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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 73 records · Page 4
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Progress of GE development of accident tolerant fuel FeCrAl cladding

The Accident Tolerant Fuel (ATF) concept proposed by General Electric (GE) utilizes a FeCrAl alloy and current uranium dioxide (UO{sub 2}) fuel pellets, resulting in a fuel assembly that leverages the performance of existing LWR fuel assembly designs and infrastructure with improved accident tolerance. The focus of this paper is to report on the maturity of the concept and the overall feasibility on the use of ferritic FeCrAl alloys as cladding for nuclear fuel in commercial light water reactors. The following aspects are reviewed: mechanical properties, corrosion performance, stress corrosion cracking, fabrication assessments and performance evaluation. The results show that FeCrAl alloy clad fuel rods (with UO{sub 2} fuel) have reasonably demonstrated capabilities to meet or exceed current fuel design technical requirements while providing increased safety benefit during design basis events and severe accident conditions. Initial characterization of the materials, in particular APMT (the current leading FeCrAl alloy candidate whose composition is Fe + 22Cr + 5Al + 3Mo) indicates excellent operational and accident condition performance in both PWR and BWR environments. Challenges on the neutron economy and tritium release have a reasonable path to mitigation. The fabrication processes for the FeCrAl/UO{sub 2} system appears to be very similar to current LWR fuel fabrication processes (pilgering/extruding, heat treatments, welding, NDE techniques, etc.) which are mature and well understood. No issues are anticipated complying with current nuclear industry quality and performance standards.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

The real-time data processing and acquisition system for Project 8 Phase II

In Phase II of the Project 8 neutrino mass experiment, electrons from the decays of tritium or 83m Kr are detected via their ≈26 GHz cyclotron radiation while contained within a circular waveguide. The signal from a given electron is characterized as a brief chirp, lasting ≲10 ms and changing in frequency by ≲1 MHz/ms. To detect these signals, the Project 8 collaboration developed a Data Acquisition (DAQ) system tailored to the signal properties. The DAQ is responsible for simultaneously selecting up to three 100 MHz-wide frequency windows to study, detect, and trigger on likely signals from different electron kinetic energies, and for writing the relevant data to disk. Finally, we describe the Phase II DAQ system in detail and address how the system was used for data-taking operations.

Data acquisition↗
DOE OSTI2025

The real-time data processing and acquisition system for Project 8 Phase II

In Phase II of the Project 8 neutrino mass experiment, electrons from the decays of tritium or 83 m Kr are detected via their ≈ 26GHz cyclotron radiation while contained within a circular waveguide. The signal from a given electron is characterized as a brief chirp, lasting ≲ 10ms and changing in frequency by ≲ 1MHz/ms. To detect these signals, the Project 8 collaboration developed a Data Acquisition (DAQ) system tailored to the signal properties. The DAQ is responsible for simultaneously selecting up to three 100MHz-wide frequency windows to study, detect, and trigger on likely signals from different electron kinetic energies, and for writing the relevant data to disk. We describe the Phase II DAQ system in detail and address how the system was used for data-taking operations.

Ashtari Esfahani, A. [Washington U., Seattle; Shar↗
DOE OSTI2015

Tritium Laboratory Karlsruhe: administrative and technical framework for isotope laboratory operation

Originally licensed in 1993 the Tritium Laboratory Karlsruhe (TLK) is a unique pilot scale isotope laboratory focused on tritium handling and processing to conduct a variety of scientific experiments and development tasks in view of future fusion power plants. TLK currently operates 15 glove boxes of 125 m{sup 3} total volume in an experimental hall measuring nearly 1500 m{sup 2}. The tritium infrastructure, comprising of the tritium storage system, the tritium transfer system and the isotope separation system, is integrated into TLK as a closed loop system to supply tritium to the experiments. Having a license for handling of up to 40 g of tritium and a closed tritium processing loop, TLK is a unique institute in non-military tritium research. In order to fulfil all requirements regarding the license, a framework of regulations is applied as a basis for the operation of TLK, as well as the setup of new experiments and the design of components. This paper will give an overview on the framework of operation in view of licensing issues, as well as administrative and technical regulations mandatory to legally and reliably operate an isotope laboratory of this scale.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2015

Novel Methods of Tritium Sequestration: High Temperature Gettering and Separation Membrane Materials Discovery for Nuclear Energy Systems

