in support of the development and qualification of tri-structural isotropic (TRISO) low enriched fuel for use in high-temperature gas-cooled reactors (HTGR). The test train contained five separate capsules that were independently controlled and monitored. Each capsule contained multiple 24.91-mm-long and 12.25-mm-dimeter compacts filled with low-enriched uranium carbide/oxide tri-structural isotropic fuel particles. The objectives of the AGR 5/6/7 experiment were to: 1. Irradiate reference-design fuel particles to support fuel qualification. 2. Establish operating margins for the fuel, beyond normal operating conditions. 3. Provide irradiated-fuel performance data and irradiated-fuel samples for post irradiation examination and safety testing. The primary objective of the AGR-5/6 test (Capsules 1, 2, 4, and 5) was to verify the successful performance of the reference-design fuel under normal operating conditions. The AGR-7 test (Capsule 3) was designed to explore fuel performance at higher temperatures. Its primary objective is to demonstrate the capability of the fuel to withstand conditions beyond normal operating conditions, in support of plant design and licensing. AGR 5/6/7 will also provide irradiated-fuel performance data based on the fission gas release from particles during irradiation. To achieve the test objectives, the AGR-5/6/7 experiment was irradiated in the northeast (NE) flux trap of the ATR with a planned duration of 500 effective full-power days (EFPDs). The NE flux trap was selected because its larger diameter provided greater flexibility for test-train design compared to the Large B positions used for the AGR-1 and AGR-2 irradiations, significantly enhancing test capabilities for the combined irradiation campaigns. Due to delays in the ATR schedule, the AGR-5/6/7 irradiation was significantly shorter than the originally planned 13-cycle schedule. Irradiation began on February 16, 2018 and ended on July 22, 2020, spanning nine ATR cycles (162B?168A) over two and a half years. Thus, the AGR-5/6/7 fuel compacts were irradiated for a total of approximately 360.9 EFPDs. Final burnup values, on a per-compact basis, ranged from 5.66% to 15.26% fissions per initial heavy metal atom, while fast fluence values ranged from 1.62 to 5.55 × 1025 n/m2 (E >0.18 MeV). Time-averaged volume-averaged fuel temperatures on a capsule basis at the end of irradiation ranged from 756°C in Capsule 5 to 1313°C in Capsule 3 excluding days with significantly lower temperature during the two short powered axial locator mechanism cycles, 163A and 167A. By the end of irradiation, 48 out of 54 installed thermocouples (TC) had failed (the bottom three capsules lost all TCs). During the first five cycles (162B ? 165A), fission-gas isotope release-rate-to-birthrate (R/B) ratios were stable in the 10-8?10-6 range, and no in-pile particle failures were observed based on the gross gamma counts. During this time, the high exposed kernel fraction and high fuel particle temperatures in Capsule 1 led to the maximum R/B value of around 2 ? 10-6 for Kr-85m. The fission gas release in all capsules started to increase from the second half of Cycle 166A, when a large number of in-pile particle failures occurred in Capsule 1 and a gas line problem in this capsule caused fission gas leakage at various degrees into the other four capsules. This gas line problem also prevented fission gas release measurement for Capsule 1 during the last three cycles due to gas flow isolation. By the end of irradiation, it is estimated that approximately 15 particle failures in Capsule 3, which was considered possible at the start of the irradiation, since the experiment was designed to operate beyond the high-temperature gas-cooled reactor (HTGR) normal operating temperature range. A few hundred in-pile particle failures were estimated for Capsule 1 by the end of Cycle 166A, but the total number of failures is unknown due to lack of fission gas release data in the later cycles. Additionally, four potential in-pile failures were identified for Capsule 2 during the last cycle, 168A. In contrast, no in-pile failures were identified in the top two capsules (4 and 5) based on absence of the typical spikes in gross gamma counts and low failure estimates using the AGR-3/4 R/B per exposed kernel model.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

Summary of Thermocouple Performance in the Advanced Gas Reactor Experiment AGR-5/6/7 During Irradiation in the Advanced Test Reactor

