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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 1,459 records · Page 81
All sourcesNASA NTRSDOE OSTI
NASA NTRS1989

Application of biocatalysts to Space Station ECLSS and PMMS water reclamation

Immobilized enzyme reactors have been developed and tested for potential water reclamation applications in the Space Station Freedom Environmental Control and Life Support System (ECLSS) and Process Materials Management System (PMMS). The reactors convert low molecular weight organic contaminants found in ECLSS and PMMS wastewaters to compounds that are more efficiently removed by existing technologies. Demonstration of the technology was successfully achieved with two model reactors. A packed bed reactor containing immobilized urease was found to catalyze the complete decomposition of urea to by-products that were subsequently removed using conventional ion exchange results. A second reactor containing immobilized alcohol oxidase showed promising results relative to its ability to convert methanol and ethanol to the corresponding aldehydes for subsequent removal. Preliminary assessments of the application of biocatalysts to ECLSS and PMMS water reclamation sytems are presented.

Jolly, Clifford D.↗
NASA NTRSDate not supplied

Kinetic Model Development of the Carbon Vapor Deposition Reactor to Predict Performance Versus Design

To enable complete closure of a Sabatier-based atmosphere revitalization system where oxygen recovery is predicated by carbon dioxide reduction to methane and water, there must be a means of recovering the hydrogen from the byproduct methane. Honeywell’s carbon vapor deposition (CVD) reactor is a potential technology that provides a thermochemical means by which methane decomposes at high temperatures (> 1,000 °C) into hydrogen, solid carbon, and a small percentage of hydrocarbons. This advanced methane pyrolysis reactor utilizes a carbon fiber substrate to capture and immobilize the solid pyrolytic carbon rather than allowing continuation toward undesirable soot. The CVD reactor is a highly dynamic system whose characteristic geometry changes while operating. The reactor model was developed to provide predictive performance assessment capabilities at different operating conditions, geometries, and configurations to aid in technology development. Modeling of the CVD reactor considered the multitude of simultaneous and interdependent physicochemical processes that are entailed, including: (1) the homogeneous gas-phase pyrolysis reactions, (2) the heterogeneous deposition reactions, and (3) the transport of gaseous species in the reactor domains. A 1-D CVD reactor model was developed based on a general packed-bed, axial-dispersion plug flow reactor model with a solid substrate balance to account for carbon deposition. The model includes 59 species and 243 reactions constituting the methane pyrolysis reaction network, carbon deposition, and interconversions of oxygenate species. The reactor model was compared to brassboard test data, where it was shown to be able to capture the pertinent physicochemical phenomena as evidenced by its agreement with short-duration run data (i.e., looking at different temperature and flow rate conditions), as well as trend agreement with the brassboard reactor’s long-duration test data. Recommendations are provided on further model development steps as well as its use cases.

Air Revitalization↗
DOE OSTI2022

Numerical study of multi-component flow and mixing in a scaled fission product venting system

Numerical simulation of isothermal multi-component gas flow was conducted to study fluid flow and mixing in a scaled low-pressure, low-temperature test facility for the fission product venting system (FPVS) of a gas cooled fast reactor (GFR). The geometry of FPVS test facility was an open loop including gradual expansion coupling and two 90°pipe elbows. First, the validation of large eddy simulation (LES) wall-adapting local eddy-viscosity (WALE) turbulent model was performed for transitional flow in a circular pipe with gradual expansion. The reactingFOAM solver in OpenFOAM v8 was employed. By introducing appropriate turbulence disturbance at the flow inlet, such as turbulence intensity, the numerical results showed good agreement in (1) the velocity profile downstream of the pipe expansion measured by particle image velocimetry (PIV) in this study, and (2) the reattachment length reported in the literature. Using this validated model, the numerical simulation of flow and component mixing was performed for the FPVS with an inlet flowrate corresponding to a Reynolds number of 2,400 to investigate the flow behavior and component mixing. Standard deviation of the mass fraction of component, referred to as absolute mixing index (AMI), was calculated to quantify the mixing of components, showing that the component mixing length determined by AMI is related to the reattachment length downstream of the expansion. Additionally, this study was able to identify the best location in the FPVS test facility to measure the velocity and concentration profies where multicomponent flow is well mixed.

