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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 235 records · Page 13
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Nuclear fuel irradiation testbed for nuclear security applications

The nuclear security community has long been interested in the identification and quantification of nuclear material signatures to understand a material’s provenance, use, and ultimate application. New forensics signatures and methods intended for non-traditional or advanced nuclear fuel applications require fuel irradiation experiments to demonstrate viability and validity. Integral fuel irradiations have historically required significant costs and long timelines to design, irradiate, and characterize. This paper describes how a recently developed nuclear fuel irradiation testbed can be used to provide a low cost, rapid turnaround, modular test environment for irradiation and evaluation of nuclear fuel specimens for nuclear security applications. The irradiation testbed houses six small ‘MiniFuel’ samples within hermetically sealed capsules inside targets that can be removed in between each ∼25-day operating cycle of the High Flux Isotope Reactor (HFIR). As many as nine targets can be irradiated using a single irradiation position (reflector region) in HFIR, allowing for varying irradiation temperatures and burnups. A suite of hot cell capabilities have been established to perform post-irradiation examination for measuring performance (e.g., fuel swelling, fission gas release) and facilitating experiment disassembly for subsequent property measurements, microstructural analysis, or chemical assay. This new testbed allows fuel irradiations to be conducted on an accelerated timeframe to enable rapid proof of concept testing and to provide reference material for nuclear fuel security applications. Recent applications using this testbed include the testing of isotopic taggants in UO 2 fuel (intentional forensics), testing of U-10Mo fuel for down-conversion of highly enriched uranium–fueled reactors, and the production of irradiated UO 2 fuel material for signature analysis of its isotopic composition (plutonium, fission gases, etc.).

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
NASA NTRS1973

Satellite nuclear power station: An engineering analysis

A nuclear-MHD power plant system which uses a compact non-breeder reactor to produce power in the multimegawatt range is analyzed. It is shown that, operated in synchronous orbit, the plant would transmit power safely to the ground by a microwave beam. Fuel reprocessing would take place in space, and no radioactive material would be returned to earth. Even the effect of a disastrous accident would have negligible effect on earth. A hydrogen moderated gas core reactor, or a colloid-core, or NERVA type reactor could also be used. The system is shown to approach closely the ideal of economical power without pollution.

Williams, J. R.↗
NASA NTRS1974

Applications of plasma core reactors to terrestrial energy systems

Plasma core reactors offer several new options for future energy needs in addition to space power and propulsion applications. Power extraction from plasma core reactors with gaseous nuclear fuel allows operation at temperatures higher than conventional reactors. Highly efficient thermodynamic cycles and applications employing direct coupling of radiant energy are possible. Conceptual configurations of plasma core reactors for terrestrial applications are described. Closed-cycle gas turbines, MHD systems, photo- and thermo-chemical hydrogen production processes, and laser systems using plasma core reactors as prime energy sources are considered. Cycle efficiencies in the range of 50 to 65 percent are calculated for closed-cycle gas turbine and MHD electrical generators. Reactor advantages include continuous fuel reprocessing which limits inventory of radioactive by-products and thorium-U-233 breeder configurations with about 5-year doubling times.-

Latham, T. S.↗
NASA NTRS1976

Georgia Institute of Technology research on the Gas Core Actinide Transmutation Reactor (GCATR)

The program reviewed is a study of the feasibility, design, and optimization of the GCATR. The program is designed to take advantage of initial results and to continue work carried out on the Gas Core Breeder Reactor. The program complements NASA's program of developing UF6 fueled cavity reactors for power, nuclear pumped lasers, and other advanced technology applications. The program comprises: (1) General Studies--Parametric survey calculations performed to examine the effects of reactor spectrum and flux level on the actinide transmutation for GCATR conditions. The sensitivity of the results to neutron cross sections are to be assessed. Specifically, the parametric calculations of the actinide transmutation are to include the mass, isotope composition, fission and capture rates, reactivity effects, and neutron activity of recycled actinides. (2) GCATR Design Studies--This task is a major thrust of the proposed research program. Several subtasks are considered: optimization criteria studies of the blanket and fuel reprocessing, the actinide insertion and recirculation system, and the system integration. A brief review of the background of the GCATR and ongoing research is presented.