The Advanced Gas Reactor -5/6/7 (AGR-5/6/7) experiment was the fourth and final experiment in the AGR experiment series and completed irradiation in July 2020. It serves as the formal fuel qualification test for the TRISO fuels under development by the US Department of Energy. This experiment was designed to irradiate fuel particles at temperatures ranging from 800°C ? 1500°C. The high end of the range created unique challenges for thermocouple-based temperature measurements. Commercially available high-temperature platinum-rhodium thermocouples (Types S, R, and B) and tungsten-rhenium thermocouples (Type C) suffer rapid decalibration due to transmutation of the thermoelements from neutron absorption. A special low neutron cross-section thermocouple system based on molybdenum/niobium thermoelements called High Temperature Irradiation Resistant thermocouple (HTIR-TC) has been under development at INL since circa 2004. Several examples of this thermocouple type were incorporated into regions of the test operating above 1200°C. For regions in the test operating in the 1000°C ? 1200°C range, high-performance versions of Type N thermocouples recently developed at Cambridge University were installed. Standard Type N thermocouples were used in regions of the test operating below 1000°C. A total of 54 thermocouples were incorporated into the test. A report of the performance of this large heterogenous set of thermocouples over the first third of the irradiation was previously provided (ANIMMA 2019). This follow-on paper discusses results from the balance of the experiment (Feb 2019 - July 2020). Special attention is focused on potential drift of thermocouples operating in the higher temperature regions.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2021

Summary of Thermocouple Performance in the Advanced Gas Reactor Experiment AGR-5/6/7 During Irradiation in the Advanced Test Reactor

This is a conference presentation ppt file. The following abstract is for the entire conference paper - which will later be the subject of a PRS artifact The Advanced Gas Reactor -5/6/7 (AGR-5/6/7) experiment was the fourth and final experiment in the AGR experiment series and completed irradiation in July 2020. It serves as the formal fuel qualification test for the TRISO fuels under development by the US Department of Energy. This experiment was designed to irradiate fuel particles at temperatures ranging from 800°C ? 1500°C. The high end of the range created unique challenges for thermocouple-based temperature measurements. Commercially available high-temperature platinum-rhodium thermocouples (Types S, R, and B) and tungsten-rhenium thermocouples (Type C) suffer rapid decalibration due to transmutation of the thermoelements from neutron absorption. A special low neutron cross-section thermocouple system based on molybdenum/niobium thermoelements called High Temperature Irradiation Resistant thermocouple (HTIR-TC) has been under development at INL since circa 2004. Several examples of this thermocouple type were incorporated into regions of the test operating above 1200°C. For regions in the test operating in the 1000°C ? 1200°C range, high-performance versions of Type N thermocouples recently developed at Cambridge University were installed. Standard Type N thermocouples were used in regions of the test operating below 1000°C. A total of 54 thermocouples were incorporated into the test. A report of the performance of this large heterogenous set of thermocouples over the first third of the irradiation was previously provided (ANIMMA 2019). This follow-on paper discusses results from the balance of the experiment (Feb 2019 - July 2020). Special attention is focused on potential drift of thermocouples operating in the higher temperature regions.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2022

2.1.1.1.401 - Wave Energy Converter Modeling

The Wave Energy Converter (WEC) Modeling Project began in 2013 with the aim of driving innovation and advancing the state of the wave energy industry by developing publicly available, easy to use software that's customizable to meet end user needs. This project is primarily focused on development of open source software for numerical design and analysis of wave energy converters (WEC), and it also supports international collaboration on standards (IEC TC 114), code verification and validation (IEA OES Task 10) and a competition (WECCCOMP).

