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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 1,711 records · Page 95
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Small-Scale Coal-Biomass to Liquids Production Using Highly Selective Fischer-Tropsch Synthesis

The research project advanced coal-to-liquids (CTL) and coal-biomass to liquids (CBTL) processes by testing and validating Chevron’s highly selective and active cobalt-zeolite hybrid Fischer-Tropsch (FT) catalyst to convert gasifier syngas predominantly to gasoline, jet fuel and diesel range hydrocarbon liquids, thereby eliminating expensive wax upgrading operations The National Carbon Capture Center (NCCC) operated by Southern Company (SC) at Wilsonville, Alabama served as the host site for the gasifier slip-stream testing/demonstration. Southern Research designed, installed and commissioned a bench scale skid mounted FT reactor system (SR-CBTL test rig) that was fully integrated with a slip stream from SC/NCCC’s transport integrated gasifier (TRIG TM ). The test-rig was designed to receive up to 5 lb/h raw syngas augmented with bottled syngas to adjust the H 2 /CO molar ratio to 2, clean it to cobalt FT catalyst specifications, and produce liquid FT products at the design capacity of 2 to 4 L/day. It employed a 2-inch diameter boiling water jacketed fixed-bed heat-exchange FT reactor incorporating Chevron’s catalyst in Intramicron’s high thermal conductivity micro-fibrous entrapped catalyst (MFEC) packing to efficiently remove heat produced by the highly exothermic FT reaction.

Gangwal, Santosh K.↗
DOE OSTI2015

BWR Spent Nuclear Fuel Integrity Research and Development Survey for UKABWR Spent Fuel Interim Storage

The objective of this report is to identify issues and support documentation and identify and detail existing research on spent fuel dry storage; provide information to support potential R&D for the UKABWR (United Kingdom Advanced Boiling Water Reactor) Spent Fuel Interim Storage (SFIS) Pre-Construction Safety Report; and support development of answers to questions developed by the regulator. Where there are gaps or insufficient data, Oak Ridge National Laboratory (ORNL) has summarized the research planned to provide the necessary data along with the schedule for the research, if known. Spent nuclear fuel (SNF) from nuclear power plants has historically been stored on site (wet) in spent fuel pools pending ultimate disposition. Nuclear power users (countries, utilities, vendors) are developing a suite of options and set of supporting analyses that will enable future informed choices about how best to manage these materials. As part of that effort, they are beginning to lay the groundwork for implementing longer-term interim storage of the SNF and the Greater Than Class C (CTCC) waste (dry). Deploying dry storage will require a number of technical issues to be addressed. For the past 4-5 years, ORNL has been supporting the U.S. Department of Energy (DOE) in identifying these key technical issues, managing the collection of data to be used in issue resolution, and identifying gaps in the needed data. During this effort, ORNL subject matter experts (SMEs) have become expert in understanding what information is publicly available and what gaps in data remain. To ensure the safety of the spent fuel under normal and frequent conditions of wet and subsequent dry storage, intact fuel must be shown to: 1.Maintain fuel cladding integrity; 2.Maintain its geometry for cooling, shielding, and subcriticality; 3.Maintain retrievability, and damaged fuel with pinhole or hairline cracks must be shown not to degrade further. Where PWR (pressurized water reactor) information is utilized or referenced, justification has been provided as to why the data can be utilized for BWR fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Catalytic Deoxydehydration of Carbohydrates and Polyols to Chemicals and Fuels

