Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “Containers”

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.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 1,639 records · Page 91

Method for forming sulfur-containing electrode using salt additive

The present disclosure relates to sulfur-containing electrodes and methods for forming the same. For example, the method may include disposing an electroactive material on or near a current collector to form an electroactive material layer having a first porosity and applying pressure and heat to the electroactive material layer so that the electroactive material layer has a second porosity. The first porosity is greater than the second porosity. The electroactive material may include a plurality of electroactive material particles and one or more salt additives. The method may further include contacting the electroactive material layer and an electrolyte such that the electrolyte dissolves the plurality of one or more salt particles so that the electroactive material layer has a third porosity. The third porosity may be greater than the second porosity and less than the first porosity.

Chen, Shuru↗

Mitigation of Leakage and Water Ingress for a Promising Phase Change Material (PCM) Containing Polythylene Glycol (PEG): Preprint

A promising and cost-effective phase change material (PCM) containing polyethylene glycol (PEG) 600, 800, and/or 1000 has been synthesized. It is shape stabilized using a sol-gel method and displays promising thermal properties. These properties make the PCMs good candidates for thermal batteries and other thermal energy storage applications in the building sector. However, PEG leakage is observed at elevated temperatures, and due to the hygroscopic nature of PEG, moisture ingress is also observed, which can have a significant effect on performance and durability. Thus, we have developed hydrophobic coatings to prevent PEG leakage and mitigate moisture ingress. The selected approach involves a secondary sol-gel process during which a thin layer formulated with hydrophobic precursors is deposited onto the surface of the PCM samples. Several coating formulations have been investigated, including formulations composed of tetraethyl orthosilicate (TEOS), diethoxy(dimethyl)silane (DEDMS), triethoxymethylsilane (TEMS), and n- octyltriethoxysilane (OTES). We find that this is effective at increasing hydrophobicity of the PCM substrates. It also decreases substrate leakage.

ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATION↗

Triple containment targets for particle irradiation

The invention provides a system for containing irradiated particles, the system having a housing having a closed upstream end and an open downstream end; an axially disposed tunnel extending from the downstream end to the upstream end, wherein longitudinally extending regions of the tunnel define a plurality of threaded surfaces; a sample cup positioned within the tunnel and proximal to the closed upstream end; threaded plugs matingly received by the threaded surfaces so as to provide at least one seal between the sample cup and the atmosphere external of the housing; and a plurality of deformable substrates disposed between the plugs.

Smith, Nicholas A.↗

Composite media for non-oxidative ethane dehydrogenation, and related ethane activation systems and method of processing an ethane-containing stream

A composite media for non-oxidative C 2 H 6 dehydrogenation comprises an aluminosilicate zeolite matrix, and an EDH catalyst on one or more of an external surface of the aluminosilicate zeolite matrix and internal surfaces within pores of the aluminosilicate zeolite matrix. The EDH catalyst comprises one or more of Fe, Zn, Pt, Ga, alloys thereof, and oxides thereof. A C 2 H 6 activation system, and a method of processing a C 2 H 6 -containing stream are also described.

Wang, Lucun↗

Synthesis, structures and properties of new hybrid solids containing ruthenium complexes and polyoxometalates

