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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 847 records · Page 47
All sourcesNASA NTRSDOE OSTI
NASA NTRS1992

Nuclear thermal propulsion transportation systems for lunar/Mars exploration

Nuclear thermal propulsion technology development is underway at NASA and DoE for Space Exploration Initiative (SEI) missions to Mars, with initial near-earth flights to validate flight readiness. Several reactor concepts are being considered for these missions, and important selection criteria will be evaluated before final selection of a system. These criteria include: safety and reliability, technical risk, cost, and performance, in that order. Of the concepts evaluated to date, the Nuclear Engine for Rocket Vehicle Applications (NERVA) derivative (NDR) is the only concept that has demonstrated full power, life, and performance in actual reactor tests. Other concepts will require significant design work and must demonstrate proof-of-concept. Technical risk, and hence, development cost should therefore be lowest for the concept, and the NDR concept is currently being considered for the initial SEI missions. As lighter weight, higher performance systems are developed and validated, including appropriate safety and astronaut-rating requirements, they will be considered to support future SEI application. A space transportation system using a modular nuclear thermal rocket (NTR) system for lunar and Mars missions is expected to result in significant life cycle cost savings. Finally, several key issues remain for NTR's, including public acceptance and operational issues. Nonetheless, NTR's are believed to be the 'next generation' of space propulsion systems - the key to space exploration.

Clark, John S.↗
NASA NTRS1991

Review of Rover fuel element protective coating development at Los Alamos

The Los Alamos Scientific Laboratory (LASL) entered the nuclear propulsion field in 1955 and began work on all aspects of a nuclear propulsion program with a target exhaust temperature of about 2750 K. A very extensive chemical vapor deposition coating technology for preventing catastrophic corrosion of reactor core components by the high temperature, high pressure hydrogen propellant gas was developed. Over the 17-year term of the program, more than 50,000 fuel elements were coated and evaluated. Advances in performance were achieved only through closely coupled interaction between the developing fuel element fabrication and protective coating technologies. The endurance of fuel elements in high temperature, high pressure hydrogen environment increased from several minutes at 2000 K exit gas temperature to 2 hours at 2440 K exit gas temperature in a reactor test and 10 hours at 2350 K exit gas temperature in a hot gas test. The purpose of this paper is to highlight the rationale for selection of coating materials used (NbC and ZrC), identify critical fuel element-coat interactions that had to be modified to increase system performance, and review the evolution of protective coating technology.

Wallace, Terry C.↗
NASA NTRS1994

A review of nuclear thermal propulsion carbide fuel corrosion and key issues

Corrosion (mass loss) of carbide nuclear fuels due to their exposure to hot hydrogen in nuclear thermal propulsion engine systems greatly impacts the performance, thrust-to-weight and life of such systems. This report provides an overview of key issues and processes associated with the corrosion of carbide materials. Additionally, past pertinent development reactor test observations, as well as related experimental work and analysis modeling efforts are reviewed. At the conclusion, recommendations are presented, which provide the foundation for future corrosion modeling and verification efforts.

Pelaccio, Dennis G.↗
NASA NTRS2006

Bench-Scale Monolith Autothermal Reformer Catalyst Screening Evaluations in a Micro-Reactor With Jet-A Fuel

Solid oxide fuel cell systems used in the aerospace or commercial aviation environment require a compact, light-weight and highly durable catalytic fuel processor. The fuel processing method considered here is an autothermal reforming (ATR) step. The ATR converts Jet-A fuel by a reaction with steam and air forming hydrogen (H2) and carbon monoxide (CO) to be used for production of electrical power in the fuel cell. This paper addresses the first phase of an experimental catalyst screening study, looking at the relative effectiveness of several monolith catalyst types when operating with untreated Jet-A fuel. Six monolith catalyst materials were selected for preliminary evaluation and experimental bench-scale screening in a small 0.05 kWe micro-reactor test apparatus. These tests were conducted to assess relative catalyst performance under atmospheric pressure ATR conditions and processing Jet-A fuel at a steam-to-carbon ratio of 3.5, a value higher than anticipated to be run in an optimized system. The average reformer efficiencies for the six catalysts tested ranged from 75 to 83 percent at a constant gas-hourly space velocity of 12,000 hr 1. The corresponding hydrocarbon conversion efficiency varied from 86 to 95 percent during experiments run at reaction temperatures between 750 to 830 C. Based on the results of the short-duration 100 hr tests reported herein, two of the highest performing catalysts were selected for further evaluation in a follow-on 1000 hr life durability study in Phase II.