As the world's fossil fuel resources are being depleted and their costs increase, there is an urgent need to discover and develop new processes for the conversion of renewable, biomass resources into fuels and chemical feedstocks. Research and development in this area have been given high priority by both governmental agencies and industry. To increase the energy content and decrease the boiling points of biomass-derived carbohydrates and polyols to the useful liquid range it is necessary to chemically remove water (dehydrate) and, preferably, oxygen (deoxygenate/reduce). The poly-hydroxylic nature of carbohydrates is attractive for their use as functionalized chemical building blocks, but it presents a daunting challenge for their selective conversion to single product chemicals or fuels. The long term, practical objective of this project is to develop catalytic processes for the deoxydehydration (DODH) of biomass-derived carbohydrates and polyols to produce unsaturated alcohols and hydrocarbons of value as chemical feedstocks and fuels; DODH: polyol + reductant --(LMOx catalyst)--> unsaturate + oxidized reductant + H2O. Limited prior studies have established the viability of the DODH process with expensive phosphine reductants and rhenium-catalysts. Initial studies in the PI's laboratory have now demonstrated: 1) the moderately efficient conversion of glycols to olefins by the economical sulfite salts is catalyzed by MeReO3 and Z+ReO4-; 2) effective phosphine-based catalytic DODH of representative glycols to olefins by cheap LMoO2 complexes; and 3) computational studies (with K. Houk, UCLA) have identified several Mo-, W-, and V-oxo complexes that are likely to catalyze glycol DODH. Seeking practically useful DODH reactions of complex polyols and new understanding of the reactivity of polyoxo-metal species with biomass-oxygenates we will employ a two-pronged approach: 1) investigate experimentally the reactivity, both stoichiometric and catalytic, of polyoxo-complexes and practical reductants with representative polyols to establish structure/reactivity relationships and reaction mechanisms; and b) carry out parallel computational studies of these reactions and their mechanisms- both analytical and predictive. Our prioritized action plan is: (1) to optimize the catalytic efficiency, assess the substrate scope/selectivity, and address key mechanistic aspects of Re-catalyzed, sulfite-driven DODH reactions; (2) use the findings from (1), together with computational predictions, to discover new, effective non-precious metal catalysts for sulfite-driven DODH reactions; and 3) to initiate exploratory studies of CO- and H2-driven DODH. Successful execution of this research project will: 1) provide practical chemical processes for the conversion of biomass into useful chemicals and fuels: 2) bring fundamental new understanding of chemical reactions involving metal-oxo catalysts; and 3) provide educational and technical training of future energy scientists.

09 BIOMASS FUELS↗
DOE OSTI2014

Qualification of Daiichi Units 1, 2, and 3 Data for Severe Accident Evaluations - Process and Illustrative Examples from Prior TMI-2 Evaluations

The accidents at the Three Mile Island Unit 2 (TMI-2) Pressurized Water Reactor (PWR) and the Daiichi Units 1, 2, and 3 Boiling Water Reactors (BWRs) provide unique opportunities to evaluate instrumentation exposed to severe accident conditions. Conditions associated with the release of coolant and the hydrogen burn that occurred during the TMI-2 accident exposed instrumentation to harsh conditions, including direct radiation, radioactive contamination, and high humidity with elevated temperatures and pressures. As part of a program initiated in 2012 by the Department of Energy Office of Nuclear Energy (DOE-NE), a review was completed to gain insights from prior TMI-2 sensor survivability and data qualification efforts. This initial review focused on the set of sensors deemed most important by post-TMI-2 instrumentation evaluation programs. Instrumentation evaluation programs focused on data required by TMI-2 operators to assess the condition of the reactor and containment and the effect of mitigating actions taken by these operators. In addition, prior efforts focused on sensors providing data required for subsequent forensic evaluations and accident simulations. To encourage the potential for similar activities to be completed for qualifying data from Daiichi Units 1, 2, and 3, this report provides additional details related to the formal process used to develop a qualified TMI-2 data base and presents data qualification details for three parameters: primary system pressure; containment building temperature; and containment pressure. As described within this report, sensor evaluations and data qualification required implementation of various processes, including comparisons with data from other sensors, analytical calculations, laboratory testing, and comparisons with sensors subjected to similar conditions in large-scale integral tests and with sensors that were similar in design to instruments easily removed from the TMI-2 plant for evaluations. As documented in this report, results from qualifying data for these parameters led to key insights related to TMI-2 accident progression. Hence, these selected examples illustrate the types of activities completed in the TMI-2 data qualification process and the importance of such a qualification effort. These details are documented in this report to facilitate implementation of similar process using data and examinations at the Daiichi Units 1, 2, and 3 reactors so that BWR-specific benefits can be obtained.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Target Housing Material Options

With gas cooling, heat transfer coefficients are low compared to water. The benefit of gas from a heat transfer point of view is that there is really no upper temperature limit for the coolant, as compared to water, which is limited ultimately by the critical point, and in practice the critical heat flux. In our case with parallel flow channels, water is limited to even lower operating limits by nucleate boiling. So gas can get as hot as the containment material will allow, but to get the density and heat transfer up to something reasonable, we must also increase pressure, thus increasing stress on the containment, namely the front and back faces. We are designing to ASME BPVC, which, for most materials allows a maximum stress of UTS/3. So we want the highest possible UTS. For reference, the front face stress in the 12 mm target at 300 psi was about 90 MPa. The inconel 718 allowable stress at 900°C is 1/3 of 517 or 172 MPa. So we are in a very safe place, but the uTS is dropping rapidly with temperature above 900°C. As we increase target diameter, the challenge will be to keep the stress down. We are probably looking at keeping the allowable at or above the present value, and at as high a temperature as possible.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2016