15 GEOTHERMAL ENERGY↗
NASA NTRS2007

Liquid Metal Pump Technologies for Nuclear Surface Power

Multiple liquid metal pump options are reviewed for the purpose of determining the technologies that are best suited for inclusion in a nuclear reactor thermal simulator intended to test prototypical space nuclear system components. Conduction, induction, and thermoelectric electromagnetic pumps are evaluated based on their performance characteristics and the technical issues associated incorporation into a reactor system. The thermoelectric electromagnetic pump is recommended for inclusion in the present system based on favorable quantitative and qualitative measures relative to the other options under consideration.

Polzin, Kurt A.↗
DOE OSTI2020

Zeolite Membrane Reactor for Pre-Combustion Carbon Dioxide Capture

Water-gas shift (WGS) reaction followed by carbon dioxide (CO 2 ) separation is a critical step in the integrated gasification combined cycle (IGCC) process for fossil-fuel-fired electrical power generation with CO 2 capture. To intensify the IGCC process hydrogen-permselective zeolite membrane reactor offers promise to replace the conventional energy-intensive fixed-bed reactors and solvent-based CO 2 capture units. The objectives of this project were to develop a bench-scale zeolite membrane reactor (total membrane area: 932 cm 2 for a 21-tube membrane bundle) for the water-gas-shift reaction of raw syngas from an oxygen-blown coal-gasifier for H 2 production with simultaneous CO 2 separation at the capability of about 2 kilograms H 2 per day (equivalent to 2 kW IGCC power plant) and to demonstrate significant progress toward achieving overall performance goal of 90% CO 2 capture rate with 95% CO 2 purity at the cost of electricity 30% less than the baseline carbon capture approaches. This report summarizes results obtained in this project on scaling up the zeolite membrane reactor by a factor of 200 in membrane area, tests of the bench-scale zeolite membrane reactor for the water-gas-shift reaction at high temperature and high-pressure, and techno-economic analysis of the integration of the zeolite membrane reactor in IGCC power plant for the electrical generation with CO 2 capture.

20 FOSSIL-FUELED POWER PLANTS↗
NASA NTRS2007

Liquid-Metal Pump Technologies for Nuclear Surface Power

Multiple liquid-metal pump options are reviewed for the purpose of determining the technologies that are best suited for inclusion in a nuclear reactor thermal simulator intended to test prototypical space nuclear system components. Conduction, induction, and thermoelectric electromagnetic pumps are evaluated based on their performance characteristics and the technical issues associated with incorporation into a reactor system. The thermoelectric pump is recommended for inclusion in the planned system at NASA MSFC based on its relative simplicity, low power supply mass penalty, flight heritage, and the promise of increased pump efficiency over earlier flight pump designs through the use of skutterudite thermoelectric elements.

Polzin, K. A.↗
DOE OSTI2021

Preliminary Procedures and Acceptance Criteria for in-situ Structural Materials Surveillance for MSR