Tomsik, Thomas M.↗
NASA NTRS2006

Science in Flux: NASA's Nuclear Program at Plum Brook Station 1955-2005

Science in Flux traces the history of one of the most powerful nuclear test reactors in the United States and the only nuclear facility ever built by NASA. In the late 1950's NASA constructed Plum Brook Station on a vast tract of undeveloped land near Sandusky, Ohio. Once fully operational in 1963, it supported basic research for NASA's nuclear rocket program (NERVA). Plum Brook represents a significant, if largely forgotten, story of nuclear research, political change, and the professional culture of the scientists and engineers who devoted their lives to construct and operate the facility. In 1973, after only a decade of research, the government shut Plum Brook down before many of its experiments could be completed. Even the valiant attempt to redefine the reactor as an environmental analysis tool failed, and the facility went silent. The reactors lay in costly, but quiet standby for nearly a quarter-century before the Nuclear Regulatory Commission decided to decommission the reactors and clean up the site. The history of Plum Brook reveals the perils and potentials of that nuclear technology. As NASA, Congress, and space enthusiasts all begin looking once again at the nuclear option for sending humans to Mars, the echoes of Plum Brook's past will resonate with current policy and space initiatives.

Bowles, Mark D.↗
NASA NTRS2013

A Nuclear Cryogenic Propulsion Stage for Near-Term Space Missions

Development efforts in the United States have demonstrated the viability and performance potential of NTP systems. For example, Project Rover (1955 - 1973) completed 22 high power rocket reactor tests. Peak performances included operating at an average hydrogen exhaust temperature of 2550 K and a peak fuel power density of 5200 MW/m3 (Pewee test), operating at a thrust of 930 kN (Phoebus-2A test), and operating for 62.7 minutes on a single burn (NRXA6 test). Results from Project Rover indicated that an NTP system with a high thrust-toweight ratio and a specific impulse greater than 900 s would be feasible. Excellent results have also been obtained by Russia. Ternary carbide fuels developed in Russia may have the potential for providing even higher specific impulses.

Houts, Michael G.↗
NASA NTRS2013

A Nuclear Cryogenic Propulsion Stage for Near-Term Space Missions

Development efforts in the United States have demonstrated the viability and performance potential of NTP systems. For example, Project Rover (1955 - 1973) completed 22 high power rocket reactor tests. Peak performances included operating at an average hydrogen exhaust temperature of 2550 K and a peak fuel power density of 5200 MW/m3 (Pewee test), operating at a thrust of 930 kN (Phoebus-2A test), and operating for 62.7 minutes on a single burn (NRXA6 test).1 Results from Project Rover indicated that an NTP system with a high thrust-toweight ratio and a specific impulse greater than 900 s would be feasible. Binary and ternary carbide fuels may have the potential for providing even higher specific impulses.