Two new organic-inorganic hybrid solids containing Keggin ions and ruthenium complexes have been synthesized and characterized by FT-IR, UV-vis, luminescence, X-ray, and TG analysis. In KNa[Ru(bpy){sub 3}]{sub 2}[H{sub 2}W{sub 12}O{sub 40}]{center_dot}8H{sub 2}O (1), the [Ru(bpy){sub 3}]{sup 2+} (bpy=2,2 Prime -bipyridine) complex ions are located in between the infinite one-dimensional double-chains formed by adjacent Keggin anions [H{sub 2}W{sub 12}O{sub 40}]{sup 6-} linked through {l_brace}KO{sub 7}{r_brace} and {l_brace}NaO{sub 6}{r_brace} polyhedra, while in K{sub 6}[Ru(pzc){sub 3}]{sub 2}[SiW{sub 12}O{sub 40}] Bullet 12H{sub 2}O (2), the [Ru(pzc){sub 3}]{sup -} (pzc=pyrazine-2-carboxylate) complex anions are confined by layered networks of the [SiW{sub 12}O{sub 40}]{sup 4-} clusters connected by potassium ions. Both compounds exhibit three-dimensional frameworks through noncovalent interactions such as hydrogen bonds and anion Midline-Horizontal-Ellipsis {pi} interactions. Additionally, compound 1 shows strong luminescence at 604 nm in solid state at room temperature. - Graphical abstract: Two three-dimensional framework solids are constructed from polyoxoanions and ruthenium-organic complexes through noncovalent interactions. Highlights: Black-Right-Pointing-Triangle Ru complexes form hybrid solids with polyoxometalates. Black-Right-Pointing-Triangle Anion Midline-Horizontal-Ellipsis {pi} interaction between polyoxometalates and metal complexes was observed. Black-Right-Pointing-Triangle Noncovalent interactions play an important role in the assembly of solids. Black-Right-Pointing-Triangle The hybrid solid shows luminescence properties.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Fuel salt and container material studies for MOSART transforming system

A study is under progress to examine the feasibility of single stream Molten Salt Actinide Recycling and Transmuting system without and with Th support (MOSART) fuelled with different compositions of actinide tri-fluorides (AnF{sub 3}) from used LWR fuel. New fast-spectrum design options with homogeneous core and fuel salts with high enough solubility for AnF{sub 3} are being examined because of new goals. The flexibility of single fluid MOSART concept with Th support is underlined, particularly, possibility of its operation in self-sustainable mode (Conversion Ratio: CR=1) using different loadings and make up. The paper summarizes the most current status of fuel salt and container material data for the MOSART concept received within ISTC-3749 and ROSATOM-MARS projects. Key physical and chemical properties of various fluoride fuel salts are reported. The issues like salt purification, the electroreduction of U(IV) to U(III) in LiF-ThF{sub 4} and the electroreduction of Yb(III) to Yb(II) in LiF-NaF are detailed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Parameter survey on heating conditions for glass-bonded sodalite ceramic waste fabrication from type-A zeolite containing simulating FPs

In the pyrometallurgical reprocessing for metal fuel cycle, fission products (FPs) in the spent salt is converted into a glass-bonded sodalite by means of Pressure-less Consolidation (PC). Although a standard condition of the PC method was reported, the reason for choosing this condition is not clear particularly for the temperature condition. In the present study, a parameter survey on the heating condition was performed by fabricating the glass-bonded sodalite containing simulating FPs under various conditions. The maximum temperature, the heating duration, the glass ratio in the initial material, and the weight load for pressing the material were chosen as the variable parameters. The mass balance of each element and the quality of the product were evaluated. It was exhibited that both the volatilized content and the free salt in the product were reduced by lowering the maximum temperature. The apparent density of the product was enhanced by both increasing the heating duration and increasing the weight load. Accordingly, lowering the maximum temperature while increasing the weight load was chosen to modify the fabricating condition. As a result of the modified condition, both the volatilized and the free salt ratios were reduced without significant change in the apparent density of the product. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

The Effect of the Presence of Ozone on the Lower Flammability Limit (LFL) of Hydrogen in Vessels Containing Savannah River Site High Level Waste - 12387