This project is aimed at addressing critical issues related to tritium sequestration in next generation nuclear energy systems. A technical hurdle to the use of high temperature heat from the exhaust produced in the next generation nuclear processes in commercial applications such as nuclear hydrogen production is the trace level of tritium present in the exhaust gas streams. This presents a significant challenge since the removal of tritium from the high temperature gas stream must be accomplished at elevated temperatures in order to subsequently make use of this heat in downstream processing. One aspect of the current project is to extend the techniques and knowledge base for metal hydride materials being developed for the ''hydrogen economy'' based on low temperature absorption/desorption of hydrogen to develop materials with adequate thermal stability and an affinity for hydrogen at elevated temperatures. The second focus area of this project is to evaluate high temperature proton conducting materials as hydrogen isotope separation membranes. Both computational and experimental approaches will be applied to enhance the knowledge base of hydrogen interactions with metal and metal oxide materials. The common theme between both branches of research is the emphasis on both composition and microstructure influence on the performance of sequestration materials.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Evaluation of hydrogen isotope exchange methodology on adsorbents for tritium removal

The Savannah River National Laboratory has demonstrated a potential process that can be used to remove tritium from tritiated water using Pt-catalyzed molecular sieves. The process is an elemental isotope exchange process in which H{sub 2} (when flowed through the molecular sieves) will exchange with the adsorbed water, D{sub 2}O, leaving H{sub 2}O adsorbed on the molecular sieves. Various formulations of catalyzed molecular sieve material were prepared using two different techniques, Pt-implantation and Pt-ion exchange. This technology has been demonstrated for a protium (H) and deuterium (D) system, but can also be used for the removal of tritium from contaminated water (T{sub 2}O, HTO, and DTO) using D{sub 2} (or H{sub 2}). (authors)

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2016

Synthesis and Performance Evaluations of SAPO-34 Membranes- Milestone Report for FCRD-MRWFD-2016-000263

This report describes the synthesis and evaluation of molecular sieve zeolite membranes to separate and concentrate tritiated water (HTO) from dilute HTO-bearing aqueous streams. Several silico alumino phosphate (SAPO-34) molecular sieve zeolite membranes were synthesized on tubular supports and characterized with gas and vapor permeation measurements. The pervaporation process performance was evaluated for the separation and concentration of tritiated water. Experiments were performed over a range of tritiated water concentration that correspond to the range anticipated in a nuclear fuel processing system that includes both acid and water streams recycling. The permeate was recovered under vacuum. The tritium concentration ranged from 0.5 to 1 mCi/mL which is about 0.1 mg/L or 0.1 ppm. The separation factors calculated from the measured tritium concentrations ranged from 0.83-0.98. The HTO concentration was three orders of magnitude lower than prior experiments performed with simulated feed containing HDO (>100 ppm) using deuterated water where high separation factors (>10) were obtained using SAPO membranes on alumina disk supports. Although the membrane performance characterization results for HTO were lower than expected, they can be explained on the basis of low feed volume and three orders of magnitude lower HTO concentration compared to HDO concentration in deuterated water. Several new approaches are proposed, such as tuning the diffusion coefficient of HTO, and optimization of membrane thickness that may help achieve preferential transport of tritium (HTO) resulting in a substantially more concentrated permeate.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2013

Roadmap for development of an advanced head-end reactor

A novel dry treatment process for used nuclear fuel (UNF) using nitrogen dioxide is being developed to remove volatile and semi-volatile fission products and convert the monolithic fuel material to a fine powder suitable as a feed to many different separations processes. The process may be considered an advanced form of voloxidation, which was envisioned to remove tritium from the fuel prior to introduction of the fuel into the aqueous separations systems, where subsequent separation of tritium from the water would be difficult and expensive. The product from NO{sub 2} reaction can be selectively chosen to be U{sub 3}O{sub 8}, UO{sub 3}, or a nitrate by adjusting the processing conditions; all products are generated at temperatures lower than those used in standard voloxidation. All the fundamental tenants of the process have been successfully demonstrated as a proof of principle, and many aspects have been corroborated multiple times at laboratory scale. The goal of this roadmap is to define the activities required to develop the process to a technology-readiness level sufficient to an engineering-scale implementation. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Benchmarking Hydrogen Isotope Separation Efficiency of Pd/k–Packed TCAP Columns

By leveraging the large isotope effect in the palladium hydrogen isotope system, the Thermal Cycling Absorption Process (TCAP) provides an efficient and advantageous means to separate protium, deuterium, and tritium. To meet increased future tritium processing demands, such as those needed for fusion power plants, current designs of the separation columns need to be adapted and optimized using the progress made in understanding hydrogen isotope science. One key to this optimization lies in understanding the baseline performance for currently employed separation packing materials. Pd/k and molecular sieves, as commonly used for the separation of hydrogen isotopes, are herein evaluated to establish a baseline for their separation efficiency. Van Deemter plots are formulated, and the influence of each parameter is evaluated to determine areas for improvement.

