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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Design and Modeling of Vacuum Pumping for Steady-State Pellet Fueling Systems

We report that future long-pulse magnetic confinement fusion reactors will require density and isotopic mixture control using steady-state repeating pellet injectors. For high-energy density burning plasmas, pellet velocities of 1 km/s and above will be required for sufficient plasma penetration to achieve high fueling efficiency. Currently, steady-state repeating injection systems utilize cryogenic extruder systems to produce an extrusion of solid deuterium or deuterium-tritium. In repeating light gas gun injectors, the solid extrusion is cut and simultaneously loaded into a barrel. Once loaded, a fast operating gas valve delivers a high pressure burst of gas to accelerate the pellet down the barrel and into the machine. This process takes ~10 ms to achieve. Adequate gas pumping of the extruder exhaust and injection line propellant gas collection chambers is necessary for optimal operation of the pellet fueling system. Excess solid from the extruder sublimates in an exhaust chamber. The gas pressure in the extruder exhaust chamber must remain low to maintain low heating loads on the cooling mechanism (cryorefrigerators or liquid helium flow) and to reduce thermal conduction to the extrusion. Pumping the injection line chambers is necessary to limit propellant gas flow into the machine. A numerical simulation code was created to predict temporal pumping performance for these repeating pellet injection systems. Lastly, this paper outlines the methods and assumptions used to create this model and compares results to the pellet injection system currently employed on DIII-D, the steady-state pellet injection system planned for the Wendelstein 7-X, and a brief analysis of the ITER conceptual pellet fueling system.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2015

Using ASCEM Modeling and Visualization to Inform Stakeholders of Contaminant Plume Evolution and Remediation Efficacy at F-Basin Savannah River, SC – 15156

Communication with stakeholders, regulatory agencies, and the public is an essential part of implementing different remediation and monitoring activities, and developing site closure strategies at contaminated sites. Modeling of contaminant plume evolution plays a critical role in estimating the benefit, cost, and risk of particular options. At the same time, effective visualization of monitoring data and modeling results are particularly important for conveying the significance of the results and observations. In this paper, we present the results of the Advanced Simulation Capability for Environmental Management (ASCEM) project, including the discussion of the capabilities of newly developed ASCEM software package, along with its application to the F-Area Seepage Basins located in the U.S. Department of Energy Savannah River Site (SRS). ASCEM software includes state-of-the-art numerical methods for simulating complex flow and reactive transport, as well as various toolsets such as a graphical user interface (GUI), visualization, data management, uncertainty quantification, and parameter estimation. Using this software, we have developed an advanced visualization of tritium plume migration coupled with a data management system, and simulated a three-dimensional model of flow and plume evolution on a high-performance computing platform. We evaluated the effect of engineered flow barriers on a nonreactive tritium plume, through advanced plume visualization and modeling of tritium plume migration. In addition, we developed a geochemical reaction network to describe complex geochemical processes at the site, and evaluated the impact of coupled hydrological and geochemical heterogeneity. These results are expected to support SRS’s monitoring activities and operational decisions.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2026

Design and Commissioning of a Deuterium-Tritium Gas Delivery System for Muon Catalyzed Fusion in a Diamond Anvil Cell

We report the design, commissioning, and operation of deuterium-deuterium (DD) and deuterium-tritium (DT) gas delivery systems developed to load a diamond anvil cell (DAC) beam target for muon-catalyzed fusion (muCF). The DAC approach enables DT fuel to be compressed to GPa pressures at more than twice the liquid density and heated from cryogenic temperatures through 500 K, opening access to a substantially expanded parameter range for muCF kinetics and yield measurements. In this approach, DT is cryo-condensed to a liquid in a minichamber and then compressed in the DAC using a helium-driven pneumatic membrane, achieving high pressures in a millimeter-scale DT sample volume. A DD gas delivery system was designed and used to validate the experimental apparatus, measure the gas quantities needed for filling, develop operational experience, and collect kinetics and yield data with DD targets. The DT gas delivery system adds tritium-specific capabilities for inventory minimization, secondary containment, and activity monitoring. The DT system integrates depleted uranium storage beds and a liquid helium cryogenic condenser used for pressure building and cryopumping. High-purity delivery is provided by a rapid-response palladium permeator. The system is housed in a helium-atmosphere glovebox held at negative pressure with continuous cleanup. We present the process and instrumentation design, a failure modes and effects analysis (FMEA), and data from the experiment's in situ Raman spectrometer, which provides direct confirmation of target loading and composition through the optically clear diamond anvils. The 2024 and 2025 DT campaigns achieved repeatable target fills and operation with no measurable tritium releases to the stack, demonstrating safe, high-purity DT loading at novel density-temperature conditions for muCF studies.

