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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 91 records · Page 5
All sourcesNASA NTRSDOE OSTI
NASA NTRS1978

General aviation internal-combustion engine research programs at NASA-Lewis Research Center

An update is presented of non-turbine general aviation engine programs. The program encompasses conventional, lightweight diesel and rotary engines. It's three major thrusts are: (1) reduced SFC's; (2) improved fuels tolerance; and (3) reduced emissions. Current and planned future programs in such areas as lean operation, improved fuel management, advanced cooling techniques and advanced engine concepts, are described. These are expected to lay the technology base, by the mid to latter 1980's, for engines whose life cycle fuel costs are 30 to 50% lower than today's conventional engines.

Willis, E. A.↗
NASA NTRS1978

General aviation internal combustion engine research programs at NASA-Lewis Research Center

An update is presented of non-turbine general aviation engine programs underway at the NASA-Lewis Research Center in Cleveland, Ohio. The program encompasses conventional, lightweight diesel and rotary engines. Its three major thrusts are: (a) reduced SFC's; (b) improved fuels tolerance; and (c) reducing emissions. Current and planned future programs in such areas as lean operation, improved fuel management, advanced cooling techniques and advanced engine concepts, are described. These are expected to lay the technology base, by the mid to late 1980's, for engines whose life cycle fuel costs are 30 to 50% lower than today's conventional engines.

Willis, E. A.↗
DOE OSTI2025

Core design and performance of the Westinghouse lead fast reactor with UO 2 and MOX configurations

For this work, Westinghouse partnered with Argonne National Laboratory to design, model and optimize UO 2 - and MOX-fueled core designs for a medium size (950 MWt) Lead Fast Reactor that was pursued by Westinghouse. Using Argonne’s suite of reactor analysis codes together with Westinghouse fuel cost economic models, thousands of candidate cores were considered to achieve the economics-optimized cores presented in this paper. This optimization process considered detailed reactor physics, fuel performance, transient performance, and fuel economics models. The reactor performance of the resulting optimized UO 2 - and MOX-fueled core designs are described and compared in this paper. Both cores show fuel performance and transient behavior that is considered acceptable for the optimization presented herein, while further testing campaigns on material performance in high-temperature liquid lead will be required to confirm acceptability at the operating conditions chosen. A multi-batch strategy was selected for the UO 2 core for best fuel utilization with minimum fuel inventory costs. A single-batch fuel management was instead selected for the MOX core to maximize cycle length and minimize the impact of the longer refueling outage resulting from the higher decay heat of the discharged MOX fuel relative to the discharged UO 2 fuel, requiring a longer cooling time before dry-lift of discharged fuel could take place.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

The STAT7 Code for Statistical Propagation of Uncertainties in Steady State Thermal Hydraulics Analysis of Plate Fueled Reactors

The STAT7 software was developed to perform steady-state thermal hydraulic analyses. Application of the software is for non-power research and test reactors, including conversion to low-enriched uranium fuel of U.S. High-Performance Research Reactors such as MITR-II. Since it can be necessary to repeat analysis during fuel reloading, STAT7 accommodates flexibility in analyzing many realistic aspects of reactor fuel management. STAT7 uses a Monte Carlo approach to model common fabrication parameters and other key reactor analysis uncertainties required for research and test reactor thermal hydraulic analyses. These safety calculations are ultimately intended to protect against high fuel plate temperatures due to critical heat flux or departure from nucleate boiling or onset of flow instability; but additional margin is obtained by basing the limiting safety settings on avoiding onset of nucleate boiling. STAT7 can simultaneously analyze all of the axial nodes of all of the fuel plates and all of the coolant channels for one lateral stripe of a fuel element. The stripes run the length of the fuel, from the bottom to the top. Power splits are calculated for each axial node of each plate to determine how much of the power goes out each face of the plate. By running STAT7 multiple times, full core analysis can be performed by analyzing the margin to onset of nucleate boiling and onset of flow instability for each axial node of each stripe of each plate of each element in the core.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1979

Commercial phosphoric acid fuel cell system technology development

Reducing cost and increasing reliability were the technology drivers in both the electric utility and on-site integrated energy system applications. The longstanding barrier to the attainment of these goals was materials. Differences in approaches and their technological features, including electrodes, matrices, intercell cooling, bipolar/separator plates, electrolyte management, fuel selection, and system design philosophy were discussed.