Fiber optic quench detection via optimized Rayleigh Scattering in high-field YBCO accelerator magnets

Yttrium barium copper oxide (YBCO) coated conductors are known for their ability to operate in the superconducting state at relatively high temperatures, even above the boiling point of liquid nitrogen (77 K). When these same conductors are operated at lower temperatures, they are able to operate in much higher magnetic fields than traditional superconductors like NiTi or Nb 3 Sn. Thus, YBCO superconducting magnets are one of the primary options for generating the high magnetic fields needed for future high energy physics devices. Due to slow quench propagation, quench detection remains one of the primary limitations to YBCO magnets. Fiber optic sensing, based upon Rayleigh scattering, has the potential for spatial resolution approaching the wavelength of light, or very fast temporal resolution at low spatial resolution, and a continuum of combinations in between. This project has studied, theoretically and experimentally, YBCO magnets and Rayleigh scattering quench detection systems to demonstrate feasibility of the systems for YBCO quench protection systems. Under this grant an experimentally validated 3D quench propagation model was used to accurately define the acceptable range of spatial and temporal resolutions for effective quench detection in YBCO magnets and to evaluate present-day and potentially improved YBCO conductors. The data volume and speed requirements for quench detection via Rayleigh scattering required the development of a high performance fiber optic based quench detection/data acquisition system and its integration with an existing voltage tap/thermo-couple based system. In this project, optical fibers are tightly co-wound into YBCO magnet coils, with the fiber on top of the conductor as turn-to-turn insulation. Local changes in the temperature or strain of the conductor are sensed by the optical fiber, which is in close thermal and mechanical contact with the conductor. Intrinsic imperfections in the fiber reflect Rayleigh backscattered laser signals that are shifted by the changes in the fiber that are induced by a local change in the YBCO temperature or strain. One goal of this project was to show that modern technology can be used to interrogate the signals from a (very expensive) YBCO magnet to detect an impending quench in time to protect it from self-destruction. The results show that Rayleigh-backscattering interrogated optical fibers (RIOF) have significant advantages over traditional techniques, including very high spatial resolution and the ability to detect a hot-spot well before the peak local temperature becomes so high that the conductor can be damaged. RIOF quench detection is intrinsically faster than voltage taps, and this intrinsic advantage is greater as the coil size and/or current margin increases. We describe the development and testing program performed under the grant.

43 PARTICLE ACCELERATORS↗
DOE OSTI2016

Characterization and Recovery of Rare Earths from Coal and By-Products

Coal is a precious resource, both in the United States and around the world. The United States has a 250-year supply of coal, and generates between 30 - 40% of its electricity through coal combustion. Approximately 1 Gt of coal has been mined annually in the US, although the 2015 total will likely be closer to 900 Mt (http://www.eia.gov/coal/production/quarterly/). Most of the coal is burned for power generation, but substantial quantities are also employed in the manufacture of steel, chemicals, and activated carbons. Coal has a positive impact upon many industries, including mining, power, rail transportation, manufacturing, chemical, steel, activated carbon, and fuels. Everything that is in the earth’s crust is also present within coal to some extent, and the challenge is always to utilize abundant domestic coal in clean and environmentally friendly manners. In the case of the rare earths, these valuable and extraordinarily useful elements are present within the abundant coal and coal by-products produced domestically and world-wide. These materials include the coals, as well as the combustion by-products such as ashes, coal preparation wastes, gasification slags, and mining by-products. All of these materials can be viewed as potential sources of rare earth elements. Most of the common inorganic lanthanide compounds, such as the phosphates found in coal, have very high melting, boiling, and thermal decomposition temperatures, allowing them to concentrate in combustion and gasification by-products. Furthermore, rare earths have been found in interesting concentrations in the strata above and below certain coal seams. Much of the recent research on coal utilization in the United States has focused upon the capture of pollutants such as acid gases, particulates, and mercury, and the greenhouse gas carbon dioxide. The possible recovery of rare earth and other critical elements from abundant coal and by-products is an exciting new research area, representing a dramatic paradigm shift for coal.