Koukina, Elena [Acceleron Fusion]↗
DOE OSTI2018

Evaluation of getters for methane and ammonia decomposition

Getters are used in many tritium purification systems for impurity removal. One getter material, SAES® ST909, can be used to decompose various impurities (methane, carbon oxides, ammonia and others), and release the hydrogen bound in these compounds for further processing. Because, ST909 absorbs eight to nine percent of its initial weight in nitrogen, alternate materials may provide greater operational flexibility to accommodate variable operating carrier gases. The gettering capability of ST909 is related to the tri-metallic alloy (zirconium, manganese, and iron) composition. However, simply using a material with the same composition may not provide similar methane and ammonia decomposition conversion efficiency. For this testing, alternative materials to ST909 are compared to track methane and ammonia decomposition performance. Bench scale tests were performed at elevated temperature using various gas mixtures. Changes in methane and ammonia decomposition performance were tracked using RGA analysis during bench scale testing. Analytical comparisons of chemically equivalent materials to ST909 were also performed and correlated to bench scale test results. Characterization testing performed to evaluate chemically equivalent materials included: X-ray fluorescence, X-ray diffraction, and inductively coupled plasma-mass spectrometry (ICP-MS). Results are expounded upon and differences in gettering capability of alternate materials are highlighted.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2016

Inertially confined fusion plasmas dominated by alpha-particle self-heating [Alpha-particle self-heating dominated inertially confined fusion plasmas]

Alpha-particle self-heating, the process of deuterium–tritium fusion reaction products depositing their kinetic energy locally within a fusion reaction region and thus increasing the temperature in the reacting region, is essential for achieving ignition in a fusion system. Here, we report new inertial confinement fusion experiments where the alpha-particle heating of the plasma is dominant with the fusion yield produced exceeding the fusion yield from the work done on the fuel (pressure times volume change) by a factor of two or more. These experiments have achieved the highest yield (26 ± 0.5 kJ) and stagnation pressures ( &#8781 220 ± 40 Gbar) of any facility-based inertial confinement fusion experiments, although they are still short of the pressures required for ignition on the National Ignition Facility (~300–400 Gbar). Finally, these experiments put us in a new part of parameter space that has not been extensively studied so far because it lies between the no-alpha-particle-deposition regime and ignition.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2018

Hydrothermal discharge from the El Tatio basin, Atacama, Chile

El Tatio in northern Chile is one of the best-studied geothermal fields in South America. However, there remain open questions about the mass and energy budgets, water recharge rates and residence time in the subsurface, origin of dissolved solutes, and processes affecting the phase and chemical composition of groundwater and surface water. Here we measured and sampled surface manifestations of the geothermal system (geysers, perpetual spouters, mud pools/volcanoes, and non-eruptive hot springs) and meteoric water. From the isotopic composition we infer that the thermal water has a meteoric origin that is different from the composition of local meteoric water. The absence of detectable tritium in thermal waters indicates that most of the recharge occurred pre-1950. Boiling and steam separation from the deep reservoir appear to be the main subsurface processes affecting the thermal fluids. A large amount of heat is lost to the atmosphere by evaporation from surface water and by steam emitted from erupting geysers. Finally, using the chloride inventory method, we estimate thermal water discharge to be 218 to 234 L/s, and the advective heat flow to be 120 to 170 MW.