Prokopius, P. R.↗
DOE OSTI2020

Scoping Analysis of Sodium Cooled Fast Spectrum Test Reactor Cores

A team from the national laboratories conducted a study that assessed advanced reactor technology options and the irradiation needs of a broad group of stakeholders. This identified a gap in the fast neutron irradiation capability - the need for a test reactor that can deliver high neutron fluxes of fast neutrons. The study compared fast and thermal neutron irradiation technology options based on technology maturity and the ability to meet strategic objectives, along with pathways to deployment and tradeoffs in mission. It also provided a recommendation on specific technology options to address the current gap in fast neutron irradiation capabilities. In this context, a detailed scope of work was prepared that pertains to a trade study focusing on the core of a fast neutron spectrum test reactor. This report presents the main results obtained by a team of Argonne National Laboratory (ANL) and Idaho National Laboratory (INL) scientists. It aims at developing an overall understanding of how the various areas important to the design and successful operation of such a facility interact with one another. Developing a sense of the relevant sensitivities at a systems level creates an understanding of where and how margins may be available to improve performances (e.g., flux level, capacity factor, and thermal hydraulic limits). Emphasis is given to the reactor (core design options and safety) and its fuel cycle (fuel supply and fabrication, as well as spent fuel management).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Safely probing the chemistry of Chernobyl nuclear fuel using micro-focus X-ray analysis

Detailed chemical analysis of the solidified molten fuel still residing in the stricken Chernobyl reactor unit 4 are inferred using multi-modal micro-focus X-ray analysis of a low-radioactivity proxy. A fascinating mixture of molten UO 2 , nuclear fuel cladding, concrete, stainless steel and other nuclear reactor components, these materials behaved like lava, solidifying to form a complex, highly radioactive glass-ceramic. Using element-specific chemical probes (micro-X-ray fluorescence and X-ray absorption spectroscopy), coupled with micro-diffraction analysis, the crystalline phase assemblage of simulants of these heterogeneous materials was established, which included “chernobylite” and a range of compositions in the (U 1-x Zr x )O 2 solid solution. Novel insight to nuclear accident fuel chemistry was obtained by establishing the oxidation state and local coordination of uranium not only in these crystalline phases, but uniquely in the amorphous fraction of the material, which varied depending on the history of the nuclear lava as it flowed through the reactor. This study demonstrates that micro-focus X-ray analysis of very small fractions of material can yield rich chemical information, which can be applied to nuclear-melt down materials to aid decommissioning and nuclear fuel management at nuclear accident sites.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

LaSalle Leaking Pin PIE Proposal

Determine the most probable root cause of the LaSalle fuel rod failure, distinguish primary failure mechanisms from post-failure degradation, and assess whether the condition is isolated or indicative of a broader issue affecting similar rods or assemblies. Results will support decisions related to core operation, fuel management, and corrective actions.

LaSalle↗
NASA NTRS1981

Status of commercial phosphoric acid fuel cell system development

In both the electric utility and onsite integrated energy system applications, reducing cost and increasing reliability are the main technology drivers. The longstanding barrier to the attainment of these goals, which manifests itself in a number of ways, was materials. The differences in approach among the three major participants (United Technologies Corporation, Westinghouse Electric Corporation/Energy Research Corporation, and Engelhard Industries) and their unique technological features, including electrodes, matrices, intercell cooling, bipolar/separator plates, electrolyte management, fuel selection and system design philosophy are discussed.

Warshay, M.↗
NASA NTRS1981

Status of commercial phosphoric acid fuel cell system development

A review of the current commercial phosphoric acid fuel cell system development efforts is presented. In both the electric utility and on-site integrated energy system applications, reducing cost and increasing reliability are important. The barrier to the attainment of these goals has been materials. The differences in approach among the three major participants are their technological features, including electrodes, matrices, intercell cooling, bipolar/separator plates, electrolyte management, fuel selection and system design philosophy.

Warshay, M.↗
DOE OSTI2023

Materials and Fuels Complex Operations Management Improvement Strategy for Fiscal Year 2023

The Materials and Fuels Complex (MFC) has experienced substantial growth in terms of staff, research, and production in recent years. MFC operational performance has effectively kept pace with this growth. However, to capture the continuous improvement actions needed to improve effectiveness and increase the efficiency of our management systems, a broad operations management strategy is necessary. The MFC Operations Management Improvement (OMI) Strategy is complementary to the MFC Five-Year Mission and Investment Strategies and the MFC Management Plan. The OMI strategy is structured to address the management systems outlined in the Nuclear Facility Management Standard Operations Model. Selected management systems are evaluated independently in chapters that describe the prior 5 years of performance improvement, a description of improvement actions for the MFC staff that directly perform within or contribute to the management system, process improvements, and any needed equipment improvements. Chapter selection is based on a management system’s need for improvement. In some cases, management systems may be combined in a single chapter. Additional chapters are added to address subject areas not formally described in the standard operations model.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