The Enhanced Chemical Cleaning (ECC) process uses ozone to effect the oxidation of metal oxalates produced during the dissolution of sludge in the Savannah River Site (SRS) waste tanks. The ozone reacts with the metal oxalates to form metal oxide and hydroxide precipitants, and the CO{sub 2}, O{sub 2}, H{sub 2}O and any unreacted O{sub 3} gases are discharged into the vapor space. In addition to the non-radioactive metals in the waste, however, the SRS radioactive waste also contains a variety of radionuclides, hence, hydrogen gas is also present in the vapor space of the ECC system. Because hydrogen is flammable, the impact of this resultant gas stream on the Lower Flammability Limit (LFL) of hydrogen must be understood for all possible operating scenarios of both normal and off-normal situations, with particular emphasis at the elevated temperatures and pressures of the typical ECC operating conditions. Oxygen is a known accelerant in combustion reactions, but while there are data associated with the behavior of hydrogen/oxygen environments, recent, relevant studies addressing the effect of ozone on the flammability limit of hydrogen proved scarce. Further, discussions with industry experts verified the absence of data in this area and indicated that laboratory testing, specific to defined operating parameters, was needed to comprehensively address the issue. Testing was thus designed and commissioned to provide the data necessary to support safety related considerations for the ECC process. A test matrix was developed to envelope the bounding conditions considered credible during ECC processing. Each test consists of combining a gas stream of high purity hydrogen with a gas stream comprised of a specified mixture of ozone and oxygen in a temperature and pressure regulated chamber such that the relative compositions of the two streams are controlled. The gases are then stirred to obtain a homogeneous mixture and ignition attempted by applying 10J of energy to a fuse wire. A gas combination is considered flammable when a pressure rise of 7% of the initial absolute pressure is observed. The specified testing methodology is consistent with guidelines established in ASTM E-918-83 (2005) 'Standard Practices for Determining Limits of Flammability of Chemicals at Elevated Temperature and Pressure'. The LFL of hydrogen in air was determined and is in good agreement with the literature data. Ozone-oxygen mixtures were found to be flammable at concentrations above 8.3 vol.% based on the ASTM E918 7% pressure rise criteria for flame propagation. This result is lower than previously reported values which can be explained through the variations in the test setup and procedure. It is believed that the lower values obtained in this work are a result of improvements of the test methodology. Tests performed with hydrogen in various concentrations of ozone in oxygen have shown that the LFL of hydrogen decreases as the concentration of ozone in the mixture increases. This testing was designed to provide data under the conditions considered most optimal to produce deflagration. The geometry and materials of construction of the testing vessel; the location of the fuse wire; the magnitude of the supplied energy; the careful minimization of diluents and other contaminants; and meticulous procedural detail to maintain integrity of the ozone to the maximum extent practical, result in data that reflect not the expected process conditions, but those that enhance the possibility of flame propagation. For this reason, there is believed to be considerable conservatism in the indicated results. Per the vendor, the maximum possible ozone concentration producible by the planned ECC Ozone generator is 8 volume percent (the typical maximum operating setpoint concentration is 6.8 vol%), less than the 8.3 minimum volume % concentration shown to be flammable in a 99.999% pure O{sub 2} environment at the optimally conservative conditions established in this testing. Further, the feed to the ECC ozone generator is only 87% oxygen, the remainder, water vapor and nitrogen, both powerful diluents. It is, thus, believed not credible that deflagration can occur at this maximum feed concentration condition. In addition, once the ozone stream contacts the waste stream, the many simultaneous oxidizing reactions will rapidly decompose available ozone to well below flammable levels. Further, because the radiolytically generated hydrogen quantity is negligible compared to the supplied ozone/oxygen stream (0.0004 moles per minute H{sub 2} vs 76 moles per minute ozone/oxygen), even the total H{sub 2}/O{sub 3} mixture, without crediting decomposition reactions, does not approach flammable concentrations. Finally, even at the 'end' of the ECC batch cycle, when most of the metal oxalates have been decomposed, testing has indicated that the ozone concentration in the vapor space of the ECC process vessel reaches a concentration of no more than 3 vol%, remaining well below concentrations of concern. The major issue for the ECC operation established by this testing is the impact of the data when applied to off normal conditions. While it is possible to discontinue ozone addition to the reaction vessel at any time, the radiolytic hydrogen generation rate continues, varying slightly as ambient pressures and temperatures change. Relative to the data generated and analyzed in this testing, the ECC hazards analysis team will re-evaluate off normal conditions (e.g. those during which process exhaust ventilation is lost) such that issues involving mixtures of hydrogen and ozone in the vapor space can be appropriately controlled. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Decontamination of organic wastes containing radionuclides - 15122

