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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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At least 145 records · Page 8
All sourcesNASA NTRSDOE OSTI
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 the 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 tool-sets 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. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Removal of water and iodine by solid sorbents: adsorption isotherms and kinetics

Tritium and iodine-129 are two major radioactive elements that are present in off-gases from spent fuel reprocessing plants. Adsorption by solid sorbents is the state-of-the-art technique for removal of these species from off-gases. Modeling and simulating adsorption processes require accurate adsorption equilibrium and kinetic data to permit reasonable estimates of process parameters. We have developed a continuous flow single-pellet adsorption system to gather accurate adsorption equilibrium and kinetic data for adsorption of water by molecular sieve 3A and for adsorption of iodine by silver exchanged mordenite. In this paper, the design of the water and iodine adsorption experimental systems are briefly described and results of water adsorption experiments are presented and discussed. Water uptake curves are fitted with the linear-driving force (LDF) model and the shrinking-core model to determine kinetic parameters. It is shown that the kinetics of water adsorption on zeolite 3A under current experimental conditions is controlled by both the external film resistance and the macro-pore diffusion and can be predicted by both the LDF model and the shrinking-core model with the former one performing slightly better. Preliminary results from iodine adsorption experiments will be presented in the conference.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Evaluating T 2 Conversion to Tritiated Water under Ambient Conditions - Case Study

In 2015, an incident released approximately 40 Ci of T 2 gas directly into the Tritium Exhaust System. Data from a bubbler system that monitored the stack effluent during the time period encompassing the accident, from 9 days prior through approximately 26 hours following the release, indicated that approximately 0.25% of the total accumulated tritium gas was in the form of tritiated water; however this value does not account for sources of tritium exhaust from other building operations and processes during the 9 days prior to this incident. Further analysis of the bubbler data around this time period considered the 9-day background contributions and shows that the actual fraction of the tritium that was released as tritiated water vapor (during and within 26 hours after the release) was likely lower than 0.1%.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Continuous flow system for controlling phases separation near λ transition

As demands on 3He are increasing and conventional 3He production through tritium decay is decreasing, alternative 3He production methods are becoming economically viable. One such possibility is to use entropy filters for extraction of the 3He isotope from natural gas. According to the phase diagram of the 3He, its solidification is impossible by only lowering of the temperature. Hence during the cooling process at stable pressure we can reach λ-point and pass to the special phase - He II. The total density of HeII is a sum of the two phases: normal the superfluid ones. It is possible to separate these two phases with an entropy filter - the barrier for the classically-behaving normal phase. This barrier can also be used to separate the two main isotopes of He: 4He and 3He, because at temperatures close to the 4He-λ-point the 3He isotope is part of the normal phase. The paper presents continuous flow schemes of different separation methods of 3He from helium commodity coming from natural gas cryogenic processing. An overall thermodynamic efficiency of the 3He/4He separation process is presented. A simplified model of continuous flow HeI -HeII recuperative heat exchanger is given. Ceramic and carbon porous plugs have been tested in entropy filter applications.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2018

Off-Gas Treatment: Evaluation of Nano-structured Sorbents for Selective Removal of Contaminants

Nuclear energy has practically unlimited potential to satisfy world’s energy needs for the foreseeable future. However, a comprehensive and reliable solution must be devised to address the key issues related to nuclear waste management in order to develop nuclear energy in a safe and responsible manner. Capture and immobilization of volatile radionuclides from nuclear operations is an essential component of an integrated nuclear waste management system. The majority of emissions occur during the treatment of the used nuclear fuel (UNF) as it is chopped and dissolved in the boiling nitric acid for subsequent extraction steps. The radionuclides contained in the off-gas include 129 I, 85 Kr, tritium (3H) and 14 C. Several alternative technologies have been investigated, with effective adsorption based processes holding the most potential for controlling these emissions, which is highly desirable for the development of the advanced fuel cycle. Proposed project is aimed at developing using a nanosorbent-based process for the capture and immobilization of the radionuclides of interest.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Effect of Minerals on the Removal of Uranium (VI) in the Presence of Humic Acid and Colloidal Silica - 19406