15 GEOTHERMAL ENERGY↗
DOE OSTI2018

Final Report For PDRD SR16009: Durable Water Splitting Using Thermochemical Cycles Of Nanostructured Metal Oxides

The Tritium Facility process for stripping tritium from the glovebox atmosphere currently relies on the use of hot magnesium beds (mag beds) to “crack” tritiated water, releasing hydrogen isotopes as a gas. This is a safe and effective process for recovering hydrogen isotopes from water for further processing. However, due to the thermodynamic properties of magnesium, mag beds are single use - once the magnesium metal is converted to magnesium oxide the bed must be replaced. Replacing mag beds is a labor-intensive and operationally complex task. In order to streamline operations and reduce costs, it is desirable to develop a reusable system that recovers hydrogen isotopes from water, while safely sequestering oxygen in a stable form. The following document reviews the results from the Plant Directed Research, Development, and Demonstration (PDRD) project “Durable Water Splitting using Thermochemical Cycles of Nanostructured Metal Oxides” (SR16009), which had the specific goal of developing a process for stripping tritium from the glovebox atmosphere that eliminates magnesium beds.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

THERMAL STRUCTURAL AND SHOCK EVENT EVALUATIONS OF THE FUELING PELLET INJECTION SYSTEM FOR ITER

The US is among seven partner nations in a collaborative effort to design, build, and demonstrate fusion’s ability to be a large-scale carbon free energy source. Each country has its own Domestic Agencies (DA) that contribute directly to the ITER project. US ITER, which is a DOE Office of Science project managed by Oak Ridge National Laboratory, is developing world class engineering solutions to the design, construction, and assembly of the burning plasma experiment that can demonstrate the scientific and technological feasibility of fusion. US ITER’s scope includes completing the preliminary and final design, qualifying materials and processes for manufacture and testing, executing manufacturing, and delivering the Fueling Pellet Injection System (FPIS) to the ITER site for assembly. The US-ITER FPIS system is designed to inject cryogenically frozen pellets of deuterium-tritium (D-T) into the plasma. The FPIS has two main functions: 1. Provide a steady supply of deuterium and tritium fuel, 2. Mitigating the impact of edge localized modes on the plasma facing components. The FPIS will have the capability to reside in three port cells within a tritium second barrier containment cask. Each pellet cask contains three flight tubes linking with penetrations on the torus cryopump housing and vacuum vessel (VV): two for the magnetic high field side (HFS) pellet injection and one for the magnetic low field side (LFS) pellet injection. This paper presents results of the thermal-structural analyses of the FPIS flight tube structural components when subjected to various loads such as electromagnetic (EM) loads, nuclear heating, seismic, and operational, and dynamic shock events. The resulting temperatures and stresses under combined conditions have been found to satisfy the design criteria to ensure safe and reliable operation of the FPIS flight tubes within the vacuum vessel.

Martinez, Oscar↗
DOE OSTI2014

Tritium Breeding Blanket for a Commercial Fusion Power Plant - A System Engineering Assessment

The goal of developing a new source of electric power based on fusion has been pursued for decades. If successful, future fusion power plants will help meet growing world-wide demand for electric power. A key feature and selling point for fusion is that its fuel supply is widely distributed globally and virtually inexhaustible. Current world-wide research on fusion energy is focused on the deuterium-tritium (DT for short) fusion reaction since it will be the easiest to achieve in terms of the conditions (e.g., temperature, density and confinement time of the DT fuel) required to produce net energy. Over the past decades countless studies have examined various concepts for TBBs for both magnetic fusion energy (MFE) and inertial fusion energy (IFE). At this time, the key organizations involved are government sponsored research organizations world-wide. The near-term focus of the MFE community is on the development of TBB mock-ups to be tested on the ITER tokamak currently under construction in Caderache France. TBB concepts for IFE tend to be different from MFE primarily due to significantly different operating conditions and constraints. This report focuses on longer-term commercial power plants where the key stakeholders include: electric utilities, plant owner and operator, manufacturer, regulators, utility customers, and in-plant subsystems including the heat transfer and conversion systems, fuel processing system, plant safety systems, and the monitoring control systems.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2017