01 COAL, LIGNITE, AND PEAT↗
DOE OSTI2015

Development of an Efficient Meso- scale Multi-phase Flow Solver in Nuclear Applications

The proposed research aims at formulating a predictive high-order Lattice Boltzmann Equation for multi-phase flows relevant to nuclear energy related application - namely, saturated and sub-cooled boiling in reactors, and liquid- liquid mixing and extraction for fuel cycle separation. An efficient flow solver will be developed based on the Finite Element based Lattice Boltzmann Method (FE- LBM), accounting for phase-change heat transfer and capable of treating multiple phases over length scales from the submicron to the meter. A thermal LBM will be developed in order to handle adjustable Prandtl number, arbitrary specific heat ratio, a wide range of temperature variations, better numerical stability during liquid-vapor phase change, and full thermo-hydrodynamic consistency. Two-phase FE-LBM will be extended to liquid–liquid–gas multi-phase flows for application to high-fidelity simulations building up from the meso-scale up to the equipment sub-component scale. While several relevant applications exist, the initial applications for demonstration of the efficient methods to be developed as part of this project include numerical investigations of Critical Heat Flux (CHF) phenomena in nuclear reactor fuel bundles, and liquid-liquid mixing and interfacial area generation for liquid-liquid separations. In addition, targeted experiments will be conducted for validation of this advanced multi-phase model.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Assessment of Gasoline Additive Containing Ditert-butoxypropanol

The Fuels, Engines, and Emissions Research Center completed analysis and testing of the CPS Powershot gasoline additive under the auspices of the Department of Energy’s Technical Assistance for US Small Businesses in Vehicle Technologies. Gas chromatography-mass spectrometry (GC-MS) was used to quantify the makeup of the additive, finding a predominance of 2,3-Ditert-Butoxypropanol, also known as Glyceryl Di-Tert-Butyl Ether (GTBE). Blends of the additive at 2 and 4 volume percent were subjected to a number of standard ASTM tests, including Research Octane Number, Motor Octane Number, distillation, and vapor pressure. Results show a high boiling range and low vapor pressure for the additive, and a very modest octane boosting effect in gasoline with and without ethanol.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2016

Development and Testing of CTF to Support Modeling of BWR Operating Conditions

This milestone supports developing and assessing COBRA-TF (CTF) for the modeling of boiling water reactors (BWRs). This is achieved in three stages. First, a new preprocessor utility that is capable of handling BWR-specic design elements (e.g., channel boxes and large water rods) is developed. A previous milestone (L3:PHI.CTF.P12.01) led to the development of this preprocessor capability for single assembly models. This current milestone expands this utility so that it is applicable to multi-assembly BWR models that can be modeled in either serial or parallel. The second stage involves making necessary modications to CTF so that it can execute these new models. Specically, this means implementing an outer-iteration loop, specic to BWR models, that equalizes the pressure loss over all assemblies in the core (which are not connected due to the channel boxes) by adjusting inlet mass ow rate. A third stage involves assessing the standard convergence metrics that are used by CTF to determine when a simulation is steady-state. The nal stage has resulted in the implementation of new metrics in the code that give a better indication of how steady the solution is at convergence. This report summarizes these eorts and provides a demonstration of CTF's BWR-modeling capabilities. CASL-U-2016-1030-000

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Low temperature dissolution flowsheet for plutonium metal

The H-Canyon flowsheet used to dissolve Pu metal for PuO 2 production utilizes boiling HNO 3 . SRNL was requested to develop a complementary dissolution flowsheet at two reduced temperature ranges. The dissolution and H 2 generation rates of Pu metal were investigated using a dissolving solution at ambient temperature (20-30 °C) and for an intermediate temperature of 50-60 °C. Additionally, the testing included an investigation of the dissolution rates and characterization of the off-gas generated from the ambient temperature dissolution of carbon steel cans and the nylon bags that contain the Pu metal when charged to the dissolver.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Severe Accident Analysis Of BWR Core Fueled With UO 2 /FeCrAl with Updated Materials and Melt Properties From Experiments

Iron-chromium-aluminum (FeCrAl) alloys are being considered as advanced fuel cladding concepts with enhanced accident tolerance. FeCrAl alloys have slower oxidation kinetics and higher strength compared to Zr-based alloys at high temperatures. FeCrAl could be used for fuel cladding and spacer or mixing vane grids in light water reactors and/or as channel box material in boiling water reactors (BWRs). There is a need to assess the potential gains afforded by the FeCrAl accident tolerant fuel concept over the existing zirconium based materials employed today.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

System Study: High-Pressure Coolant Injection 1998-2014

This report presents an unreliability evaluation of the high-pressure coolant injection system (HPCI) at 25 U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2014 for selected components were obtained from the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The unreliability results are trended for the most recent 10 year period, while yearly estimates for system unreliability are provided for the entire active period. No statistically significant increasing or decreasing trends were identified in the HPCI results.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