Houts, Michael G.↗
NASA NTRS2014

The Nuclear Cryogenic Propulsion Stage

Nuclear Thermal Propulsion (NTP) development efforts in the United States have demonstrated the technical viability and performance potential of NTP systems. For example, Project Rover (1955 - 1973) completed 22 high power rocket reactor tests. Peak performances included operating at an average hydrogen exhaust temperature of 2550 K and a peak fuel power density of 5200 MW/m3 (Pewee test), operating at a thrust of 930 kN (Phoebus-2A test), and operating for 62.7 minutes in a single burn (NRX-A6 test). Results from Project Rover indicated that an NTP system with a high thrust-to-weight ratio and a specific impulse greater than 900 s would be feasible. Excellent results were also obtained by the former Soviet Union. Although historical programs had promising results, many factors would affect the development of a 21st century nuclear thermal rocket (NTR). Test facilities built in the US during Project Rover no longer exist. However, advances in analytical techniques, the ability to utilize or adapt existing facilities and infrastructure, and the ability to develop a limited number of new test facilities may enable affordable development, qualification, and utilization of a Nuclear Cryogenic Propulsion Stage (NCPS). Bead-loaded graphite fuel was utilized throughout the Rover/NERVA program, and coated graphite composite fuel (tested in the Nuclear Furnace) and cermet fuel both show potential for even higher performance than that demonstrated in the Rover/NERVA engine tests.. NASA's NCPS project was initiated in October, 2011, with the goal of assessing the affordability and viability of an NCPS. FY 2014 activities are focused on fabrication and test (non-nuclear) of both coated graphite composite fuel elements and cermet fuel elements. Additional activities include developing a pre-conceptual design of the NCPS stage and evaluating affordable strategies for NCPS development, qualification, and utilization. NCPS stage designs are focused on supporting human Mars missions. The NCPS is being designed to readily integrate with the Space Launch System (SLS). A wide range of strategies for enabling affordable NCPS development, qualification, and utilization should be considered. These include multiple test and demonstration strategies (both ground and in-space), multiple potential test sites, and multiple engine designs. Two potential NCPS fuels are currently under consideration - coated graphite composite fuel and tungsten cermet fuel. During 2014 a representative, partial length (approximately 16") coated graphite composite fuel element with prototypic depleted uranium loading is being fabricated at Oak Ridge National Laboratory (ORNL). In addition, a representative, partial length (approximately 16") cermet fuel element with prototypic depleted uranium loading is being fabricated at Marshall Space Flight Center (MSFC). During the development process small samples (approximately 3" length) will be tested in the Compact Fuel Element Environmental Tester (CFEET) at high temperature (approximately 2800 K) in a hydrogen environment to help ensure that basic fuel design and manufacturing process are adequate and have been performed correctly. Once designs and processes have been developed, longer fuel element segments will be fabricated and tested in the Nuclear Thermal Rocket Element Environmental Simulator (NTREE) at high temperature (approximately 2800 K) and in flowing hydrogen.

Houts, Michael G.↗
NASA NTRS2015

Increased Oxygen Recovery from Sabatier Systems Using Plasma Pyrolysis Technology and Metal Hydride Separation

State-of-the-art life support carbon dioxide (CO2) reduction technology is based on the Sabatier reaction where less than 50% of the oxygen required for the crew is recovered from metabolic CO2. The reaction produces water as the primary product and methane as a byproduct. Oxygen recovery is constrained by the limited availability of reactant hydrogen. This is further exacerbated when Sabatier methane (CH4) is vented as a waste product resulting in a continuous loss of reactant hydrogen. Post-processing methane with the Plasma Pyrolysis Assembly (PPA) to recover hydrogen has the potential to dramatically increase oxygen recovery and thus drastically reduce the logistical challenges associated with oxygen resupply. The PPA decomposes methane into predominantly hydrogen and acetylene. Due to the highly unstable nature of acetylene, a separation system is necessary to purify hydrogen before it is recycled back to the Sabatier reactor. Testing and evaluation of a full-scale Third Generation PPA is reported and investigations into metal hydride hydrogen separation technology is discussed.

Zachary W Greenwood↗
NASA NTRS2015

Nuclear Thermal Propulsion

Development efforts in the United States for nuclear thermal propulsion (NTP) systems began with Project Rover (1955-1973) which completed 22 high-power rocket reactor tests. Results indicated that an NTP system with a high thrust-to-weight ratio and a specific impulse greater than 900 s would be feasible. John F. Kennedy, in his historic special address to Congress on the importance of Space on May 25, 1961, said, "First, I believe that this nation should commit itself to achieving the goal, before this decade is out, of landing a man on the Moon and returning him safely to the Earth..." This was accomplished. He also said, "Secondly ... accelerate development of the Rover nuclear rocket. This gives promise of someday providing a means for even more exciting and ambitious exploration of space... to the very end of the solar system itself." The current NTP project focuses on demonstrating the affordability and viability of a fully integrated NTP system with emphasis on fuel fabrication and testing and an affordable development and qualification strategy. The goal is to enable NTP to be considered a mainstream option for supporting human Mars and other missions beyond Earth orbit.

Mitchell, Sonny↗
NASA NTRS2017

Methane Post-Processing and Hydrogen Separation for Spacecraft Oxygen Loop Closure

State-of-the-art life support oxygen recovery technology on the International Space Station is based on the Sabatier reaction where only about half of the oxygen required for the crew is recovered from metabolic carbon dioxide (CO2). The Sabatier reaction produces water as the primary product and methane as a byproduct. Oxygen recovery is constrained by both the limited availability of reactant hydrogen from water electrolysis and Sabatier methane (CH4) being vented as a waste product resulting in a continuous loss of reactant hydrogen. Post-processing methane with the Plasma Pyrolysis Assembly (PPA) to recover this hydrogen has the potential to substantially increase oxygen recovery and thus dramatically reduce the logistical challenges associated with oxygen resupply. The PPA decomposes methane into predominantly hydrogen and acetylene. A purification system is necessary to purify hydrogen before it is recycled back to the Sabatier reactor. Testing and evaluation of acetylene removal systems and PPA system architectures are presented and discussed.