A Tritium Management Tool for Light Water Detritiation - 17320

Kurion{sup R}, Inc. (Kurion) has developed the Modular Detritiation System (MDS{sup R}) based on the working principle of combined electrolysis catalytic exchange (CECE), offering a combination of efficiency, maturity, and cost among alternatives. The technology builds on proven heavy water solutions, and although is focused on light water, it can be adapted for heavy water detritiation. The application of MDS{sup R} is a modular approach and allows for a flexible design that is adaptable to a range of facility sizes and processing needs. This modular approach delivers a factory quality and plug-n-process system that utilizes commercially available equipment and minimizes impact to a facility's day-to-day operations. In addition to focusing on optimization of the MDS{sup R} technology, Kurion has strategies for a complete detritiation solution of which MDS{sup R} is a part of. The strategy, supported by modeling and numerical analysis, is a systematic approach to managing front-end pretreatment (if required), overall volume reduction, and amount and method of tritium retention. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Fission Product Yields from 232Th, 238U, and 235U Using 14 MeV Neutrons

Neutron-induced fission yield studies using deuterium-tritium fusion-produced 14 MeV neutrons have not yet directly measured fission yields from fission products with half-lives on the order of seconds (far from the line of nuclear stability). Fundamental data of this nature are important for improving and validating the current models of the nuclear fission process. Cyclic neutron activation analysis (CNAA) was performed on three actinide targets–thorium-oxide, depleted uranium metal, and highly enriched uranium metal–at the University of Michigan's Neutron Science Laboratory (UM-NSL) using a pneumatic system and Thermo-Scientific D711 accelerator-based fusion neutron generator. This was done to measure the fission yields of short-lived fission products and to examine the differences between the delayed fission product signatures of the three actinides. The measured data were compared against previously published results for 89Kr, -90, and -92 and 138Xe, -139, and -140. The average percent deviation of the measured values from the Evaluated Nuclear Data Files VII.1 (ENDF/B-VII.1) for thorium, depleted-uranium, and highly-enriched uranium were -10.2%, 4.5%, and -12.9%, respectively. In addition to the measurements of the six known fission products, 23 new fission yield measurements from 84As to 146La are presented.

Pierson, Bruce D.↗
DOE OSTI2014

Investigation of Radiation Cross-Talk Effects on Biological Dose Rates in Maintenance Areas Inside the ITER Tokamak - Paper 107

The intense plasma neutron source of ITER requires careful consideration of radiation safety during the design of the machine, its operation plan and licensing process. One important issue is that of occupational radiation exposure (ORE) during in-situ maintenance activities inside the cryostat, and particularly in the port inter-space area (PI) where up keeping of port systems should take place (diagnostics, tritium breeding test modules, heating antennas). ITER ports and port plugs (PP) become activated by neutron irradiation during plasma operation; from the subsequent decay and gamma emission of irradiated materials, the so-called shutdown dose rate field (SDR) appears. A project requirement of 100 uSv/h after 12 days exists for the SDR in the PI in order to limit exposure during maintenance and meet ORE regulatory limits. This requirement is particularly challenging for equatorial ports (EP) due to, among others, background neutron radiation and activation from/by neighbouring components (a.k.a. 'cross-talk'), and particularly the lower ports (LP). For a particular port inter-space, the SDR field is a combination of: (i) the gamma dose due to the activation of components local to the PI, typically referring to the PI itself plus the port plug, port duct and bellows, vessel extension and blankets in its vicinity; (ii) the gamma dose due to activation of components far away from the PI, a.k.a. the cross-talk, which typically refers to other ports, cryostat and rest of the vacuum vessel and blankets. The situation is complicated to some extent by the fact that both these components can arise from a combination of irradiation by local neutrons, i.e. those permeating through the vicinity of the port itself, and by cross-talk neutrons, i.e. those permeating via other paths. In simple terms, the so-called 'neutron cross-talk' refers to neutron leakage through other ports and components producing activation, decay gammas and dose locally in the PI of interest, whereas the 'gamma cross-talk' refers to activation and decay gammas in other ports producing dose in the PI from far away. The study reported here focuses on the evaluation of the effect of radiation cross-talk from a torus cryopump (TCP) LP located directly underneath a generic diagnostics equatorial port plug (DEPP) in the equatorial port itself. This is considered the worst case scenario, as the TCP-LP has a large aperture and has to be mostly empty in order to meet TCP pumping requirements. For the reference case and R1, the neutron cross-talk from the LP activating the EP is responsible for 109 uSv/h. In addition, a gamma cross-talk of 51 uSv/h from LP activation has been found. The total cross-talk contribution is therefore responsible for 160 uSv/h of SDR at EP PI. The local SDR contribution, i.e. that from activation of the EP with 'local-only' neutrons, is 166 uSv/h. The total SDR at R1 is 360 uSv/h, including some minor contributions from the UP, lateral EPs, cryostat and bio-shield. The total SDR at R2 is 349 uSv/hr, with a cross-talk contribution of 203 uSv/h. In R2 the gamma cross-talk is more intense, as it is closer to the highly activated TCPH. These results illustrate how the SDR field produced by radiation cross-talk through a TCP-LP is comparable in magnitude with that produced by neutron leakage and activation local to the EP. Shielding options are currently under consideration. Local gamma shielding at the EP PI is found to be effective in reducing gamma cross-talk from the LP and TCPH, and also in reducing the local contribution from the port duct and bellows. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2017