System Study: High-Pressure Core Spray 1998-2014

This report presents an unreliability evaluation of the high-pressure core spray (HPCS) at eight U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2014 for selected components were obtained from the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The unreliability results are trended for the most recent 10 year period, while yearly estimates for system unreliability are provided for the entire active period. No statistically significant increasing or decreasing trends were identified in the HPCS results.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

System Study: Isolation Condenser 1998-2014

This report presents an unreliability evaluation of the isolation condenser (ISO) system at four U.S. boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2014 for selected components were obtained from the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The unreliability results are trended for the most recent 10 year period, while yearly estimates for system unreliability are provided for the entire active period. No statistically significant increasing trends were identified. A statistically significant decreasing trend was identified for ISO unreliability. The magnitude of the trend indicated a 1.5 percent decrease in system unreliability over the last 10 years.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

System Study: Reactor Core Isolation Cooling 1998-2014

This report presents an unreliability evaluation of the reactor core isolation cooling (RCIC) system at 31 U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2014 for selected components were obtained from the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The unreliability results are trended for the most recent 10 year period, while yearly estimates for system unreliability are provided for the entire active period. No statistically significant trends were identified in the RCIC results.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

RELAP-7 Pressurizer Component Development Updates

RELAP-7 is a nuclear systems safety analysis code being developed at the Idaho National Laboratory (INL). RELAP-7 development began in 2011 to support the Risk Informed Safety Margins Characterization (RISMC) Pathway of the Light Water Reactor Sustainability (LWRS) program. The overall design goal of RELAP-7 is to take advantage of the previous thirty years of advancements in computer architecture, software design, numerical methods, and physical models in order to provide capabilities needed for the RISMC methodology and to support nuclear power safety analysis. The code is being developed based on Idaho National Laboratory’s modern scientific software development framework – MOOSE (the Multi-Physics Object-Oriented Simulation Environment). The initial development goal of the RELAP-7 approach focused primarily on the development of an implicit algorithm capable of strong (nonlinear) coupling of the dependent hydrodynamic variables contained in the 1-D/2-D flow models with the various 0-D system reactor components that compose various boiling water reactor (BWR) and pressurized water reactor nuclear power plants (NPPs). As part of the efforts to expand the capability for PWR simulation, an equilibrium single-region pressurizer model has been implemented in RELAP-7. The pressurizer component can simulate pressure and water level change through insurge, spray, and heating processes. Two simple tests – one for insurge process and another for outsurge process – have been reported to demonstrate and verify the functions of the pressurizer model. The typical single-phase PWR system model presented in the first RELAP-7 milestone report has been updated, as part of system level test for the new pressurizer model. The updated PWR system model with the pressurizer component can be used for more realistic transient simulations. The addition of the equilibrium single-region pressurizer model represents the first step of developing a suite of pressurizer models with different levels of fidelities. Non-equilibrium two-region two-zone model and non-equilibrium five-region three zone model will be implemented in the near future. Cross comparison and validation will be pursued when higher fidelity pressurizer models are completed.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Studies of Heavy-Ion Reactions and Transuranic Nuclei

Studies of heavy-ion reactions and transuranic nuclei performed by the University of Rochester Nuclear Science Research Group have been successful in furthering experimental systematics and theoretical understanding of the behavior of nuclear systems excited to their limits of stability. The theoretical results explain specifically the “boiling” and “vaporization” of atomic nuclei, but are more generally applicable to isolated, quantal many-particle systems which, under thermal or mechanical stresses, all disintegrate by evaporation, via surface cluster emission, or via fission-like processes. Accompanying experimental investigations by the group have demonstrated several new types of dynamical instability of nuclei: In central, “head-on” collisions, target nuclei exhibit limited ability to stop energetic projectile nuclei and to dissipate the imparted linear momentum. Substantial matter overlap (“neck”) between projectile and target nuclei, which is observed at elevated collision energies, can be stretched considerably and break at several places simultaneously. These results provide new testing grounds for microscopic theory of the cohesion of nuclear matter. This property has remained elusive, even though the elementary nucleon-nucleon forces are well known since some time. Technical R&D has resulted in a detailed characterization of a novel plastic material, which can now be used in the design of sensitive diagnostic systems for various types of radio-activity. Innovative application of powerful laser systems has produced intense, controllable sources of exotic particle radioactivity for nuclear investigations. Several students have received their Ph.D. degree in experimental nuclear science for their work on basic nuclear research or R&D projects.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