Clement, J. D.↗
NASA NTRS1980

Health and safety: Preliminary comparative assessment of the Satellite Power System (SPS) and other energy alternatives

Data readily available from the literature were used to make an initial comparison of the health and safety risks of a fission power system with fuel reprocessing; a combined-cycle coal power system with a low-Btu gasifier and open-cycle gas turbine; a central-station, terrestrial, solar photovoltaic power system; the satellite power system; and a first-generation fusion system. The assessment approach consists of the identification of health and safety issues in each phase of the energy cycle from raw material extraction through electrical generation, waste disposal, and system deactivation; quantitative or qualitative evaluation of impact severity; and the rating of each issue with regard to known or potential impact level and level of uncertainty.

Habegger, L. J.↗
NASA NTRS1980

Comparative health and safety assessment of the SPS and alternative electrical generation systems

A comparative analysis of health and safety risks is presented for the Satellite Power System and five alternative baseload electrical generation systems: a low-Btu coal gasification system with an open-cycle gas turbine combined with a steam topping cycle; a light water fission reactor system without fuel reprocessing; a liquid metal fast breeder fission reactor system; a central station terrestrial photovoltaic system; and a first generation fusion system with magnetic confinement. For comparison, risk from a decentralized roof-top photovoltaic system with battery storage is also evaluated. Quantified estimates of public and occupational risks within ranges of uncertainty were developed for each phase of the energy system. The potential significance of related major health and safety issues that remain unquantitied are also discussed.

Habegger, L. J.↗
NASA NTRS1994

Tele/Autonomous Robot For Nuclear Facilities

Fail-safe tele/autonomous robotic system makes it unnecessary for human technicians to enter nuclear-fuel-reprocessing facilities and other high-radiation or otherwise hazardous industrial environments. Used to carry out experiments as exchanging equipment modules, turning bolts, cleaning surfaces, and grappling turning objects by use of mixture of autonomous actions and teleoperation with either single arm or two cooperating arms. System capable of fully autonomous operation, teleoperation or shared control.

Backes, Paul G.↗
DOE OSTI2017

Room temperature electrodeposition of actinides from ionic solutions

Uranic and transuranic metals and metal oxides are first dissolved in ozone compositions. The resulting solution in ozone can be further dissolved in ionic liquids to form a second solution. The metals in the second solution are then electrochemically deposited from the second solutions as room temperature ionic liquid (RTIL), tri-methyl-n-butyl ammonium n-bis(trifluoromethansulfonylimide) [Me.sub.3N.sup.nBu][TFSI] providing an alternative non-aqueous system for the extraction and reclamation of actinides from reprocessed fuel materials. Deposition of U metal is achieved using TFSI complexes of U(III) and U(IV) containing the anion common to the RTIL. TFSI complexes of uranium were produced to ensure solubility of the species in the ionic liquid. The methods provide a first measure of the thermodynamic properties of U metal deposition using Uranium complexes with different oxidation states from RTIL solution at room temperature.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Thermodynamic Properties of Rare Earth Chlorides in Molten Salts: A Novel Approach to Experimental Electrochemistry of Molten Salts

The thermodynamic and electrochemical properties of rare earth chlorides in molten salt solutions are of interest from the standpoint of both pyrochemical processing of spent nuclear fuel and for the development of Generation IV nuclear reactors. Knowledge of key properties like activity coefficients and phase diagrams will accelerate the development of safe reliable carbon free energy. In this work, three conduits of research have been pursued and reported within the larger theme of minimization of waste in spent fuel reprocessing and associated technology development. The activity coefficient of rare earth chlorides has been investigated using ground breaking theory and experimental approach. Furthermore, an ionic size dependent activity coefficient model is presented. This book also contains research into quaternary phase diagrams for LiCl-KCl-CsCl-RECl3 molten salt systems. This book should prove an excellent introduction to anyone pursuing a career in experimental high temperature electrochemistry and molten salt chemistry.