INL Site ARG-US Implementation – FY22 Activities and FY23 Plans

The Packaging Certification and Life Cycle Management program at Argonne National Laboratory (ANL) developed a suite of monitoring systems collectively referred to here as ARG-US. ARG-US provides necessary data for nuclear facility and system operation and maintenance, and has been previously demonstrated in hot cells, radioactive material (RAM) storage areas, and RAM shipment trucks. It has also been installed directly on RAM storage and shipment containers. ARG-US offers some unique advantages over other commercially available systems by using wireless data connections, battery power supplies, and customizable monitoring methods. Idaho National Laboratory (INL) has been tasked with investigating applications for ARG-US at INL site facilities, which are operated by several different contractors. The initial investigation scope centered on CPP-603 Irradiated Fuel Storage Facility in relation to the upcoming Department of Energy (DOE) Spent Nuclear Fuel (SNF) Packaging Demonstration. The investigation has been led by the Used Fuel Management Department in INL’s Nuclear Science and Technology (NS&T) directorate. This investigation is expected to recommend INL site processes or facilities in which to implement ARG-US systems. INL has engaged local site technical and oversight representatives, security, nuclear safeguards, and program management personnel to identify good candidates for ARG-US test implementations. INL has identified the following high-level goals for any INL site implementation of ARG-US: provide unique testing environments, prompt development of new monitoring methods/techniques for the ARG-US suite, and acquire useful monitoring for the user facility. As a result, INL suggests three program areas for further investigation: CPP-603 Fuel Handling Cave (FHC), legacy mixed waste storage systems at Idaho Nuclear Technology & Engineering Center (INTEC) known as the “Tank Farm,” and periodic and emergency environmental monitoring. This progress report relates activities undertaken in this investigation, describes the preliminary areas of interest for limited scope ARG-US testing or implementation, and relays expected actions for completing the task scope. This report offers an opportunity to the program sponsor, technical leads at ANL, and INL site representatives to give feedback on the initial assessment and make recommendations on the forthcoming activities.

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

Investigation of INL Site Implementation of ARG-US

The Packaging Certification and Life Cycle Management department at Argonne National Laboratory (ANL) developed a suite of monitoring systems collectively referred to here as ARG-US. ARG-US is used to provide necessary data for nuclear facility and system operation and maintenance. ARG-US has been previously demonstrated in hot cells, radioactive material (RAM) storage areas, and RAM shipment trucks. It has also been installed directly on RAM storage and shipment containers. ARG-US offers some unique advantages over other commercially available systems by using wireless data connections, battery power supplies, and customizable monitoring methods. Idaho National Laboratory (INL) has been tasked with investigating applications for ARG-US at INL site facilities, which are operated by several different contractors. The initial investigation scope centered on CPP-603 Wet & Dry Fuel Storage Facility in relation to the upcoming DOE Spent (SNF) Packaging Demonstration. The investigation has been led by the Used Fuel Management Department in the Nuclear Science and Technology Directorate. This investigation is expected to recommend processes or facilities at the INL site at which to implement ARG-US systems. INL has engaged local site technical and oversite representatives, security, nuclear safeguards, and program management to identify good candidates for ARG-US test implementations. INL has identified high-level goals for any INL site implementation of ARG-US: provide unique testing environments, prompt development of new monitoring methods/techniques for the ARG-US suite, and acquire useful monitoring for the user facility. To these ends, INL suggests 3 program areas for further investigation: CPP-603 Fuel Handling Cave (FHC), legacy mixed waste storage systems at Idaho Nuclear Technology & Engineering Center (INTEC) known as the “tank farm”, and periodic and emergency environmental monitoring. This progress report relates activities undertaken in this investigation, describes the preliminary areas of interest for limited scope ARG-US testing or implementation, and relays expected actions needed to complete the task scope. This report offers an opportunity to the program sponsor, technical leads at ANL, and INL site representatives to give feedback on the initial assessment and make recommendations on the forthcoming activities.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

A strategy for automated core design to increase economic viability and minimize fuel fragmentation, relocation, and dispersal susceptibility in high-burnup cores