Evaluation of a Zirconium Recycle Scrubber System

A hot-cell demonstration of the zirconium recycle process is planned as part of the Materials Recovery and Waste Forms Development (MRWFD) campaign. The process treats Zircaloy® cladding recovered from used nuclear fuel with chlorine gas to recover the zirconium as volatile ZrCl4. This releases radioactive tritium trapped in the alloy, converting it to volatile tritium chloride (TCl). To meet regulatory requirements governing radioactive emissions from nuclear fuel treatment operations, the capture and retention of a portion of this TCl may be required prior to discharge of the off-gas stream to the environment. In addition to demonstrating tritium removal from a synthetic zirconium recycle off-gas stream, the recovery and quantification of tritium may refine estimates of the amount of tritium present in the Zircaloy cladding of used nuclear fuel. To support these objectives, a bubbler-type scrubber was fabricated to remove the TCl from the zirconium recycle off-gas stream. The scrubber was fabricated from glass and polymer components that are resistant to chlorine and hydrochloric acid solutions. Because of concerns that the scrubber efficiency is not quantitative, tests were performed using DCl as a stand-in to experimentally measure the scrubbing efficiency of this unit. Scrubbing efficiency was ~108% ± 3% with water as the scrubber solution. Variations were noted when 1 M NaOH scrub solution was used, values ranged from 64% to 130%. The reason for the variations is not known. It is recommended that the equipment be operated with water as the scrubbing solution. Scrubbing efficiency is estimated at 100%.