The Arvia Process is being adapted for use in the nuclear industry for the decontamination of oils containing radionuclides. This waste treatment process uses adsorption followed by electrochemical oxidation as an alternative technique to conventional incineration. A study has been carried out on components of the Arvia Process with the aim of understanding the changes in chemistry due to ionizing radiation. Samples of adsorbent in water have been exposed to gamma-rays from a Co-60 irradiator at absorbed doses of 3, 19, and 304 kGy. Preliminary results show that the concentration of inorganic ions and carbon in the water increased as a function of dose. This increase suggests that degradation of the adsorbent may take place at high doses. Further characterization of the microstructure and adsorption studies post-irradiation are planned. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Conditioning of Pu-containing radioactive waste generated in the Hot-lab: where are we now - 15400

To achieve an optimal strategy for interim storage and final disposal, solid radioactive waste has to be sorted and treated. This strategy is based on the requirement to reduce the volume of waste designated for long term disposal, mainly for economical and safety reasons. In addition, the radiological limitations for shipping of the waste packages to a disposal site have to be taken into account as well as the improved regulatory requirements for material under IAEA control. In case of the operational Pu-contaminated and -containing waste from the PSI-Hot laboratory in Switzerland, the existing and over years practiced conditioning method needed improvement. The existing authorized specifications of the conditioning process had to be revised for an optimization of the interim-storage and the deep geological disposal. This was done by the PSI Section for Dismantling and Waste Management. The improvement focused on technical aspects as well as improvement of the documentation for the certification by final depositor (Nagra) and the safety authority (ENSI). This work started in 2008. In 2012 ENSI issued an authorization to realize the first conditioning according the new specification, formally as a certification test ('Type Test') prior to final approval by the ENSI. But, although a licensed method by the competent authority is available, modifications could be necessary and new tasks come up. In 2013 the PSI has produced 13 waste-packages as a Type Test. To obtain the legal authorization ('Type Approval'), some open issues ought to be answered. Waste is continuously generated and so the question arises: When will the next conditioning be done and applying which method? (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Copolymers containing pendant ionomeric carbosilane groups

Cationic polymers are provided that comprise monomeric units of Formula (V). (V) Each asterisk (*) indicates an attachment position to another monomeric unit; R is hydrogen or methyl; each R 2 is each independently an alkyl, aryl, or a combination thereof; L is a linking group comprising an alkylene group; and +R 3</sup? is a cationic nitrogen-containing group free of any N—H bonds. Membranes formed from said cationic polymers, devices including such membranes, and methods of making such cationic polymers are also provided.

Hartmann-Thompson, Claire↗

Investigation of Fission Product Interaction with the SiC Containment Layer of High Turnup TRISO Fuel Particles