Greenwood, Zachary W.↗
NASA NTRSDate not supplied

Recent Discoveries in Simulant Behavior and Regolith Handling

The past few years have seen an exciting increase in visibility for the use of in-situ resources to achieve sustainability at the Moon and Mars. This visibility and inclusion in the mission architectures has led to an explosion in the research and development of mining and processing of regolith at the Moon. However, there are still many fundamental issues regarding digging, transporting, and processing of lunar and Mars regolith that may slow the progress by multiple teams who need to understand these issues before proceeding with hardware development. The objective of this project is to utilize the same team of soil and thermal experts and unique facilities to allow for efficient and fast generation of fundamental data and dissemination of critical results to the larger community. The capability gaps being addressed include how to dig up hard, frozen resources, transport to the processing site, lifting/depositing into a hopper or reactor, and extracting the water, all without losing the primary product to the environment. Soil Seals: Much attention has been focused on the challenges with regolith inlet/outlet valves that must seal in process gases after being repeatedly ex-posed to abrasive regolith and dust particles. One possible solution that has been proposed is to design the inlet hopper or duct such that the bulk of regolith pro-vides sufficient sealing of the reactor. Tests are planned to parametrically measure the maximum del-ta-pressure that can be maintained across a column of soil before ‘blow-out’ occurs. Variables to be examined include column height and width, simulant type, and compaction.

Simulant↗
NASA NTRSDate not supplied

The Use of Immersion Rigs for High Temperature Hydrogen Exposure Testing within the Nuclear Thermal Rocket Element Environmental Simulator (NTREES): Thermal Soak Rig (TSR)

The Nuclear Thermal Rocket Element Environmental Simulator (NTREES) facility was purpose constructed to perform non-nuclear evaluations of nuclear thermal propulsion (NTP) system fuel materials and structures within prototypic thermochemical environments. This system has been utilized steadily in its ability to subject test specimens to thermochemical and thermohydraulic environments simulating that of an operating nuclear rocket engine. Fission heat is simulated by induction power and experiments are conducted within a ~1000 psi pressure vessel. Hydrogen is conventionally passed through the heated fuel surrogate test specimen while pressure, temperature, and gas species data are collected at various points along the experiment. In order to test fuel and material coupon samples, a class of test apparatus named “immersion rigs” are being developed and employed to more rapidly test these smaller and more technically challenging test specimen. One example of a promising potential fuel structure, Tristructural-isotropic (TRISO) particles, presents unique challenges for testing of this type. TRISO fuel micro-particles are spheroids typically on the order of 500 – 1000 μm in diameter, and exposing a batch sample to hot hydrogen requires purpose-built special test equipment. Thusly, an immersion rig was developed and successfully demonstrated to expose ~1 g of micro-particles to hydrogen gas at temperatures and pressures relevant to NTP systems for the purpose of fuel evaluation. The rig, comprised primarily of graphite and pure tungsten, houses in its core a batch of micro-particles between pucks of porous silicon carbide (SiC). This approach permits gas flow while simultaneously retaining the particles in place. Herein is a discussion of the design, analysis, fabrication, and testing of the NTREES Thermal Soak Rig (TSR).

Space Nuclear Propulsion↗
DOE OSTI2020

As-Run Physics Analysis for the EPRI-1 Experiment

The purpose of this ECAR is to document the as-run physics analysis for the Electric Power Research Institute (EPRI)-1 experiment located in the Center Flux Trap (CFT). EPRI-1 was irradiated during cycle 157C. The experiment irradiation ended on February 15, 2015. The as-run heat rates, DPA, and flux/fluence results were calculated using the Monte Carlo N-Particle (MCNP) Advanced Test Reactor (ATR) full core model. The heat rates, DPA, and flux are calculated based on an average center lobe power of 21.1 MW. The fluence and DPA is based on the operating time of 5.4 days. ORIGEN2 Version 2.2 was used to calculate the isotopic concentration in grams for the EPRI-1 experiment. The source term analysis was performed as requested by the project staff. The results of this analysis may be used to support Post-Irradiation Examination (PIE)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Actions to Evaluate Tritium in ATR Complex Perched Water Well 12