Uranium is one of the key contaminants of concern in groundwater as a result of past nuclear processing activities at the Department of Energy (DOE) facilities such as the Savannah River Site (SRS) in Aiken, South Carolina. SRS was one of the major nuclear processing facilities during the Cold War where plutonium was produced. Beginning with the implementation of the environmental cleanup program in 1981, SRS has become a hazardous waste management facility responsible for the storage and remediation of contaminated soil and groundwater from radionuclides. Approximately 1.8 billion gallons of acidic waste containing radionuclides and dissolved heavy metals was disposed in the F/H Area seepage basins. This led to the unintentional creation of highly contaminated groundwater plumes, consisting of radionuclides with an acidic pH between 3 and 5.5. The acidity of the plumes contributed to the mobility of several constituents of concern (COC) such as tritium ({sup 3}H), uranium-238 ({sup 238}U), iodine-129 ({sup 129}I), and strontium-90 ({sup 90}Sr) for the F-Area plume and {sup 3}H, {sup 90}Sr and mercury for the H-Area plume. This investigation will focus on uranium in the oxidation state +6 (U(VI)), which is a key contaminant of concern in the groundwater plume of the FArea. Initially, the pump-treat-reinject system was implemented from 1997 to 2003 to remove contaminants from polluted groundwater. Downgrade groundwater within the system would be pumped to the water treatment facility and re-injected upgrade within the aquifer. The effectiveness and sustainability of this process diminished over time. Therefore, it was discontinued and replaced with a funnel-and-gate process in 2004. This new process injects sodium hydroxide directly into the gates of the F-Area groundwater to effectively raise pH levels. By raising the pH of the groundwater, a treatment zone is created by reversing the acidic nature of the contaminated sediments and producing a negative net charge on the surface of sediment particles. As a result, the adsorption of cationic contaminants is expected to be enhanced. So far, this process has resulted in a decrease of {sup 90}Sr and {sup 238}U concentrations, though the concentration of iodine has been unaffected by this treatment. The solution used for the injections contains a high carbonate alkalinity in order to overcome the surface acidic conditions and natural partitions in the groundwater system. To maintain the neutral pH in the treatment zone, systematic injections are required. The continuous use of high concentrations of a carbonate solution to raise pH could re-mobilize uranium previously adsorbed within the treatment zone, though this has not been observed in the monitoring data. Humic substances (HS) are major components of soil organic matter. HS are polyfunctional organic macromolecules formed by the chemo-microbiological decomposition of biomass or dead organic matter. These substances are usually divided into three main fractions: humin (insoluble at all pH values), humic acid (soluble at pH greater than 3.5), and fulvic acid (soluble at all pH values). Humic acid carries a large number of functional groups such as aromatic rings, carboxyl groups, phenols, aliphatic chains, etc. Humic acid is an important ion exchange and metal complexing ligand with a high complexation capacity. The ability of humic acid to chemically bind to metals influences in their migration behavior and fate in natural systems. Various studies have suggested that the retention of U(VI) via sorption in the presence of humic acid is a complex process. For instance, humic acid can form an organic coating at the surface of minerals, which can enhance the sequestration of metals. Ivanov et al. (2012) studied the sorption of U(VI) sorption onto bentonite in the presence and absence of humic acid. The study found that uranium sorption in the presence of humic acid was enhanced at low pH (below 3.8) while at moderate pH (3.8 and 6.5) uranium sorption was reduced compared to the absence of humic acid. At high pH (7 and 9), uranium sorption was reduced for both the presence and absence of humic acid. Also, Krepelova et al. (2007) not only found that U(VI) sorption onto kaolinite was influenced by pH, U(VI) concentration, the presence of inorganic carbon species, and humic acid but also that U(VI) preferred to be adsorbed onto kaolinite as a uranyl-humate complex. This investigation analyzed the mechanisms of the synergetic interactions of humic acid and colloidal silica on the removal of uranium in the presence of pure minerals such as quartz, kaolinite and goethite which are characteristic of the F/H area. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Transmutation Scoping Studies for a Chloride Molten Salt Reactor