The nuclear industry aims to increase the cycle length of pressurized water reactors from 18 to 24 months to increase power plant capacity factors and economic viability. These cycle length extensions will inherently require fuel rods to exceed the current peak rod average burnup limit of 62 GWd/MTU. A chief concern of operating beyond the current burnup limit is the fuel fragmentation, relocation, and dispersal (FFRD) phenomenon in which pulverized fuel fragments can axially relocate and escape through a burst in the cladding formed during a loss-of-coolant accident. In this work, we demonstrate an approach for automating core design employing an optimization tool based on a penalty-free, parallel simulated annealing algorithm to produce pressurized water reactor core designs with two different optimization objectives. The two objectives were to produce core designs with (1) mitigated FFRD susceptibility while achieving 24-month cycle lengths (2) maximum cycle length with no regard for the likelihood of FFRD. Batch size was considered in tandem with both cases to maximize economic viability. The PARCS nodal model was the primary reactor physics tool used in the optimizations and used nuclear cross sections calculated with 2D Polaris lattice physics models. Reactor performance and safety characteristics of the optimized cores were verified using high-fidelity Virtual Environment for Reactor Applications models. The core designs produced by the optimization tool are compared with each other and to a high-burnup core design produced and analyzed in previous works to highlight the fuel management strategies that may enhance high-burnup reactor safety and economic viability. The optimized cores satisfied their respective objective functions, producing a maximum cycle length of 720 effective full-power days in one core design and one that may reduce FFRD susceptibility by up to 50% based on the first-order approximation to FFRD risk formulated in this work. The optimized cores met most constraints but exceeded the hot channel factor limit, especially in FFRD cases where fresh fuel carried more power. Furthermore, this highlights the need for future lattice-level optimizations and broader assembly options.

Cycle length↗
NASA NTRS1979

Commercial phosphoric acid fuel cell system technology development

A review of the current commercial phosphoric acid fuel cell system technology development efforts is presented. In both the electric utility and on-site integrated energy system applications, reducing cost and increasing reliability are the technology drivers at this time. The longstanding barrier to the attainment of these goals, which manifests itself in a number of ways, has been materials. The differences in approach among the three major participants (United Technologies Corporation (UTC), Westinghouse Electric Corporation/Energy Research Corporation (ERC), and Engelhard Industries) and their unique technological features, including electrodes, matrices, intercell cooling, bipolar/separator plates, electrolyte management, fuel selection and system design philosophy are discussed.

Prokopius, P. R.↗
NASA NTRS1982

From Lift-Off to Light-Off

Shuttle's propellant measurement system is produced by Simmonds Precision. Company has extensive experience in fuel management systems and other equipment for military and commercial aircraft. A separate corporate entity, Industrial Controls Division was formed due to a number of non-aerospace spinoffs. One example is a "custody transfer" system for measuring and monitoring liquefied natural gas (LNG). LNG is transported aboard large tankers at minus 260 degrees Fahrenheit. Value of a single shipload may reach $15 million. Precision's LNG measurement and monitoring system aids accurate financial accounting and enhances crew safety. Custody transfer systems have been provided for 10 LNG tankers, built by Owing Shipbuilding. Simmonds also provided measurement systems for several liquefied petroleum gas (LPG) production and storage installations. Another spinoff developed by Simmonds Precision is an advanced ignition system for industrial boilers that offers savings of millions of gallons of fuel, and a computer based monitoring and control system for improving safety and reliability in electrical utility applications. Simmonds produces a line of safety systems for nuclear and non-nuclear electrical power plants.

Source record↗
DOE OSTI2022

Nuclear data uncertainty quantification for the nuclide inventory of a Calvert Cliffs spent fuel sample

The impact of nuclear data cross section uncertainties and covariance matrices on the nuclide vector of spent nuclear fuel was investigated; This exercise was carried out for the Calvert Cliffs fuel assembly D047 benchmark available in the SFCOMPO database. Sample P irradiated in rod MKP109 for 4 cycles up to a burnup of approximately 44 GWd/MTU was selected for the analysis. Nuclear data uncertainties were taken from the most recent libraries released by evaluation projects JEFF, ENDF/B and JENDL, and were propagated using the SANDY code via a stochastic sampling approach. This paper provides a quantification of the uncertainty on the concentration of several actinides and fission products relevant for spent fuel management. Uncertainties generally below 5 % were predicted for the concentrations of most of the uranium, neptunium and plutonium isotopes relevant for SNF applications. Curium isotopes carry larger uncertainties that might exceed 10 %. The contribution of cross section uncertainties on the concentrations of fission products was found to be marginal with the exception of a few nuclides. These results can be significantly affected by the lack of evaluated covariance matrices for the capture cross section of several fission products. Burnup tracers such as {sup 148}Nd and {sup 137}Cs have negligible uncertainties because of the power normalisation imposed in every stochastic calculation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