On April 23, 2019, elevated tritium concentrations in monitoring well PW-12 were reported to the Advanced Test Reactor (ATR) Complex environmental manager by the Fluor Idaho Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) lead. The monitoring well is one of a series of wells located in and around the ATR Complex that are sampled on an annual basis as part of the CERCLA monitoring program. The Fluor Idaho CERCLA lead indicated the elevated concentrations had been trending upward from Fiscal Year (FY) 2017 through FY 2019. The tritium levels went from 1,880 pCi/L in October 2016 to 7,890 pCi/L in October 2017 and increased to 13,000 pCi/L in October 2018. In order to identify a baseline, the safe drinking water level (e.g. 20,000 pCi/L) can be used as a reference point for comparison only, as PW-12 is a perched water well and does not penetrate the aquifer. Although well below safe drinking water levels, an investigation was initiated to identify the source of tritium.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2020

JMOCUP Physics Depletion Calculations for the As-Run AGR-5/6/7 TRISO Particle Experiment in ATR Northeast Flux Trap

This ECAR documents the as-run Jim Sterbentz’s MCNP ORIGEN Coupled Utility Program (JMOCUP) physics depletion calculation and the calculated results for the AGR-5/6/7 irradiation experiment in the Advanced Test Reactor (ATR). AGR 5/6/7 was irradiated for nine power cycles in the northeast flux trap. The depletion calculations were performed to provide input data for a variety of other engineering analyses supporting the AGR-5/6/7 experiment along with post-irradiation characterization of the tri-structural isotropic (TRISO) particle fuel compacts. Detailed full-core MCNP models and Oak Ridge Isotope Generation (ORIGEN2) radionuclide generation models were specifically developed as part of the JMOCUP Monte Carlo depletion calculations. The MCNP ORIGEN2 computer codes were coupled using the well-established JMOCUP utility modules to couple the two codes and run the depletion calculations. The physics calculations were performed in support of the Advanced Gas Reactor (AGR) program.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

As-Run Physics Analysis for the AGC-4 Experiment Irradiated in the ATR

This Engineering Calculations Analysis Report (ECAR) documents the results of the Advanced Test Reactor (ATR) detailed physics analyses performed to calculate the displacements per atom (DPA) and the fast neutron fluence (E > 0.1 MeV) of the Advanced Graphite Creep (AGC) experiment, AGC-4, irradiated in the ATR East Flux Trap (EFT) (see Figure 1) during ATR Cycle 157D, 158A, 162A, 162B, 164A, 164B, 166A, and Cycle 166B. This ECAR also reports the neutron and photon heat rates for the materials of the AGC-4 experiment for ATR Cycle 158A (timestep 19), which provides to the maximum heating. The results for these evaluations and analysis are reported herein. The AGC-4 as-run specimen neutron fast fluence (E > 0.1 MeV), DPA, and material heat rate calculations were performed using a general-purpose Monte Carlo N-Particle (MCNP) code. All calculated results are tabulated herein.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

As-Run Thermal Analysis for the AGC-4 Experiment Irradiated in the ATR

The Advanced Graphite Capsule (AGC) irradiation experiment will provide irradiation creep rate data for the new graphite proposed for the Next Generation Nuclear Plant (NGNP) program. The fourth experiment in the series (AGC-4) was designed to irradiate various types of graphite specimens at a temperature of 900 ºC and targeted displacements per atom (DPA) of 6. This experiment has been irradiated in the east flux trap of the Advanced Test Reactor (ATR) during the cycles of 157D, 158A, 162A, 162B, 164A, 164B, 166A, and 166B. Temperatures were monitored using twelve thermocouples (TC) located at various elevations in the reactor core, and variable helium-argon gas mixtures were used for gas gap temperature control of the specimens. The purpose of this Engineering Calculation and Analysis Report (ECAR) is to calculate the specimen temperature after the model is calibrated by the measured TC data with the as-run heating rates of the components, DPA of the graphite, and the gas mixture compositions during the experiment. As-run specimen mean temperature and the tolerance will be obtained.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