Bagri, Prashant↗
DOE OSTI2019

A Self-Irradiation Correction for the Hybrid K-Edge Densitometer

Hybrid x-ray fluorescence and K-edge densitometry is used at nuclear fuel reprocessing facilities to determine the concentrations of uranium and plutonium in the input accountability vessel. The uranium concentration is determined by K-edge densitometry, that is, from the step-difference in transmission on either side of the K-absorption edge measured using a continuous x-ray spectrum on a vial of the solution defining a well-known geometry. Historically this was done using a basic region of interest analysis but recently to accommodate more challenging solution full spectral analysis has been developed with considerable benefits. In parallel with the transmission measurement, the relative plutonium-to-uranium concentration is obtained from the relative strength of the x-ray–induced K-shell x-ray fluorescence production measured at a backward angle in energy-dispersive mode. However, gamma-ray emissions from fission products present in the solution also cause fluorescence of the characteristic actinide x-rays. Uncorrected, these self-irradiation–induced x-rays can introduce a bias into the reported U:Pu result. The commonly implemented method to correct this interference requires acquisition of a long (~8 hour) passive spectra for the sample. Although effective, it is often impractical to perform a passive measurement on every sample such that a representative passive spectrum is used. We have developed an alternative correction methodology using only the active x-ray fluorescence spectrum that eliminates the requirement for the time-consuming passive count. The active self-interrogation methodology will be discussed and initial performance results presented.

Mcelroy Jr, Robert↗
DOE OSTI2020

Treatment of Problematic Reactive Metal Wastes Using GeoMelt ICV - WM2020 Conference Paper

Decommissioning of sodium-cooled reactors and fast reactor technologies has generated a number of reactive metal waste configurations that are problematic to treat and typically lack cost effective treatment methods and disposition options. As a result, Veolia Nuclear Solutions, under contract with Idaho National Laboratory (owned by the U.S. Department of Energy and managed and operated by Battelle Energy Alliance, LLC) demonstrated its GeoMelt® In-Container Vitrification (ICV)™ technology to safely convert sodium metal to a non-reactive vitrified oxide form. The demonstration project, supported by glass formulation and crucible testing, consisted of a series of ICV™ melts that processed elemental sodium into stable non-reactive glass. INL is currently implementing GeoMelt® technology as a means to safely and reliably convert radioactive reactive metal residues that contaminate sodium cooled reactor components into waste forms that comply with existing disposition pathways. Reactive metal wastes require treatment in order to remove the Resource Conservation and Recovery Act (RCRA) reactivity and ignitability characteristics to comply with land disposal restrictions. GeoMelt®, which is an alternative to other potential treatment approaches, provides a robust approach that chemically converts the reactive metals to an inert oxide while also immobilizing radionuclides in a vitrified waste form with durability equal to or better than vitrified nuclear fuel reprocessing wastes (very robust and inert waste forms). Most other treatment approaches generate hydrogen gas which is problematic. In 2016, Veolia Nuclear Solutions first demonstrated the effectiveness of the GeoMelt® ICV™ process in deactivating reactive sodium metal. Crucible, bench-scale, and engineering-scale demonstrations were conducted on several surrogate waste configurations with various ratios of sodium metal and glass-formers. Each ratio and configuration demonstrated complete deactivation of the surrogate sodium metal. Follow-on work in 2017 demonstrated the deactivation of reactive sodium by GeoMelt® ICV™ at a higher waste loading relative to previously demonstrated work performed in 2016; the higher waste loading optimized glass chemistry while enhancing the economical full-scale treatment of reactive metals. Additionally, follow-on demonstration testing in 2018 and 2019 focused on more complex shapes and other reactive-metals (mocked up Experimental Breeder Reactor II [EBR-II] subassembly, sodium filled heat exchanger, and a can containing sodium potassium alloy) which were all performed at engineering-scale. Since 2018, a 10-metric ton full-scale GeoMelt unit (GeoMelt® Richland) was designed, installed, and commissioned at Perma-Fix Northwest in Richland, Washington for the treatment of reactive metal wastes. As of September 2019, over 900 55-gallon drums containing a total of around 3,500 lb of sodium with low levels of radioactivity have been treated at GeoMelt ® Richland, with resulting glass monoliths disposed at the Nevada National Security Site (NNSS). A full-scale radiological demonstration melt on an actual EBR-II subassembly has also been performed using the full-scale melter in 2019. The GeoMelt® technology is a proven radioactive waste treatment technology capable of immobilizing radioactive wastes, including bulk rubble such as drums and other steel vessels usually without pretreatment. Utilizing the GeoMelt® technology to treat reactive metals eliminates pretreatment steps resulting from having to separate the reactive metal from steel containers or jackets as GeoMelt® can easily operate at temperatures sufficient to melt the steel and expose the reactive metal for treatment. Eliminating handling steps of reactive metals is a significant safety advantage since reactive metals are pyrophoric. The results generated as a part of the 2018-2019 demonstration program are presented in the paper.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Gamma Radiolysis of Biphasic Acetohydroxamic Acid (AHA) Solvent Systems