MATHEMATICS AND COMPUTING,TIDAL AND WAVE POWER↗

DOE OSTI2023

A Revised International Standard for Gearboxes in Wind Turbine Systems

Gearbox and wind turbine design and application standards have contributed significantly to improvements in reliability over the past two decades. The International Electrotechnical Commission (IEC) 61400-4 standard of wind turbine gearbox design is currently being revised by a joint working group (JWG) of experts in IEC TC 88 (wind energy) and International Organization for Standardization (ISO) TC60 (gears) to further that effort. Experts from ISO TC4 (rolling bearings) and ISO TC28 (lubricants) have actively participated. This revision has implemented lessons learned from industry use of edition 1 since its publication in 2012. The main document, IEC 61400-4, was pared down to essential design requirements and application-specific recommendations along with a design verification framework. The JWG leveraged concurrent development of other standards, such as IEC 61400-8 on wind turbine structures, to replace edition 1 content. These are described along with how this works with the IEC Renewable Energy certification scheme for wind turbines (IECRE-WE). The JWG recognized the interest in maintaining informative parts of edition 1 including annexes on wind turbine architecture and loads, bearing and gear arrangements, bearing selection, lubrication system descriptions and lubricant performance recommendations. This information was retained in two technical reports: IEC/TR 61400-4-2 Lubrication and IEC/TR 61400-4-3 Explanatory Notes. Additionally, a technical specification, IEC/TS 61400-4-1, was drafted to provide a reliability calculation method for comparing different design options or conditions. Salient elements of these documents are described. All four documents were recently distributed for IEC/ISO review, ballot, and comment. Publication is expected in 2023.

gearbox↗

DOE OSTI2023

A Revised International Standard for Gearboxes in Wind Turbine Systems: Preprint

ENGINEERING,WIND ENERGY↗

DOE OSTI2023

First Results From Nb3Sn Coatings of 2.6 GHz Nb SRF Cavities Using DC Cylindrical Magnetron Sputtering System

A DC cylindrical magnetron sputtering system has been commissioned and operated to deposit Nb3Sn onto 2.6 GHz Nb SRF cavities. After optimizing the deposition conditions in a mock-up cavity, Nb-Sn films are deposited first on flat samples by multilayer sequential sputtering of Nb and Sn, and later annealed at 950°C for 3 hours. X-ray diffraction of the films showed multiple peaks for the Nb3Sn phase and Nb (substrate). No peaks from any Nb- Sn compound other than Nb3Sn were detected. Later three 2.6 GHz Nb SRF cavities are coated with ~1 μm thick Nb3Sn. The first Nb3Sn coated cavity reached close to Eacc = 8 MV/m, demonstrating a quality factor Q0 of $3.2 \times 108$ at Tbath = 4.4 K and Eacc = 5 MV/m, about a factor of three higher than that of Nb at this temperature. Q0 was close to $1.1 \times 109$, dominated by the residual resistance, at 2 K and Eacc = 5 MV/m. The Nb3Sn coated cavities demonstrated Tc in the range of $17.9 -- 18$ K. Here we present the commissioning experience, system optimization, and the first results from the Nb3Sn fabrication on flat samples and SRF cavities.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗

DOE OSTI2023

Demonstration of Niobium Tin in 218 MHz Low-beta Quarter Wave Accelerator C

A 218 MHz quarter wave niobium cavity has been fabricated for the purpose of demonstrating Nb3Sn technology on a low-beta accelerator cavity. Niobium-tin has been established as a promising next generation SRF material, but development has focused primarily in high- beta elliptical cell cavities. This material has a significantly higher TC than niobium, allowing for design of higher frequency quarter wave cavities (that are subsequently smaller) as well as for significantly lowered cooling requirements (possibly leading to cryocooler based designs). The fabrication, initial cold testing, and Nb3Sn coating are discussed as well as test plans and details of future applications.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗

DOE OSTI2023

Exploration of Parameters That Affect High Field Q-Slope

The onset of high field Q-slope (HFQS) around 25 MV/m prevents cavities in electropolished (EP) condition from reaching high quality factors at high gradients due to the precipitation of niobium hydrides during cooldown. These hydrides are non-superconducting at 2 K, and contribute to losses such as Q disease and HFQS. We are interested in exploring the parameters that affect the behavior of HFQS. We study a high RRR cavity that received an 800 C by 3 hour bake and EP treatment to observe HFQS. First, we explore the effect of trapped magnetic flux. The cavity is tested after cooling slowly through Tc while applying various levels of ambient field. We observe the onset of the HFQS and correlate this behavior with the amount of trapped flux. Next, we investigate the effect of the size/concentration of hydrides. The cavity is tested after holding the temperature at 100 K for 12 hours during the cooldown to promote the growth of hydrides. We can correlate the behavior of the HF QS with the increased hydride concentration. Our results will help further the understanding of the mechanism of HFQS.