Thermal Cycling Absorption Process (TCAP)↗
DOE OSTI2017

Tritium Recovery From Mixed Waste - 17004

Nuclear Sources and Services (NSSI) in Houston, TX houses a unique infrastructure to recover tritium from a variety of different streams for re-use. NSSI accepts gaseous, aqueous, and tritiated-solid feedstock. For example, titanium-tritide foils from spent neutron generator sources are de-tritiated at 1000 deg. C. Tritium gas is recovered from a variety of metal getter beds. Tritiated liquid that is desorbed from zeolite drier beds or generated while reconditioning nickel beds is transferred to a Combined Electrolysis Catalytic Exchange (CECE) system to concentrate the tritium fraction before the gas is transferred to a hydrogen isotope separator for enrichment. Mixed-waste liquids, typically generated in the pharmaceutical industry, are converted to tritiated water and inactive gaseous effluents using a high-temperature, two-stage, catalytic-oxidation destruction process. The destruction efficiency of solvents is extremely high, typically in the 99.9999% to 99.99999% range. The resultant water containing tritium is also directed to the CECE subsystem. At present, liquid activities ranging from the 0.37 MBq/L (10 μCi/L) to 740 GBq/L (20 Ci/L) can be processed economically. The CECE subsystem has de-tritiated 2 metric tons of heavy water per year and is designed to de-tritiate up to 30 metric tons of light water per year. NSSI is licensed to operate within a 2-gram tritium inventory limit. Recently the Mixed-Waste Oxidation Facility that can process up to 40 liters of liquid per day has been upgraded to reduce the corrosive effects of halogen-based mixed wastes on the effluent scrubbing system. This paper will outline the key features of this system and discuss the effectiveness of the upgrades and the operational parameter space. (authors)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2015

Recent progress on tritium technology research and development for a fusion reactor in Japan Atomic Energy Agency

JAEA (Japan Atomic Energy Agency) manages 2 tritium handling laboratories: Tritium Processing Laboratory (TPL) in Tokai and DEMO-RD building in Rokkasho. TPL has been accumulating a gram level tritium safety handling experiences without any accidental tritium release to the environment for more than 25 years. Recently, our activities have focused on 3 categories, as follows. First, the development of a detritiation system for ITER. This task is the demonstration test of a wet Scrubber Column (SC) as a pilot scale (a few hundreds m{sup 3}/h of processing capacity). Secondly, DEMO-RD tasks are focused on investigating the general issues required for DEMO-RD design, such as structural materials like RAFM (Reduced Activity Ferritic/Martensitic steels) and SiC/SiC, functional materials like tritium breeder and neutron multiplier, and tritium. For the last 4 years, we have spent a lot of time and means to the construction of the DEMO-RD facility and to its licensing, so we have just started the actual research program with tritium and other radioisotopes. This tritium task includes tritium accountancy, tritium basic safety research such as tritium interactions with various materials, which will be used for DEMO-RD and durability. The third category is the recovery work from the Great East Japan earthquake (2011 earthquake). It is worth noting that despite the high magnitude of the earthquake, TPL was able to confine tritium properly without any accidental tritium release.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Acquired experience resulting from transforming a chemical installation into a nuclear one

ICIT-Valcea has developed an experimental pilot-scale installation for tritium and deuterium separation. The main objective of this pilot was to demonstrate the water detritiation technology and to transfer this technology to the CANDU reactors of the Cernavoda nuclear power plant. The pilot-scale installation was initiated in 1992. The initial design and construction were performed similarly to chemical plants as the separation of isotopes was focused on only hydrogen and deuterium to assess feasibility. In a second phase we have begun to transform it into a nuclear facility with the aim of separating tritium. Moving to tritium separation has imposed a lot of changes. Changes consisted mainly of: -) re-design of the technological systems for nuclear material processing, applying specific codes and standards (ASME, Romanian nuclear specific pressure boundary prescriptions for code classification); -) design and implementation of new systems, classified as safety systems; -) re-design and implementation of command and control systems, complying with the requirements of reliability and maintenance required for the project promoted; -) revaluation of auxiliary systems (utilities, power supply); -) implementing radiation protection systems, including secondary barriers; -) implementing and maintaining environment operational program specific to the new nuclear plant; -) developing and conducting safety analyzes; and -) the production of specific documentation to obtain the necessary permits for construction, commissioning and operation of the plant.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