Acetohydroxamic acid (AHA) has been proposed as a substitute for uranium(IV) and hydrazine as a plutonium complexant and neptunium reductant for simplified, single-cycle used nuclear fuel reprocessing flowsheets. However, the chemical behavior of AHA in an intense multi-component radiation field is poorly understood, especially under representative biphasic reprocessing solvent system conditions. In response to this critical knowledge gap, this study has investigated the gamma radiolytic integrity of AHA in aqueous nitric solutions in contact with an organic phase, comprising current and future reprocessing ligands (TBP, DEHBA, and DEHiBA) dissolved in n-dodecane diluent. Our data show negligible effect of the organic phase on the radiolytic behavior of AHA compared to complementary single-phase experiments.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2017

De-nitration Process in Glass Melter Off-Gas Treatment System - 17092

Vitrification technology has been generally applied to stabilize High-Level Radioactive Wastes (HLW) generated from spent fuel reprocessing for safe interim storage and final geological disposal. High temperature operation of the glass melter system generates substantial amounts of NOx with various hazardous radioactive elements, due to HLW containing a lot of nitrate. Hence the off-gas treatment system (OGTS) is one of the essential systems for the removal of hazardous radioactive elements including NOx, to ensure the stable and safe operation of the glass melter system. In general, the wet scrubbing process (e.g. Sub-merged Bed Scrubber: SBS) is installed in the upstream of the OGTS for HLW vitrification in order to remove volatile ruthenium, cesium and other radioactive mists and dusts. NOx together with ruthenium, iodine and submicron particles are removed in the wet absorber equipment to below the regulated limit by law. In the Rokkasho Vitrification Laboratory (RVL) constructed recently for inactive full scale mock-up tests, we have adopted the de-nitration process using reduction catalyst in addition to together with the wet scrubbing process considering the future scenarios that are compliant to the even stricter Air Pollution Control Law. By introducing this new system, we have successfully demonstrated a good de-nitration performance in the full scale mock-up test of the glass melter system. On the other hand, by using the wet scrubbing process, nitric acid would be generated by the absorption of NOx on this recovery process and should be reworked or disposed of as secondary liquid wastes. This is one of the deciding factors for enlarging the facility and increasing construction cost. That is the reason why IHI has developed the OGTS omitting the wet scrubbing process. In this paper, we have summarized the results on the de-nitration process of the OGTS in RVL and the progress of the optimizing denitration process for future reprocessing system applications. (authors)