43 PARTICLE ACCELERATORS↗

DOE OSTI2023

Demonstration of Niobium Tin in 218 MHz Low-beta Quarter Wave Accelerator Cavity

A 218 MHz quarter wave niobium cavity has been fabricated for the purpose of demonstrating Nb3Sn technology on a low-beta accelerator cavity. Niobiumtin has been established as a promising next generation SRF material, but development has focused primarily in high-beta elliptical cell cavities. This material has a significantly higher TC than niobium, allowing for design of higher frequency quarter wave cavities (that are subsequently smaller) as well as for significantly lowered cooling requirements (possibly leading to cryocooler based designs). The fabrication, initial cold testing, and Nb3Sn coating are discussed as well as test plans and details of future applications.

43 PARTICLE ACCELERATORS↗

DOE OSTI2023

First Results from Nb3Sn Coatings of 2.6 GHz Nb SRF Cavities Using DC Cylindrical Magnetron Sputtering System

A DC cylindrical magnetron sputtering system has been commissioned and operated to deposit Nb3Sn onto 2.6 GHz Nb SRF cavities. After optimizing the deposition conditions in a mock-up cavity, Nb-Sn films are deposited first on flat samples by multilayer sequential sputtering of Nb and Sn, and later annealed at 950 {\deg}C for 3 hours. X-ray diffraction of the films showed multiple peaks for the Nb3Sn phase and Nb (substrate). No peaks from any Nb3Sn compound other than Nb3Sn were detected. Later three 2.6 GHz Nb SRF cavities are coated with ~1 $\mu$m thick Nb3Sn. The first Nb3Sn coated cavity reached close to Eacc = 8 MV/m, demonstrating a quality factor Q0 of 3.2 x 108 at Tbath = 4.4 K and Eacc = 5 MV/m, about a factor of three higher than that of Nb at this temperature. Q0 was close to 1.1 x 109, dominated by the residual resistance, at 2 K and Eacc = 5 MV/m. The Nb3Sn coated cavities demonstrated Tc in the range of 17.9 - 18 K. Here we present the commissioning experience, system optimization, and the first results from the Nb3Sn fabrication on flat samples and SRF cavities.

43 PARTICLE ACCELERATORS↗

DOE OSTI2023

Advanced ground thermal conductivity testing

Determination of Ground Thermal Conductivity (TC) in the field of Geothermal Heat Pumps (GHPs). The invention provides a simple mathematical approach using minimal terms with apparatus not dependent on stable electric power generation.