Software For Advanced Large-scale Analysis Of Magnetic Confinement For Numerical Design, Engineering & Research (salamander)

As magnetic confinement fusion energy gains traction internationally to enable abundant energy production, designing components for fusion systems is a pressing challenge. During the planned lifetime of a fusion device, components evolve in extreme environments and must withstand large, repeated thermal loads and bombardment by 14 MeV neutrons, plasma ions, and neutral particles (deuterium, tritium, and helium), corrosive conditions, etc. All these physical processes take place simultaneously, interact in intricate ways, and impose important constraints that can affect performance. Experimental data is rare and costly to obtain, making design particularly challenging. Predictive computational frameworks must be an integral part of an accelerated and cost-effective design process by modeling fusion system performance in simulated environments. To better understand component degradation and operational impacts on their performance, the Software for Advanced Large-scale Analysis of MAgnetic confinement for Numerical Design, Engineering & Research (SALAMANDER) is designed as an open-source, fully integrated, multiphysics, multiscale, NQA-1 compliant framework facilitating 3D, high-fidelity fusion system modeling. To that end, SALAMANDER is a MOOSE-based framework, and therefore leverages MOOSE upstream libraries such as PETSc and libMesh to deliver sophisticated finite element, finite volume, and nonlinear solver technology for fusion energy simulations. SALAMANDER couples MOOSE physics module capabilities—such as thermal hydraulics, heat conduction, Navier-Stokes, and thermomechanics—with tritium transport via TMAP8, neutronics via Cardinal, and nascent particle-in-cell capabilities. Direct simulation Monte Carlo methods will be used to address neutral transport near the walls. By coupling all these physics in an integrated application, SALAMANDER will enable high-fidelity modeling of irradiation levels and plasma exposure conditions of plasma facing components and their impact on heat and tritium distributions, as well as the resulting mechanical constraints experienced by the plasma facing components and performance of blanket systems. Furthermore, SALAMANDER will be particularly suited for engineering studies thanks to the stochastic tool module readily available in MOOSE, allowing for extended uncertainty quantification and risk analysis studies. It is also able to use computer-aided design (CAD) meshes to model complex geometries, which is indispensable for fusion systems. SALAMANDER therefore supports design, safety, engineering, and research projects for magnetic confinement fusion systems

Simon, Pierre-Clement [Idaho National Laboratory (↗
DOE OSTI2018

Jefferson Lab Tritium Program

Electromagnetic probes of few-body nucleon systems play a pivotal role in our attempts to understand the structure of both nuclei and the protons and neutrons that comprise them. Of particular interest, are light nuclei where, among other motivating factors, models are more easily developed. The simplest asymmetric nuclear system is the A=3 system. While experiments involving 3He are relatively common, the combined study of the mirror nuclei 3He and 3H can provide much deeper insight. This provided the onus for the development of a tritium target system. After considerable effort the Jefferson Lab Tritium Program became operational with first beam in December of 2017. Since that time, data have been successfully collected for two experiments with two more now in process and expected to be completed by late 2018. These experiments range in scope from the study of Short Range Correlations to the Hypernuclear ?N interaction to the A=3 EMC effect and ratio of u/d in the nucleon. In this talk, the challenges of working with tritium and the development of the tritium target system are presented. Selected preliminary results and target performance will also be discussed.

Meekins, David↗
DOE OSTI2015

Proposal for secondary enclosure setup for experiments to expose plasma facing materials to tritiated plasma in VISIONI

VISIONI is an equipment at SCK-CEN that allows the exposure of candidate plasma facing materials to tritium - deuterium plasmas at ITER first wall conditions. VISIONI itself, being a vacuum setup, acts as primary confinement. To protect operators against exposure to a tritiated atmosphere VISIONI must be placed in a secondary confinement. The current Tritium lab at SCK-CEN has a walk-in process cell which can be used to enclose the plasma chamber and diagnostics of the VISIONI setup, which have a limited tritium inventory. This allows easy accessibility to the setup in a well-ventilated environment. Routine operations should be conducted from outside the process cell and maintenance operations can be conducted from within the process cell with proper protections. The tritium storage and supply can be enclosed in a glove box with a dedicated air detritiation system which is activated during an experiment or in case of an incident. The detritiation system will oxidize tritium and capture it on molecular sieves. By using this confinement approach it is possible to expose materials to a tritiated plasma while maintaining good accessibility of the VISIONI setup. This paper describes the proposed confinement system and compares it to the most common approach where the entire system is enclosed into one large glovebox.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Analytical Assessment of Impurities Found in Process Gases using the InfiTOF, a High Resolution, Low Mass, Variable Time-of-Flight Mass Spectrometer