Tri-structural Isotropic (TRISO) encapsulated fuel particles are used in high-temperature gas-cooled nuclear reactors. Fuel performance is limited by the propensity of metallic fission products to be released under high temperature testing to simulate hypothesized accident scenarios. The current particle design consists of a 350 μm or 425 μm UO{sub 2} or UCO fuel kernel surrounded by a 100μm porous carbon buffer layer, a 40μm inner pyrolytic carbon (IPyC) layer, a 35μm high-density silicon carbide (SiC) layer, and a 40μm outer pyrolytic carbon (OPyC) layer. Of interest is the fission product interaction with the SiC layer, which provides the most structural support and is the main barrier to the diffusion of fission products. Previous laboratory examination studies have indicated the release of Ag through seemingly intact particles and the localized corrosion of the SiC due to Pd. Although a number of studies have been conducted, the specifics of both the Ag and Pd reaction mechanisms, in addition to U and Pu, through SiC are currently not well understood. This work aims to understand the pathways and kinetics of these interactions under normal operation as well as accident conditions. In order to improve our understanding of this, we have the unique opportunity to examine irradiated TRISO particles to study post-irradiation effects on these materials. Experimental data is necessary to constrain the current models used to predict the behavior of fission products in the SiC layer of TRISO fuels. This work aims to understand the interaction of fission products with the SiC containment layer in TRISO particles. We have systematically examined three SiC shells from irradiated TRISO particles using XAFS, TEM, and EDS. Results indicate the formation of palladium silicides, and uranium and plutonium carbides. Additionally, silver is seen to be metallic. Preliminary microscopy results indicate fission products throughout the SiC layer of the TRISO particles, and microstructural changes have been observed between the as-irradiated and safety tested particles. These results have provided key information on the fission product transport behavior through irradiated SiC at varying temperatures. Knowledge of these reaction pathways will allow for better simulation of the long-term behavior of TRISO fuels. They may also suggest ways to modify the SiC layer to improve fuel performance and mitigate fission product release, which is critical for safety strategies required to commercialize these nuclear fuel technologies. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Investigation of instrumentation containing an embedded digital device - 257

The U.S. nuclear power industry primarily relies on mature, well-proven analog technology as the basis for most instrumentation and control (I and C) equipment employed in its nuclear power plants (NPPs). However, the market for this legacy instrumentation is very small and availability of such equipment for spare parts or replacement is increasingly limited. Frequently, modern I and C equipment contains embedded digital devices (EDDs), such as microprocessors and programmable logic devices. While the enhanced functional capabilities provided by the digital components offer improved performance, reliability, and maintainability, the corresponding complexity of sophisticated digital implementations also exacerbate concerns about an increased potential for undetected systematic faults, which can lead to common-cause failure (CCF). As part of a research project sponsored by the U.S. Department of Energy, an investigation has been conducted of commercially available instrumentation marketed for nuclear or industrial application. The objective was to identify the usage of EDDs and to determine the types of functional roles allocated to the devices. Examples of the equipment surveyed include smart sensors, sequencers for emergency diesel generators, pumps, valve actuators, motor control centers, breakers, priority logic modules, and uninterruptible power sources. Where sufficient detail was found, an evaluation was performed to ascertain the nature and role of the digital technology implemented in the instrumentation. The information gained through this investigation serves as technical input for defining a classification approach for EDDs based on the functional impact of the device. For example, the role of an EDD can involve any of the following functions: performance monitoring and diagnostics, information extraction and communication, instrument condition monitoring and protection, or execution of core instrument functions. Classifying digital devices according to their role in the operation of the instrumentation in which they are embedded can enable determination of the potential functional impact of failure of that device and contribute to determination of the potential safety significance of a CCF event disabling all instances of that device. This paper will report on the findings of the investigation of instrumentation with EDDs and indicate considerations for the emerging classification approach. The higher-level goal of this effort is to determine whether a graded approach to qualification testing may be possible based on classification. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Sodium Spray Fire Simulations using CONTAIN-LMR

License applications for sodium fast reactors will include sodium fire analysis and subsequent mitigation strategies. Sodium leaks or ruptures from piping and support systems can lead to significant fires due to the reactive nature of sodium with air and water. Understanding the behavior of sodium fires is crucial for modeling and predicting accidents in sodium fast reactors. This paper describes the initial results of a control volume code, CONTAIN-LMR (CLMR), modeling effort for a sodium spray fire experiment conducted at Sandia National Laboratories (SNL) in Albuquerque, New Mexico in 2010. Additional sodium spray and pool fire tests were conducted at SNL in 2009 and 2010 and will be compared to CLMR results in future papers. The purpose of this analysis was extract information from the evolution of the SNL tests to better understand how to use CLMR in future sodium fire analysis. Thus the following CLMR simulations were not completely the result of blind benchmarking study. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Atalante research facility new mass limits in a unit containing plutonium and minor actinides - 14402