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

Self-Absorption Corrections for Plutonium Measurements With the Tomographic Gamma Scanner - 17325

The Tomographic Gamma Scanner (TGS) is a Non-Destructive Assay (NDA) instrument that combines high-resolution gamma spectroscopy with low spatial resolution three-dimensional image reconstruction in order to reduce bias and improve precision for measurements of drummed waste. The TGS technique is especially suited to measure containers with non-uniform source distributions and low-to me-dium-density heterogeneous matrices, compared with traditional NDA methods. Canberra recently built, installed, and commissioned a TGS system, intended primarily for the measurement of plutonium (Pu) bearing waste, at a European nu-clear fuel reprocessing facility. Pu assays using the TGS technique present technical challenges distinct from those typically encountered in traditional LLW applications. Among these is the need to apply corrections for self-absorption effects that occur on scales smaller than the TGS spatial resolution, and can compromise the assay result if not included in the TGS imaging methodology. We present a brief review of the Self Absorption Correction (SAC) technique, and discuss the complications involved with its application to the TGS methodology with examples from our recent experience. This work continues a Canberra history of further developing and refining a technique to solve emerging problems, having commercialized the original methodology by partnering with the research institution that pioneered it. The TGS and the SAC methodologies were originally developed by Los Alamos National Laboratory (LANL) and John Fleissner respectively, and this work illustrates how Canberra has implemented an advancement of these techniques in a new TGS system. (authors)

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

The Micro-Optical Ring Electrode: A Sensor for Multiple Actinide Ions Monitoring - 17586

We present on the development of the Micro-Optical Ring Electrode (MORE) for the analysis of trans-uranium elements in aqueous mixtures. The MORE is a photo-electrochemical device based on a ring microelectrode that uses the insulator interior to the ring as a light guide. This single device exploits the unique photophysical and electrochemical properties of multiple analytes present in mixtures to quantify them. Our study aims to develop a protocol for the analysis of ions of uranium, neptunium and plutonium, which are most relevant to the nuclear industry, especially in the areas of decommissioning and fuel reprocessing, where speedy and safe identification of radioactive contaminants is essential. The technique offers a number of advantages over traditional methods in the area of radiochemical analysis such as real time monitoring in-situ, reduced radiation dose impact to the analyst as a result of greatly reduced sample handling and preparation, improved analysis times allowing quick decision making, at a much reduced cost (e.g. compared to ICP-MS and Scintillation counting). Here we present results of our study on non-radioactive surrogates for the target elements and discuss the advantages the technique can deliver to the nuclear industry. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Electrochemical Manipulation and Radiolytic Evaluation of Organic Phase Neptunium

Post for LDRD Annual Poster Session Under envisioned used nuclear fuel reprocessing conditions, neptunium (Np) is present in a mixture of extractable Np(IV)/Np(VI) and inextractable Np(V) species, the distribution of which is dependent on several factors that lead to the unintentional partitioning of Np into various phases and product streams, reducing process and cost efficiency. With this in mind, we tested an innovative approach to precisely control the oxidation state distribution of Np using novel, high surface area, optically transparent, ligand modified tin-doped indium oxide electrodes (LMEs). These proof-of-concept experiments employed a variety of radiation and electrochemistry (echem) techniques to determine: electrode radiation stability; aqueous echem behavior of Np at nITO|P3 electrode surfaces; radiation-induced Np-extractant reaction kinetics; and non-aqueous echem of Np in diethylhexyl butyramide (DEHBA) solutions.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2023

Solids from Fuel Dissolution, Process Solutions, and Waste Evaporation

Irradiated light water reactor (LWR) fuel contains material which is insoluble in nitric acid, and process solutions are unstable with respect to precipitation of compounds insoluble in nitric acid. Accordingly, solids will be encountered during fuel reprocessing, and operations must be conducted in a way to minimize formation of additional solids and to prevent their interference with reprocessing operations.

Campbell, D. O.↗