Clemenzi, Richard A.↗

DOE OSTI2020

Disulfonamide Ligands as f-Element Extractants from Alkaline High-Level Waste

Almost 100 million gallons of high level waste (HLW) has been generated from defense reprocessing programs to support nuclear weapons production. HLW is currently stored at the Hanford and Savannah River Sites (SRS) [1]. It contains radioactive components, such as {sup 137}Cs, {sup 99}Tc and {sup 90}Sr, as well as large amounts of non-radioactive species, including solvated cations and insoluble metal hydroxides. HLW has high concentration of soluble hydroxides ([OH{sup -}] = 2 M at SRS [2]), complexing inorganic anions ([NO{sub 3}{sup -}] ∼ 0.65 - 3.7 M at SRS [2]) and a high ionic strength. Even though the majority of actinide component in alkaline HLW is precipitated, studies of An(III) and Ln(III) complexation in highly alkaline solutions in the presence of high nitrate concentrations have showed that soluble nitrate complexes can be formed [3]. The current treatment of HLW at SRS currently consists of two processes: The Actinide Removal Process (ARP) and the Next-Generation Caustic-Side Solvent Extraction process (NG-CSSX). Strontium and actinides are removed by sorption on monosodium titanate through the Actinide Removal Process (ARP) commonly referred to as the alpha-strike step [4]. Then, cesium is selectively extracted from the alkaline media via the Caustic-Side Solvent Extraction (CSSX) process [5]. The CSSX solvent consists of a calix[4]arene-crown-6 extractant dissolved in an inert hydrocarbon matrix containing i) a solvent modifier (alkylphenoxy alcohol) which increases extractant solubility and prevents third phase formation, and ii) a suppressor (trioctylamine or guanidine) - that mitigates surfactant effects. The CSSX process removes {sup 137}Cs selectively and rapidly, yet the post- CSSX aqueous stream may still contain high amounts of An, thus requiring a second ARP treatment (referred to as 'alpha-finishing') for some tanks. Despite the success of the ARP process in removing Sr and An, it often represents the kinetic bottleneck of integrated processing, as it is slower than solvent extraction. Understanding the complexation of actinides by organic ligands that are compatible with the CSSX process could eventually lead to a combined caustic-side Cs/Sr/actinide extraction process with better economics due to a reduced amount of monosodium titanate and/or a shorter required contact time with titanate during ARP. This would ensure a low-activity waste (LAW) stream with no actinides without additional ARP processing post-CSSX. Disulfonamides were studied for extraction of Sm(III) from alkaline aqueous media of pH 10-14 into dichloromethane. Up to 82% of Sm(III) was extracted from solutions of pH 12.5 -13.5 and up to 84% from solutions of pH 10.5 - 11.5. These results show some resemblance with Am extraction results by calixarene ligands previously reported [10]. Kinetic studies demonstrated that even 5 min is enough to complete stripping, whereas extraction is time-limiting process and requires up to 20 h for efficient removal of Sm(III) from alkaline aqueous media. Determination of composition of complexes in solution after extraction by the equilibrium shift method showed a 1:1 Sm(III):dsa-2 complexation ratio for the extracted species.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2020

Sorption of Anionic Iodine and Molybdenum to Multi-sorbing Barrier (MSB) Materials - 20030

A multi-sorbing barrier (MSB) material containing sodium bentonite (NaB), organoclay, and anion resin is being evaluated for containment of anionic radionuclides. Batch tests on the MSB were conducted using two non-radioactive analog isotopes ({sup 127}I, {sup 95}Mo) as surrogates for {sup 129}I and {sup 99}Tc due to their similar valence and molecular weight relative to the radionuclides. Tests were conducted on the organoclay, anion resin, and NaB individually and on NaB-resin blends with resin contents of 1% and 5% by mass. The anion resin exhibits very strong sorption of {sup 127}I and {sup 95}Mo relative to the other materials. The organoclay exhibited modest sorption of {sup 95}Mo and {sup 127}I, and NaB little sorption. The sorption isotherms follow a linear model for NaB, a Freundlich model for the organoclay, and a Langmuir model for the anion resin and NaB-resin mixtures. Combining resin with NaB significantly enhances sorption relative to NaB alone, with a larger resin fraction yielding greater sorption. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2020

Modeling Efforts to Gain Insight into Historical Leak Events from the Single-Shell Tank A-105 in the Hanford 241-A Tank Farm - 20113