ATALANTE, located in Marcoule, is one of the main Nuclear Facilities of the French Alternative Energies and Atomic Energy Commission (CEA). This facility is mainly dedicated to research and development for nuclear energy related to the back end of the fuel cycle: spent fuel and final waste management. It is a large research complex housing chemistry and analytical laboratories and shielded cells (240 glove boxes and 11 shielded cells), equipped for basic studies through to demonstration experiments. Within the facility, one unit can contain plutonium and curium at the same time. In this case, nuclear criticality safety is achieved by exercising control over mass. A subcritical mass limit is defined for plutonium (280 g for {sup 239}Pu) and an exemption mass limit (3 g for {sup 245}Cm) is used for the actinide. In order to increase the maximum mass (in particular {sup 245}Cm) which can be handled in the unit, new mass control limits were studied, based on control over the sum of the mass ratio to the upper subcritical mass, noted m{sub lim}, for each nuclide (350 g for {sup 239}Pu; 34.8 g or 23 g for {sup 245}Cm) as shown in equation (1) below. m({sup 239}Pu)/ m{sub lim}({sup 239}Pu)+ m({sup 245}Cm)/m{sub lim}({sup 245}Cm)≤1 (1) To validate this criticality safety control method, a study was conducted to demonstrate that, for different masses of {sup 239}Pu and {sup 245}Cm meeting equation (1), the calculated effective multiplication factor is always lower than the effective multiplication factor corresponding to a sphere of 350 g of pure {sup 239}Pu, moderated, with 20 cm water reflection. The two pseudo fissile media are: - metallic plutonium (100% {sup 239}Pu, density 19.86), moderated with water, - metallic curium (100% {sup 245}Cm, density 13.566), moderated with water. The calculation model is a sphere of {sup 245}Cm metal positioned in a sphere of {sup 239}Pu metal, with optimum moderation for each fissile material, and a full water reflection (20 cm). The results show that, under certain conditions, the effective multiplication factor of the mixture can be higher than the effective multiplication factor of each pure fissile material when using the upper subcritical mass of 34.8 g for {sup 245}Cm. When using 23 g of {sup 245}Cm, the k{sub eff} of the mixture is always lower than the k{sub eff} of pure {sup 239}Pu. These calculations enable the safety documentation to be changed by introducing a new process operational procedure, based on equation (1). This change in procedure allows the operator to work with a larger quantity of curium in the unit. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Steam Condensation Scaled Experiment in the Presence of Non-condensable Gas for Reactor Containment Passive Safety Analysis

This study presents scaled experiments using steam condensation with non-condensable gas (NCG)—helium, simulating hydrogen—as these experiments are pivotal for water-cooled reactor passive containment cooling system (PCCS) design and analysis. Research into PCCSs for small modular reactors (SMRs) is especially important in light of SMR system design; however, studies in the literature reflect limitations due to test geometry and operational condition variations, without considering SMR prototypic design. To address these challenges, a scaled test facility was developed to accurately replicate SMR PCCSs. This facility includes vertical down-flow condensing test sections with 1-, 2-, and 4-in.-diameter condensing tubes, accompanied by annular water cooling. Experiments were conducted using both superheated and saturated steam, with steam mass flow rates varying from 55 to 66 kg/hr., in the presence of helium as the NCG mass flow rate ranges from 1.8 to 22 kg/hr. Test data were collected on (a) the axial temperatures of the annular cooling water; (b) the outer wall temperature of the condensers; and (c) the mass flow rate, temperature, and pressure at the test section inlets and outlets. These primary test data were used in conjunction with a standard data reduction methodology to estimate essential thermal parameters such as heat fluxes, heat transfer coefficients, and condensation rates. The effects of NCGs on steam condensation within the geometry of the scaled test sections were then presented in regard to various testing conditions.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