This report describes the initial development of the technical basis for a materials surveillance technology that, when fully developed and validated, can be used by stakeholders to develop and implement a materials surveillance program to manage materials degradation during reactor operations. The procedures described in this report aim to bound the detrimental effects of the reactor environment on the creep-fatigue life of the reactor components through materials surveillance. Specifically, the report describes procedures for selecting critical locations within the reactor components for surrogate materials surveillance and the sizing and placements of passively-actuated materials surveillance test articles under development by the US Department of Energy, Advanced Reactor Technologies Program. These test articles apply cyclic thermomechanical load to a surrogate sample of the component structural material passively through thermal expansion mismatch within the articles, without requiring any penetrations in the reactor coolant boundary. The procedures can be used to size materials surveillance test articles so that they will fail before the corresponding structural component. These articles would be exposed to the component operating environment, located to experience a bounding radiation fluence and representative coolant exposure. They can be monitored during reactor operations to ensure that they have not failed under the combination of creep-fatigue damage and environmental degradation. The report describes the key technical decisions and the corresponding rational for using these procedures to provide advanced warning of any impending material failure. An appendix describes the technical basis and the procedure for sizing a family of passively-actuated test articles. Finally, the report describes future work needed to complete the technical basis underlying this materials surveillance technology.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Elevated Temperature Graphite Mechanical Testing

High purity graphite will be used for core components within most High Temperature Reactor (HTR) designs. Several "high tech” industries currently utilize synthetic, high-purity, commercially available graphite components to fabricate photovoltaic cells, semi-conductors, optical fibers, and other high value electronic industry products. New advanced HTR designs are also interested in using these graphite grades for long-term, internal core component applications. Consequently, the US Department of Energy, Advanced Reactor Technologies (DOE-ART) program has spent several years testing different high purity graphite grades for potential use within these new nuclear reactor designs. A significant part of that effort has been in the development and improvement of American Society for Testing and Materials (ASTM) test standards specifically for nuclear graphite grades. Nearly all ASTM test standards have either been developed or improved by the DOE-ART program over the past 25 years. The ART program continues to assist in the development of new ASTM test standards in support of the future commercial HTR fleet currently being designed and built in the USA. The newest effort undertaken by ART is the development of high temperature mechanical testing practices that may be acceptable for a future ASTM test standard (or guide) for this critical material property measurement.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Results of the KRUSTY Warm Critical Experiments

The Kilowatt Reactor Using Stirling TechnologY (KRUSTY) was a prototypic nuclear-powered test of a 5-kW(thermal) Kilopower space reactor. This paper presents results from the KRUSTY warm critical experiments, which were completed prior to the final system test. The first set of criticals comprised cold or zero-power criticals; i.e., the core was not heated by fission power. These were followed by three warm criticals, where fission power heated the core to 200°C, 300°C, and 450°C, respectively. These criticals provided the data, confidence, and regulatory framework that were needed to proceed with the KRUSTY nuclear system test. The criticals also provided valuable data for the benchmarking of codes applicable to all nuclear systems. Finally, a comparison of KRUSTY results to pretest predictions is provided, and overall, the models matched the experimental results very closely.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Thermochemical Nonequilibrium Modeling in a Continuous-Galerkin, Finite-Element Framework

The presented work discusses the implementation, verification, and validation of Park's two-temperature model in a scalable, computational fluid dynamics (CFD) code developed at the US Department of Energy's Oak Ridge National Laboratory (ORNL). The implementation of Park's two-temperature model was verified through 0D test cases involving an adiabatic reactor and a nitrogen thermal bath. The implementation was then validated through comparisons with other validated CFD codes and experimental data on a hypersonic cylinder and double cones. These are standard benchmark test cases for thermochemical non-equilibrium (TCNE) modeling, and all data are shared publicly. The verification and validation results showed that ORNL's in-house CFD code could model complex, high-speed flow problems with and without TCNE modeling. This work is essential for future research involving 3D shock wave/boundary layer interactions (SBLIs).