The 241-A Tank Farm is a single-shell tank (SST) farm constructed to store process waste from Hanford nuclear operations. Millions of gallons of nuclear waste were stored in the 241-A Tank Farm SSTs and some of the SSTs leaked in the past. In addition, spills and pipeline leaks during transfers and storage and intentional discharges to cribs and trenches resulted in releasing waste to the ground. Tank liner leaks are referred to as 'leaks' and all other discharges to the soil are referred to as 'releases.' Liquid waste that could be removed by pumping has been removed from all of the SSTs to reduce the potential for future leaks. The most significant historical leak event in 241-A Tank Farm occurred in 1965 when tank A-105 experienced a pressurized steam event. This event resulted in damage to the inner steel liner flooring of A-105, which was separated from the sidewalls over part of its circumference and buckled up to 2.5 m, vertically. During this event up to 7,570 L of contaminated liquid may have leaked to the adjacent soil. This leak occurred under extreme temperature and pressure conditions. Modeling efforts to better understand the nature and extent of historical leaks from subsurface SSTs that were used to store highly radioactive, self-boiling liquid wastes at 241-A Tank Farm have recently been undertaken. These efforts have compiled the available historical information to support a preliminary non-isothermal, multiphase flow and chemical transport modeling effort to re-create the conditions under which the leaks occurred and to formulate a conceptual model as to the extent and distribution of leaked radioactive contaminants in the adjacent soil material. The focus of this work was on developing an understanding of the key features, processes and bounding conditions related to tank A-105 leak events that occurred in the 1960's. The activities include estimation of leak composition, non-isothermal multiphase flow and transport modeling, and geochemical modeling. The STOMP{sup C} Water-Air-Energy modeling code was utilized to implement the three-dimensional representation of the subsurface tank and the surrounding flow field. Once constructed, the model was used to evaluate the conceptual understanding of leaks originating from different parts of the tank under the elevated temperature boundary conditions that were imposed by surrounding tanks at the time of the historical leaks. The chemical evolution of the liquid waste was also evaluated as it leaked from the tank under transient pressure and temperature gradients. The results of the preliminary evaluation suggested the presence of a heat-pipe effect beneath tank A-105 in which water vapor at an elevated temperature is driven away from the base of the tank to a position where water vapor cools, condenses and is then drawn back toward the tank by the strong capillary attraction of the dry soil. A non-sorbing contaminant (Tc-99) was introduced into this flow field to better understand the potential distribution patterns of leaked contaminants that may have occurred during the historical leak events. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2020

The Use of Silver Chloride Injection in Remediation of Iodine-129 by In Situ Capture as Silver Iodide at the F-Area Seepage Basin - 20225

The Savannah River Site (SRS) produced tritium, plutonium, and special nuclear materials for national defense, medicine, and the space programs. As part of operations, the F-Area Seepage Basins operated until 1988 for the disposition of deionized acidic wastewater from the F Separations Facility. The wastewater contained dilute nitric acid and low concentrations of non-radioactive metals, and radionuclides, with the major isotopes being Cs-137, Sr-90, U-235, U-238, Pu-239, Tc-99, I-129, and tritium. The seepage basins were closed in 1988 and backfilled and capped by 1991. The groundwater emanating from beneath the closed and capped seepage basins is acidic and contains elevated levels of both chemical and radiological contaminants. Releases from the groundwater plumes sourced from the F-Area Seepage Basins have impacted the water quality of Fourmile Branch, which is a small tributary to the Savannah River, a regional water source. A large pump-and-treat system was constructed in 1997 and operated until 2003 in an attempt to capture the releases to Fourmile Branch. The system in F Area and a similar system in H Area were expensive (∼$1.3 M/month) to operate and produced large quantities of radioactive waste, with concentrations of I- 129 too high to be disposed of at the SRS. In 2004, SRS replaced pump-and-treat with a funnel and gate system that along with operation of a base injection system at the gates reduces the flux of contaminants to the wetlands adjacent to Fourmile Branch. The alkaline solution injected into the aquifer neutralizes the acidic plume and immobilizes many of the cationic constituents. However, base injection is not effective in managing the release of iodine-129, an anionic contaminant. To address iodine-129 SRS and the Savannah River National Laboratory developed an in situ technology that uses ultra-fine ground silver chloride (AgCl) as an injectable capture medium for the sequestration of iodine-129. The AgCl amendment has a very small particle size and is designed to be injected into the contaminated aquifer to capture iodine-129. Dissolved iodine-129 forms a stable and highly insoluble solid (silver iodide) upon contact with AgCl. Laboratory studies, a field scale pilot test (2009), and three deployments (2011, 2015 and 2019) of AgCl have been successfully performed at the F-Area Seepage Basins. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