Over the past few years, there has been strong renewed interest from private industry, mostly from start-up enterprises, in molten salt reactor (MSR) technologies because of the unique properties of this class of reactors. These are reactors in which the fuel is homogeneously mixed with the coolant in the form of liquid salts and is circulated continuously into and out of the active core region with on-line fuel management, salt treatment, and salt processing. In response to such wide-spread interest, Argonne National Laboratory is expanding its well-established reactor modelling and simulation expertise and infrastructure to enable detailed analysis and design of MSRs. The tools being developed are able to simulate the continuous fuel flow, the complex on-line fuel management and elemental removal processes (e.g., fission product removal) using depletion steps representative of a real MSR system. Leveraging these capabilities, a parametric study on the transmutation performance of a simplified actinide-burning MSR concept that uses a chloride-based salt was performed. This type of salt has attracted attention over the more commonly discussed fluoride-based salts since no tritium is produced as a result of irradiation and it is compatible with a fast neutron spectrum. The studies discussed in this paper examine the performance of a burner MSR design with a fixed core size and power density over a range of possible fuel salt molar ratios with NaCl-MgCl2 as the carrier salt. The intent is to quantify the impact on the required transuranics content of the make-up fuel, the actinide transmutation rates, and other performance characteristics for typical burner MSR designs.

Heidet, Florent↗
DOE OSTI2022

2021 Annual Site Environmental Report

This report provides the U.S. Department of Energy (DOE) and the public with information on the level of radioactive and non-radioactive pollutants (if any) that are added to the environment as a result of Princeton Plasma Physics Laboratory’s (PPPL) operations. The results of PPPL’s 2021 environmental surveillance and monitoring program are presented and discussed. The report also summarizes environmental initiatives, assessments, and community involvement programs that were undertaken in 2021. PPPL’s on-site operations were significantly curtailed in 2021 due to the global coronavirus pandemic. PPPL has engaged in fusion energy research since 1951. The Laboratory’s mission is to develop the scientific knowledge and advanced engineering to enable fusion to power the U.S. and the world, and to developing the understanding of plasmas from the nano- to the astrophysical scale. PPPL’s primary experiment, the National Spherical Torus Experiment-Upgrade (NSTX-U) is a collaboration among national laboratories, universities, and national and international research institutions and is a major element in the US Fusion Energy Sciences Program. Its design tests the physics principles of spherical torus (ST) plasmas, playing an important role in the development of smaller, more economical fusion reactors. Due to previous operational issues, NSTX-U did not operate in 2021. PPPL is engaged in a project to replace key NSTX-U components and systems to enable operation of this international fusion user facility. In 2021, PPPL’s radiological environmental monitoring program measured tritium in the air at the NSTX-U Stack and at onsite sampling stations. Using highly sensitive air monitors, PPPL is capable of detecting small changes in the ambient levels of tritium. The operation of an in-stack monitor located on D-site is used to demonstrate compliance with the National Emission Standard for Hazardous Air Pollutants (NESHAPs) regulations. Also included in PPPL’s radiological environmental monitoring program, are water monitoring – ground, surface, and waste waters. PPPL’s radiological monitoring program characterized the background levels of tritium in the environment and those data are presented in this report. Ground water monitoring continued under New Jersey Department of Environmental Protection’s (NJDEP) Site Remediation Program regulations. PPPL monitored for non-radiological contaminants, mainly volatile organic compounds (components of common degreasing solvents). In 2021, PPPL complied with permit limits for surface and sanitary discharges. PPPL was honored with awards for EPEAT-certified electronics purchasing and use of peracetic acid as an alternative water treatment chemical in its non-potable process water systems an NJDEP recycling award.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2024

Radiation Monitoring for Volatilized Zinc Contamination Using Gamma-Ray Imaging and Spectroscopy

Gamma-ray imaging is a tool that has grown in importance in the applications of non-destructive assay (NDA) for radioactive survey and analysis of nuclear facilities. Imaging techniques have shown great promise in providing valuable information involving radioactive waste management and contamination prevention. For the application studied in this work, 65 Zn has been identified as a radioactive contaminant during tritium extraction. Due to the volatile nature of 65 Zn under the pressure and temperature changes during extraction operations, 65 Zn can easily travel through components of the extraction system as vapor, making it difficult to trap. Previous research involving the development of a filtration system showed that the 65 Zn can be trapped, mitigating product contamination. However, during the extraction process, direct analysis of the equipment to confirm that zinc contamination is trapped in the filter and has not spread to other components is impractical. Here, in this situation, the need to assay the location of the contamination with little-to-no interference with operations is vital. In this work, we demonstrate the use of a commercialized 3D position-sensitive CdZnTe (CZT) gamma-ray imaging spectrometer to provide analysis of the 65 Zn contamination. Onsite measurements during an extraction process are studied to assess the location and migration of the 65 Zn. The results obtained from real-time glovebox monitoring demonstrate the feasibility of gamma-ray imaging for localizing the contamination and providing a preliminary qualitative assessment that is intended to be used in future work quantifying the contamination build-up and activity over time.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2015