Nutter, Nicole↗
DOE OSTI2025

MARVEL Utilization Plan

This report provides a high-level overview of the operational plan for the MARVEL microreactor. The main focus is on discussing testing and application demonstration opportunities that leverage the reactor. With the reactor rapidly progressing towards demonstration, along with the short operational window (2 years), it was deemed critical to start establishing a basis for how stakeholders can engage with the program and leverage the reactor as a testbed. The report discusses potential opportunities for using MARVEL leveraging from data/design access, novel controls testing, novel nuclear-electric applications, to novel nuclear-heat applications. It also provides guidance for interested stakeholders on potential funding opportunities that can be pursued to support tests during the operational lifetime of the reactor. The report also discusses the recommended strategy for outreach and the organization structure to review and down select project. It is important to emphasize that this is intended to be a living document with revisions issued as the MARVEL demonstration timeline evolves. Similarly, the framework for engagement is subject to change as the project nears construction. As such, this first revision to the original test plan highlights the new Microreactor Program-sponsored pathway for engaging with stakeholders via an Expression of Interest and Notice of Opportunity.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

SAS4A/SASSYS-1 Validation with EBR-II Tests Performed During the SHRT Testing Program (Rev. 2)

The EBR-II sodium-cooled fast reactor supported a wide range of fast reactor research activities. The final fifteen years of operations at EBR-II were used for experiments and tests to demonstrate the importance of passive safety in liquid metal reactors and the capability of such systems to provide strong passive responses during off-normal conditions. The Shutdown Heat Removal Test program at EBR-II provided test data supporting the validation of computer codes for design, licensing, and operation of LMRs, among other objectives. The program included nearly sixty tests, most of which were protected and unprotected loss of flow and loss of heat sink tests run at various initial powers and flow rates. This report documents the results of modeling and simulation of tests from the Shutdown Heat Removal Test program to support validation of Argonne’s SAS4A/SASSYS-1 fast reactor safety analysis code. SHRT-17 and SHRT-6 were protected loss of flow tests where a loss of electrical power to all sodium coolant pumps was simulated to demonstrate the effectiveness of natural circulation cooling characteristics. SHRT-45R, SHRT-39, and SHRT-43R were unprotected loss of flow tests where the control rod scram function of the plant protection system was disabled to demonstrate the effectiveness of EBR-II’s passive reactivity feedbacks. BOP-301 and BOP-302R were unprotected loss of heat sink tests that demonstrated how reactivity feedback effects driven by changes to the core inlet temperature can shut down the fission process. PICT-5/6 demonstrated how core power could be increased and decreased via changes in the intermediate sodium and feedwater flow rates. This validation activity leverages modeling and simulation efforts that originated under an International Atomic Energy Agency Coordinated Research Project for benchmark analysis of the EBR-II Shutdown Heat Removal Tests. The SAS4A/SASSYS-1 models originally developed for the CRP have been modernized to utilize new code features and modeling practices. Differences between the predicted and measured test data were explored and the causes of these differences were identified as being due to modeling approximations, uncertainty in transient component behavior, and instrumentation uncertainties. Overall, it was concluded that for all tests simulated there was good agreement between the flow, power, and temperature predictions and the measured data.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRS2017

Evolving Maturation of the Series-Bosch System

Human exploration missions to Mars and other destinations beyond low Earth orbit require highly robust, reliable, and maintainable life support systems that maximize recycling of water and oxygen. In order to meet this requirement, NASA has continued the development of a Series-Bosch System, a two stage reactor process that reduces carbon dioxide (CO2) with hydrogen (H2) to produce water and solid carbon. Theoretically, the Bosch process can recover 100% of the oxygen (O2) from CO2 in the form of water, making it an attractive option for long duration missions. The Series Bosch system includes a reverse water gas shift (RWGS) reactor, a carbon formation reactor (CFR), an H2 extraction membrane, and a CO2 extraction membrane. In 2016, the results of integrated testing of the Series Bosch system showed great promise and resulted in design modifications to the CFR to further improve performance. This year, integrated testing was conducted with the modified reactor to evaluate its performance and compare it with the performance of the previous configuration. Additionally, a CFR with the capability to load new catalyst and remove spent catalyst in-situ was built. Flow demonstrations were performed to evaluate both the catalyst loading and removal process and the hardware performance. The results of the integrated testing with the modified CFR as well as the flow demonstrations are discussed in this paper.