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

PDRD Project 19039 - All metal vacuum scroll pump development for tritium

Currently the United States relies on a single source, foreign company, Eumeca (formerly Normetex) for the supply of an all metal vacuum scroll pump to process tritium. This “one of a kind”, all metal, vacuum scroll pump utilizes no “wetted”, polymeric materials that degrade prematurely when exposed to tritium. U.S. based vacuum scroll pump suppliers typically utilize a polymer, Teflon®, as a tip seal material at the edge of the scroll impellers for proper pump function. Teflon® is known to degrade within a relatively short time frame (approximately 1 year) when continually exposed to tritium gas. This degradation results in “flaking” of the Teflon® material which results in particulate contamination of the process gas stream. While filtration of the particulate is possible, doing so requires additional maintenance which is costly in a glovebox environment. In addition, filtration can degrade pumping system performance as particulate collects in the filter media over time. To date, all of the well established U.S and European based vacuum scroll pump manufacturers have been uninterested in developing an all metal vacuum scroll pump that does not use a polymeric scroll tip seal. This position is taken as the vacuum scroll pump manufacturers do not see a large enough market to make it profitable for them to develop and market an all metal vacuum scroll pump. In the mid 2000s, one company, Air Squared, Inc., a small Original Equipment Manufacturer (OEM) specializing in design and fabrication of custom scroll vacuum pumps and compressors, was willing to pursue the effort. The first design effort incorporated Vespel® as a polymeric tip seal material. Results of this effort led to realizing the affect that the Met-Bel MB-601 metal bellows pump has on performance when used as a backing pump for an all metal vacuum scroll pump without a scroll tip seal. As a result, Air Squared subsequently designed an all metal vacuum scroll pump without a tip seal material for evaluation. Although the initial, all metal version of the Air Squared pump was found deficient in regard to scroll impeller stiffness, which resulted in scroll contact, general design and pump performance was verified to be adequate. SRS PDRD Project 19039 was launched in 2019 as a 3rd attempt with Air Squared, Inc. in regard to development of an all metal vacuum scroll pump for consideration as an equivalent to the Eumeca/ Normetex pump. The effort specified 300 series stainless steel for the scroll impellers, all metal sealing, and design to meet tritium throughput and ultimate vacuum requirements. In addition, a 60 psig design pressure rating requirement was specified for the pump containment boundary. The 3rd generation Air Squared pump has been designed, fabricated, and has received preliminary testing. Preliminary testing has indicated proper leak tightness and pump performance. However, work continues regarding development of a pumping chamber containment bellows that can be subjected to 66 psig without damage to the bellows. Work will continue in 2020 and 2021 to pursue a pumping chamber containment bellows that meets the pressure rating requirement.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Evaluation of the Waste Streams from Management of Reprocessing Off-Gas Streams - 18483

Nuclear fuel reprocessing is being considered in the United States as part of an effort to help meet US energy needs and as a means to improve waste management. During the chemical processing of UNF to recover the valuable components for reuse, a number of volatile radionuclides will be released into the off-gas streams of the reprocessing plant. These volatile radionuclides include H-3, C-14, Kr-85, and I-129. Off-gas abatement systems for the volatile radionuclides will be required to ensure that air emissions from a UNF reprocessing facility can meet or exceed emission requirements set forth by the US Environmental Protection Agency and the US Nuclear Regulatory Commission. An initial engineering evaluation of offgas abatement systems was completed recently. The evaluation addressed the designs of the head-end offgas treatment system, which typically includes both the shear off-gas and the dissolver off-gas and vessel off-gas treatment system. Based on this study, the waste volumes for each of the four primary volatile radionuclides (H-3, C-14, Kr-85, and I-129) were estimated. Included in these estimates were the nonradioactive isotopes and the 'tramp' elements found in air (e.g., Kr and C-12 as CO{sub 2}) and process chemicals (e.g., I and Cl). Because the final waste form for these radionuclides has not been selected, estimated volumes were calculated for various alternative waste forms as appropriate. The resulting waste volume for I was found to be highly dependent on the final waste form selected and increased by a factor of 2 to 4 by tramp halogens that would be released into the off-gas stream from the process chemicals. The volume of tritiated water from the fuel was quite small, but the estimated volume of the final waste form depended on the tritium management approach employed within the reprocessing plant. The Kr waste volume was calculated assuming pure Kr (i.e., assuming complete separation of Xe and Ar). For the case of no Xe separation, the estimated waste volume increased by factor of ∼10. Argon from the air, if not removed, increased the estimated volume by a factor of 425. The estimated volume of C-14 waste was dominated by the CO{sub 2} from the air leaking into the processing cell. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Real-Time Detection of Hydrogen and Ammonia Isotopologues for Impurity Removal and Recovery of Tritium

To accommodate gas measurements for impurity removal and recovery of tritium, a silver-coated optical or waveguide is employed for collecting Raman scattered signals to determine relative hydrogen and ammonia isotopologue populations in real time. The data and results presented here demonstrate an analytical methodology for the analysis of four ammonia and three hydrogen isotopologues in a hydrogen–deuterium exchange reaction by gas phase Raman spectroscopy. Standard chemometric modeling techniques effectively unravel the signatures of the isotopologues involved observed here; however, a sophisticated quantum chemical approach supports the spectral assignments. An interpretation of the data presented here can emphasize the practicality and reliability of the gaseous monitoring system in complex chemical environments for the hydrogen fuel economy as well as the more distant energy source from a facility that handles tritium. There are still considerable concerns about the measurement of tritium in isotope separation and radiological impurities from gas processing. A common impurity in gas processing is ammonia, which can form readily in the presence of nitrogen and tritium. Substituted ammonia (NQ 3 ), where Q = H, D, or T, is traditionally removed through getters or diffusers along with other non-hydrogen contaminants. A preferable analytical approach is noninvasive and can be deployed for real-time process evaluation in radiological environments.