Adsorption characteristics of water vapor on gear-pellet and honeycomb-pellet types of adsorbents containing A-type zeolite

It is necessary to recover or process tritiated species that are extensively coexistent in nuclear fusion installations. A conventional way to recover tritium release to atmosphere is catalytic oxidation of tritiated species and adsorption of tritiated water vapor on adsorbents with high surface areas. Therefore, new adsorbents with low pressure loss and high surface areas need to be developed and utilized for such large-scale adsorption systems. In this study, attention was focused on new adsorbents, which are gear-type pellet MS5A adsorbent, gear-type pellet MS4A adsorbent and honeycomb-type pellet MS5A adsorbent. The adsorption characteristics of the new adsorbent were comparatively studied with conventional type of adsorbents (pellet-type MS5A adsorbent and pebble-type MS5A adsorbent), in terms of adsorption capacity, pressure loss and adsorption rate. It was found that the adsorption capacity of water vapor on the gear-type adsorbents is higher than that on a honeycomb-type adsorbent. The experimental breakthrough curves indicate that the adsorption rates of water vapor on gear-type and honeycomb-type adsorbents are smaller than that on conventional type adsorbents. Various adsorption models were also tested to correlate the experimental isotherms. It was found that the Langmuir-Freundlich model could properly correlate the experimental adsorption isotherms.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2015

Conceptual design and optimization for JET water detritiation system cryo-distillation facility

The aim of the Exhaust Detritiation System (EDS) of the JET Active Gas Handling System (AGHS) is to convert all Q-based species (Q{sub 2}, Q-hydrocarbons) into Q{sub 2}O (Q being indifferently H, D or T) which is then trapped on molecular sieve beds (MSB). Regenerating the saturated MSBs leads to the production of tritiated water which is stored in Briggs drums. An alternative disposal solution to offsite shipping, is to process the tritiated water onsite via the implementation of a Water Detritiation System (WDS) based, in part, on the combination of an electrolyser and a cryo-distillation (CD) facility. The CD system will separate a Q{sub 2} mixture into a de-tritiated hydrogen stream for safe release and a tritiated stream for further processing on existing AGHS subsystems. A sensitivity study of the Souers' model using the simulation program ProSimPlus (edited by ProSim S.A.) has then been undertaken in order to perform an optimised dimensioning of the cryo-distillation system in terms of available cooling technologies, cost of investment, cost of operations, process performance and safety. (authors)

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2015

Decommissioning of the NPP A-1 heavy water management system - 15121

The paper deals with experience and techniques in the application of decontamination technology and remotely controlled robotic devices for the decontamination and dismantling of the heavy water management system during undergoing decommissioning process of the A-1 NPP as well as treatment of arising liquid waste All of these activities are characterized by high level of radioactivity and contamination. During operation of the A-1 NPP the heavy water management system fulfilled several functions: (1) cooling of the moderator, (2) topping up of the moderator in the active zone, (3) drainage, feed and storage of heavy water, (4) capture of leaked heavy water. Closely associated with heavy water management were systems for separation and isotopic purification of heavy water (distilling station, removing organic impurities station, isotopic purification station). The NPP heavy water system is located inside the power block building in several rooms where the inner surface contamination is up to the level of 10 E4 Bq.cm{sup -2}, dose rate up to 15,0 mGy.h{sup -1} and the feeding pipeline contained liquid RAW with high tritium content. Pre-dismantling decontamination of the heavy water management system and associated systems had to be performed with the intention of improving the radiation situation, i.e. reducing contamination of internal surface and dose rates in the vicinity of the equipment. Various sophisticated procedures such as retrieval of sludge using remotely operated devices and loop chemical decontamination of internal surfaces were applied. The generated radioactive waste such as spent decontamination solutions and sludge were immediately solidified on-site into a geo-polymer matrix using a mobile conditioning facility. For dismantling and fragmentation of some devices, mobile robotic systems MT80 and MT15 were applied which had been developed, designed and constructed as general-purpose decommissioning equipment. Special tooling was developed for application with the robots, such as hydraulic shears, circular saw, reciprocating saw, circular pipe cutter and a system for quick tool-change without direct intervention of the operators. Most of the operations are remotely controlled on basis of visual information from four cameras, with consistent radiation protection of the operators. The implementation of partial decommissioning tasks in such hard and unfavorable radiation conditions (moreover accompanied by generation of specific high-content alpha radwaste) has introduced in practice several new unique decontamination and dismantling techniques and new procedures for the effective on-site processing of generated secondary radioactive waste, including sludge and sorbents. These advantageous techniques, equipment and experience will be exploited in the continuation of the A-1 NPP decommissioning process which has been organized according to distinctively licensed phases and are also being applied in the decommissioning of other NPP in Slovakia as well as abroad. (authors)