Stanley, Christine↗
NASA NTRS2017

Evolving Maturation of the Series-Bosch System

Human exploration missions to Mars and other destinations beyond low Earth orbit require highly robust, reliable, and maintainable life support systems that maximize recycling of water and oxygen. In order to meet this requirement, NASA has continued the development of a Series-Bosch System, a two stage reactor process that reduces carbon dioxide (CO2) with hydrogen (H2) to produce water and solid carbon. Theoretically, the Bosch process can recover 100% of the oxygen (O2) from CO2 in the form of water, making it an attractive option for long duration missions. The Series Bosch system includes a reverse water gas shift (RWGS) reactor, a carbon formation reactor (CFR), an H2 extraction membrane, and a CO2 extraction membrane. In 2016, the results of integrated testing of the Series Bosch system showed great promise and resulted in design modifications to the CFR to further improve performance. This year, integrated testing was conducted with the modified reactor to evaluate its performance and compare it with the performance of the previous configuration. Additionally, a CFR with the capability to load new catalyst and remove spent catalyst in-situ was built. Flow demonstrations were performed to evaluate both the catalyst loading and removal process and the hardware performance. The results of the integrated testing with the modified CFR as well as the flow demonstrations are discussed in this paper.

Abney, Morgan B.↗
DOE OSTI2022

Results of Initial Alloy 617 High Temperature Crack Growth Testing

Crack propagation data can provide valuable insights when performing a safety evaluation for a component. Alloy 617 is qualified in Section III, Division 5 of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC) for elevated-temperature nuclear service up to 950°C. Little is known, however, about subcritical crack-growth phenomena in Alloy 617. Previous crack-growth studies of Alloy 617 did not investigate temperatures across the range which is currently qualified in Section III, Division 5. High-temperature crack-growth testing in air and in reactor-grade helium can provide data for establishing the crack-growth correlations in support of an ASME BPVC Section XI high-temperature flaw evaluation Code Case. Idaho National Laboratory (INL) has previously performed crack-growth testing, but the equipment requires revitalization. This report provides the status of crack-growth testing in Alloy 617 at INL.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS2020

Results of the KRUSTY Warm Critical Experiments

The Kilowatt Reactor Using Stirling TechnologY (KRUSTY) was a prototypic nuclear-powered test of a 5-kW(thermal) Kilopower space reactor. This paper presents results from the KRUSTY warm critical experiments, which were completed prior to the final system test. The first set of criticals comprised cold or zero-power criticals; i.e., the core was not heated by fission power. These were followed by three warm criticals, where fission power heated the core to 200°C, 300°C, and 450°C, respectively. These criticals provided the data, confidence, and regulatory framework that were needed to proceed with the KRUSTY nuclear system test. The criticals also provided valuable data for the benchmarking of codes applicable to all nuclear systems. Finally, a comparison of KRUSTY results to pretest predictions is provided, and overall, the models matched the experimental results very closely.

David I. Poston↗
DOE OSTI2021

The high-power helicon program at DIII-D: gearing up for first experiments

Helicon current drive, also called fast wave current drive in the lower hybrid range of frequencies, has long been regarded as a promising current drive tool for reactor grade plasmas. A newly installed MW-level system at DIII-D will be the first test of this technology in reactor-relevant plasmas, in the sense that full single-pass absorption is expected. A 30-module traveling wave antenna has been installed and optimized in-vessel in early 2020. The linear electromagnetic characteristics of the unloaded module array have been extensively tested both on the bench and in the vessel at instrumentation power levels. Excellent performance has been achieved, ~2% reflected power and ~1.5% dissipated power per module in air, in a 10 MHz band around 476 MHz. Stripline feeds on both ends of the antenna allow either co or counter current drive. The installation of a 1.2 MW klystron and associated high-power electronics was completed in Fall 2020. Commissioning of the antenna is ongoing. An important goal of this experiment is to validate the helicon current drive physics basis using an extensive set of new and upgraded diagnostics.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