The JEOL InfiToF is a high-resolution, low mass (HRLM) mass spectrometer (MS) with the application potential to meet analytical needs for tritium handling systems, such as those at the Savannah River Tritium Facilities (SRTE) and those used in fusion research. The Service Level Agreement (SLA) #144 line 27 “InfiTOF R&D” has concluded its first year of efforts under Tritium Recycle and Renewal. The project goals for FY22 were: (1) Production of a finalized modification list for tritium hardened components in the InfiTOF, (2) perform an assessment on impurity gases commonly found in process gases from Tritium Facilities and (3) begin developing a set of standardized work instructions that will be used as a basis for operations on the InfiTOF.

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

Tritium Content in and Release from Pressurized-Water-Reactor Fuel Cladding

It is expected that tritium pretreatment will be required in future reprocessing plants to prevent the release of tritium to the environment (except for long-cooled fuels; >30 yrs). To design and operate future reprocessing plants in a safe and environmentally compliant manner, the amount and form of tritium in the used nuclear fuel (UNF) must be understood and quantified. Tritium in light water reactor (LWR) fuel is dispersed between the fuel matrix and the fuel cladding, and some tritium may be in the plenum, probably as tritiated water (THO or T{sub 2}O). In a standard processing flowsheet, tritium management would be accomplished by treatment of liquid streams within the plant. Pretreating the fuel prior to dissolution to release the tritium into a single off-gas stream could simplify tritium management, so the removal of tritium in the liquid streams throughout the plant may not be required. The fraction of tritium remaining in the cladding may be reduced as a result of tritium pretreatment. Since Zircaloy'R cladding makes up roughly 25% by mass of UNF in the United States, processes are being considered to reduce the volume of reprocessing waste attributed to Zircaloy'R clad fuel by recovering the zirconium from the cladding for reuse. Treatment options for Zircaloy cladding include recycling to recover the significant value of the zirconium and eliminate costs associated with the disposal of transuranic-contaminated Zircaloy. If LWR fuel is reprocessed and the fuel is dissolved without tritium pretreatment, the tritium in the cladding will remain bound to the cladding. If tritium pretreatment is included in the flow sheet, some portion of the tritium in the cladding may be released along with the tritium from the fuel matrix. To design and operate reprocessing plants in a safe and environmentally compliant manner, the amount and form of tritium in the UNF must be understood and quantified. For Zircaloy-clad fuels from light water reactors, the tritium produced from ternary fission and other sources is expected to be divided between the fuel, where it is generated, and the cladding. It has been previously documented that a fraction of the tritium produced in uranium oxide fuel from LWRs can migrate and become trapped in the cladding. Estimates of the percentage of tritium in the cladding typically range from 0-96%. There is relatively limited data on how the tritium content of the cladding varies with burnup and fuel history (temperature, power, etc.) and how pretreatment impacts its release. The present study was undertaken to understand how tritium pretreatment at standard tritium pretreatment conditions (oxidation in air at 480 to 600 deg. C) affects the tritium content in the Zircaloy cladding and the extent to which the tritium content could be reduced with modest increases in the tritium pretreatment temperature. The data indicate that <20% of the tritium in PWR cladding will be removed by pretreatment at 500-700 deg. C for 24 h. Essentially 100% of the tritium will likely be released if the heating time at 700 deg. C is extended to 96 h. Small amounts of tritium are expected to be released if the cladding is heated at ≤ 600 deg. C for up to 96 h. The results of this study indicate that the amount of tritium released from tritium pretreatment systems will be dependent on both the operating temperature and length of time in the pretreatment system. Under certain conditions, a significant fraction of the tritium could remain bound in the cladding and would need to be considered in any subsequent processing of the cladding to recover/recycle the zirconium. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Recent advances in SRS on hydrogen isotope separation using thermal cycling absorption process

TCAP (Thermal Cycling Absorption Process) is a gas chromatograph in principle using palladium in the column packing, but it is unique in the fact that the carrier gas, hydrogen, is being isotopically separated and the system is operated in a semi-continuous manner. TCAP units are used to purify tritium. The recent TCAP advances at Savannah River Site (SRS) include compressor-free concept for heating/cooling, push and pull separation using an active inverse column, and compact column design. The new developments allow significantly higher throughput and better reliability from 1/10 of the current production system's footprint while consuming 60% less energy. Various versions are derived in the meantime for external customers to be used in fusion energy projects.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