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

Investigating the application of Kalman Filters for real-time accountancy in fusion fuel cycles

Here tritium accountancy in the fusion fuel cycle is a significant concern for the operation of commercial devices. It is expected that a limit on the maximum amount of tritium inventory in the system will be implemented, meaning that accountancy of the tritium inventory in the fuel cycle will need to be as accurate as possible. This is difficult since not all locations along the fuel cycle can benefit from the implementation of a tritium accountancy sensor and measurements will inherently contain error. Commercial fusion plants will also operate continuously, challenging current accountancy techniques that rely on static processes in well-controlled environments. In this paper a simple fusion fuel cycle concept is defined and the tritium inventory over time of each component is modelled using the Euler Approximation of a series of differential Equations in Python. This information is then used to simulate sensor measurements at specific points on the fuel cycle and then passed through a Kalman Filter (KF) to improve the accuracy of the true measurement of the sensors.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2019

Tritium Extraction from Water

Tritiated water production is ubiquitous in facilities that handle tritium gas. Sources range from decontamination efforts, to the deliberate conversion of elemental tritium to tritiated water in processes that strive to reduce emissions to the environment, to gaseous effluents to the environment. At low concentrations, ranging from a few μCi/L to mCi/L, high throughputs are required to process the high-volume, low-activity water. Combined Electrolysis Catalytic Exchange (CECE) shows promise by offering high throughput, reliability, economic viability, and facile coupling to isotopic separation systems if necessary. This paper will discuss the features of two production-scale CECE facilities: a 7-m 3 /h throughput system that uses an alkaline electrolysis cell and a 21-m 3 /h throughput system that uses a proton exchange membrane (PEM) electrolysis cell. Finally, the former is in service and has been modified to improve reliability; the latter is in the initial stages of commissioning.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2015

Management of tritium European Spallation Source

The European Spallation Source (ESS) will produce tritium via spallation and activation processes during operational activities. Within the location of ESS facility in Lund, Sweden site it is mandatory to demonstrate that the management strategy of the produced tritium ensures the compliance with the country regulation criteria. The aim of this paper is to give an overview of the different aspects of the tritium management in ESS facility. Besides the design parameter study of the helium coolant purification system of the target the consequences of the tritium releasing into the environment were also analyzed. Calculations show that the annual release of tritium during the normal operations represents a small fraction from the estimated total dose. However, more refined calculations of migration of activated-groundwater should be performed for higher hydraulic conductivities, with the availability of the results on soil examinations. With the assumption of 100% release of tritium to the atmosphere during the occurring of the extreme accidents, it was found as well that the total dose complies with the constraint. (authors)

43 PARTICLE ACCELERATORS↗
DOE OSTI2022

Chemical Thermodynamic Modeling of Molten Salts to Support Off-Gas Abatement Systems