Ammonia↗
DOE OSTI2015

Replacement of tritiated water from irradiated fuel storage bay

Recently, AECL developed a novel method to reduce tritium emissions (to groundwater) and personnel doses at the NRU (National Research Universal) reactor irradiated fuel storage bay (also known as rod or spent fuel bay) through a water swap process. The light water in the fuel bay had built up tritium that had been transferred from the heavy water moderator through normal fuel transfers. The major advantage of the thermal stratification method was that a very effective tritium reduction could be achieved by swapping a minimal volume of bay water and warm tritiated water would be skimmed off the bay surface. A demonstration of the method was done that involved Computational Fluid Dynamics (CFD) modeling of the swap process and a test program that showed excellent agreement with model prediction for the effective removal of almost all the tritium with a minimal water volume. Building on the successful demonstration, AECL fabricated, installed, commissioned and operated a full-scale system to perform a water swap. This full-scale water swap operation achieved a tritium removal efficiency of about 96%.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2016

Migration and Distribution of Natural Organic Matter Injected into Subsurface Systems - 16523

The Savannah River Site (SRS) was one of the major U.S. Department of Energy facilities that produced large amounts of radioactive and hazardous waste during the Cold War era, a legacy of material production for defense purposes. The unlined F-Area seepage basins received approximately 1.8 billion gallons of acidic waste solutions containing low-level radioactivity from numerous isotopes generated by uranium slug and irradiated fuel processing. Sufficient quantities of uranium isotopes, I-129, Tc-99, and tritium migrated into the groundwater, creating an acidic plume with a pH between 3 and 5.5. Humic substances (HS) are complex heterogeneous mixtures of poly-dispersed materials formed by biochemical and chemical reactions during the decay and transformation of plant and microbial remains. They are known for their excellent binding capacity for metals, making them a strong candidate for remediation efforts to reduce the mobility of uranium (VI) in the subsurface. Studies showed that HS are an important ion exchange and metal-complexing ligand, carrying a large number of functional groups with high complexing capacity that can greatly affect the mobility behavior of actinides in natural systems. Hence, it is a potential in situ amendment for treating groundwater contaminated with these constituents. Column experiments were conducted to address the complexities of HS in flow-through porous media representative of the acidic aquifer at the SRS F-Area Seepage Basins. The humate source used, Huma-K, is high in humic and fulvic compounds. The column studies were done using the soil obtained from SRS's FAW-1 core-hole at a depth interval of 60'-70'. A bromide tracer test was performed to obtain the pore volume (PV), variance and Peclet number. The columns were saturated with an artificial groundwater solution prepared to mimic the groundwater at SRS that was pH adjusted to 3.5 and 5 until the effluent pH reached equilibrium. Approximately one (1) PV of 5000 ppm of Huma-K was injected into the columns at 2 ml/min to study the sorption of Huma-K on the sediment. Desorption of Huma-K was studied by injecting approximately four (4) PVs of the pH adjusted artificial groundwater solution. A faster increase in pH was observed in column 1 (saturated with pH 3.5 AGW) compared to column 2 (saturated with pH 5 AGW). After injection of 2 PV of AGW, the pH of the columns reached 6.5 and 7, respectively, for columns 1 and 2. After injecting an additional PV of AGW, the pH remained relatively stable. With an increase in pH from 3.5 to 5.0, the overall retention of HA increased from 461 mg/kg for the column with pH 3.5 to 642 mg/kg for the column with pH 5. This suggests that retention is not due to adsorption alone, but adsorption coupled with precipitation and re-dissolution of Huma-K as it flows through the columns is a possible explanation for greater retention at pH 5. This negates the use of a simple Langmuir isotherm to explain humate behavior in the acidic environment of column 1. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