The reprocessing of used nuclear fuel by any means will liberate gaseous fission products, such as hydrogen ( 3 H), carbon ( 14 C), noble gases ( 85 Kr), and halogens ( 129 I), from the irradiated fuel. These elements will distribute through chemical processing operations and partition into process off-gas streams, and the specific volatile release fractions will be dictated by the chemical and physical properties of the system. A recent assessment found that there were significant knowledge gaps regarding the release quantities of volatile radionuclides from individual unit operations. These knowledge gaps limited the ability to determine what dedicated off-gas treatment technologies could be required for electrochemical-based reprocessing facilities. Unfortunately, experimental efforts to quantify release fractions are limited by the challenges associated with performing experiments using irradiated fuel. This report documents preliminary thermodynamic predictions of iodine and tritium release from chloride-based molten salts as part of an effort to better direct resources toward those experiments (both simulant and irradiated) that will be of the greatest impact. It was predicted that less than 0.5% of tritium was expected to be released and that nearly all of that amount would be released as H 2 . Thermochemical data for hydrogen (H 2 ) are of high fidelity, and no additional validation is recommended. Less than 0.05% of iodine was predicted to be released, with the primary volatile species being Cs 2 I 2 . Unlike the tritium predictions, the data underpinning the iodine release predictions are of low quality. It is recommended that a limited experimental program be dedicated to expanding the current physical property and thermodynamic property data for iodine in electrochemical processing and molten salt conditions, including vapor pressure measurements for the dimerized species predicted to comprise the majority of iodine release. Validating and improving the thermodynamic properties used in predictive modeling can reduce the need for expensive testing with irradiated fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

A multi-channel x-ray temporal diagnostic for measurement of time-resolved electron temperature in cryogenic deuterium–tritium implosions at OMEGA

Electron-temperature (T e ) measurements in implosions provide valuable diagnostic information, as Te is unaffected by residual flows and other non-thermal effects unlike ion temperature inferred from a fusion product spectrum. In OMEGA cryogenic implosions, measurement of T e (t) can be used to investigate effects related to time-resolved hot-spot energy balance. The proposed diagnostic utilizes five fast-rise (~15 ps) scintillator channels with distinct x-ray filtering. Furthermore, titanium and stepped aluminum filtering were chosen to maximize detector sensitivity in the 10 keV–20 keV range, as it has been shown that these x rays have similar density and temperature weighting to the emitted deuterium–tritium fusion neutrons. Initial data collected using a prototype nosecone on the existing neutron temporal diagnostic demonstrate the validity of this diagnostic technique. The proposed system will be capable of measuring spatially integrated T e (t) with 20 ps time resolution and <10% uncertainty at peak emission in cryogenic DT implosions.

47 OTHER INSTRUMENTATION↗
DOE OSTI2024

Results of Speciation Determination Measurements of Insulating Sulfur Hexafluoride Gas

Under the direction of the NNSA NA-231 Mo-99 Program, SRNL has provided technical assistance to U.S. companies under cooperative agreements. One such area of technical assistance has been for the recommendation for contamination management methods in the eventuality that tritium contaminates the insulating gas of a high-voltage ion source used in the accelerator-based process developed by SHINE Medical Technologies to produce Mo-99. The SHINE accelerator process uses tritium as the target of a deuterium ion beam, where the ion beam is produced within a high-voltage ion source insulated with sulfur hexafluoride (SF 6 ), and is separated from the tritium target by an extensive pump train. Normal operating conditions preclude the conditions necessary for contamination of the insulating gas, however certain atypical conditions could result in a tritium contamination event. Accelerator based processes for medical isotope production represent an anomalous tritium contamination challenge due to the presence of SF 6 . SF 6 decomposes through normal use as an insulator gas, and the exact speciation of the decomposition byproducts depends on a variety of factors. This creates a scenario where the anticipation of species present in the insulating gas mixture is difficult. In turn, the fate of tritium within such a chemical system is unclear. This document discusses the results of measurements made in-situ of the SF 6 insulating gas in use within a SHINE Medical Technology’s high-voltage ion source. This was done in an effort to determine the chemical speciation of the gas, allowing for the recommendation of management strategies in the event that tritium contaminates the insulating gas. The results of the measurements reported in this work suggest that tritium contamination in the insulating gas may result in two chemical forms of tritium in the system, tritiated water and molecular tritium. These two forms of tritium would be captured by two methods. Tritiated water may be captured and removed from the SF 6 by an adsorbent canister already present within the pressure vessel that houses the insulating gas while in use. It may also be separated from the SF 6 during routine purification of the SF 6 , which is performed by SHINE personnel routinely. The purification cart will perform all non-embedded systems cleanup necessary to remediate tritiated species in the system. The purification cart passes the insulating gas through several filters, followed by the liquefication of the SF 6 . The filters will separate tritiated water from the SF 6 at this stage. It is expected that molecular tritium will accumulate in the gas-phase during the liquification step, which is separated and stored. Tritium may be recovered from this mixture through more conventional hydrogen purification methods such as a palladium diffuser.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