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At least 415 records · Page 23

Antimatter Driven P-B11 Fusion Propulsion System

One of the major advantages of using P-B11 fusion fuel is that the reaction produces only charged particles in the form of three alpha particles and no neutrons. A fusion concept that lends itself to this fuel cycle is the Magnetically Insulated Inertial Confinement Fusion (MICF) reactor whose distinct advantage lies in the very strong magnetic field that is created when an incident particle (or laser) beam strikes the inner wall of the target pellet. This field serves to thermally insulate the hot plasma from the metal wall thereby allowing thc plasma to burn for a long time and produce a large energy magnification. If used as a propulsion device, we propose using antiprotons to drive the system which we show to be capable of producing very large specific impulse and thrust. By way of validating the confinement propenies of MICF we will address a proposed experiment in which pellets coated with P-B11 fuel at the appropriate ratio will be zapped by a beam of antiprotons that enter the target through a hole. Calculations showing the density and temperature of the generated plasma along with the strength of the magnetic field and other properties of the system will be presented and discussed.

Kammash, Terry↗

Nuclear Thermal Rocket/Vehicle Design Options for Future NASA Missions to the Moon and Mars

The nuclear thermal rocket (NTR) provides a unique propulsion capability to planners/designers of future human exploration missions to the Moon and Mars. In addition to its high specific impulse (approximately 850-1000 s) and engine thrust-to-weight ratio (approximately 3-10), the NTR can also be configured as a 'dual mode' system capable of generating electrical power for spacecraft environmental systems, communications, and enhanced stage operations (e.g., refrigeration for long-term liquid hydrogen storage). At present the Nuclear Propulsion Office (NPO) is examining a variety of mission applications for the NTR ranging from an expendable, single-burn, trans-lunar injection (TLI) stage for NASA's First Lunar Outpost (FLO) mission to all propulsive, multiburn, NTR-powered spacecraft supporting a 'split cargo-piloted sprint' Mars mission architecture. Each application results in a particular set of requirements in areas such as the number of engines and their respective thrust levels, restart capability, fuel operating temperature and lifetime, cryofluid storage, and stage size. Two solid core NTR concepts are examined -- one based on NERVA (Nuclear Engine for Rocket Vehicle Application) derivative reactor (NDR) technology, and a second concept which utilizes a ternary carbide 'twisted ribbon' fuel form developed by the Commonwealth of Independent States (CIS). The NDR and CIS concepts have an established technology database involving significant nuclear testing at or near representative operating conditions. Integrated systems and mission studies indicate that clusters of two to four 15 to 25 klbf NDR or CIS engines are sufficient for most of the lunar and Mars mission scenarios currently under consideration. This paper provides descriptions and performance characteristics for the NDR and CIS concepts, summarizes NASA's First Lunar Outpost and Mars mission scenarios, and describes characteristics for representative cargo and piloted vehicles compatible with a reference 240 t-class heavy lift launch vehicle (HLLV) and smaller 120 t HLLV option. Attractive performance characteristics and high-leverage technologies associated with both the engine and stage are identified, and supporting parametric sensitivity data is provided. The potential for commonality of engine and stage components to satisfy a broad range of lunar and Mars missions is also discussed.

Borowski, Stanley K.↗

Nuclear Thermal Propulsion Ground Test History

Nuclear Thermal Propulsion (NTP) was started in ~1955 under the Atomic Energy Commission as project Rover and was assigned to Los Alamos National Laboratory. The Nevada Test Site was selected in 1956 and facility construction began in 1957. The KIWI-A was tested on July 1, 1959 for 5 minutes at 70MW. KIWI-A1 was tested on July 8, 1960 for 6 minutes at 85MW. KIWI-A3 was tested on October 10, 1960 for 5 minutes at 100MW. The National Aeronautics and Space Administration (NASA) was formed in 1958. On August 31, 1960 the AEC and NASA established the Space Nuclear Propulsion Office and named Harold Finger as Director. Immediately following the formation of SNPO, contracts were awarded for the Reactor In Flight Test (RIFT), master plan for the Nuclear Rocket Engine Development Station (NRDS), and the Nuclear Engine for Rocket Vehicle Application (NERVA). From December 7, 1961 to November 30, 1962, the KIWI-B1A, KIWI-B1B, and KIWI-B4A were tested at test cell A. The last two engines were only tested for several seconds before noticeable failure of the fuel elements. Harold Finger called a stop to any further hot fire testing until the problem was well understood. The KIWI-B4A cold flow test showed the problem to be related to fluid dynamics of hydrogen interstitial flow causing fuel element vibrations. President Kennedy visited the NTS one week after the KIWI-B4A failure and got to see the engine starting to be disassembled in the maintenance facility. The KIWI-B4D and KIWI-B4E were modified to not have the vibration problems and were tested in test cell C. The NERVA NRX program started testing in early 1964 with NRX-A1 cold flow test series (unfueled graphite core), NRX-A2 and NRX-A3 power test series up to 1122 MW for 13 minutes. In March 1966, the NRX-EST (Engine System Test) was the first breadboard using flight functional relationship and total operating time of 116 minutes. The NRX-EST demonstrated the feasibility of a hot bleed cycle. The NRX-A5 had multiple start-ups in May-June 1966 with 30.75 minutes accumulative operating time at or above 1GW. The NRX-A6 was tested in December 1969 and ran for 62 minutes at 1100 MW. Each engine had post-test examination and found various structure anomalies which were identified for correction and the fuel element corrosion rate was reduced. The Phoebus series of research reactors began testing at test cell C, in June 1965 with Phoebus 1A. Phoebus 1A operated for 10.5 minutes at 1100 MW before unexpected loss of propellant and leading to an engine breakdown. Phoebus 1B ran for 30 minutes in February of 1967. Phoebus 2A was the highest steady state reactor built at 5GW. Phoebus 2A ran for 12 minutes at 4100 MW demonstrating sufficient power is available. The Peewee test bed reactor was tested November- December 1968 in test cell C for 40 minutes at 500MW with overall performance close to pre-run predictions. The XE' engine was the only engine tested with close to a flight configuration and fired downward into a diffuser at the Engine Test Stand (ETS) in 1969. The XE' was 1100 MW and had ~28 start-ups. The nuclear furnace NF-1 was operated at 44 MW with multiple test runs at 90 minutes in the summer of 1972. The NF-1 was the last NTP reactor tested. The Rover/NERVA program was cancelled in 1973. However, before cancellation, a lot of other engineering work was conducted by Aerojet on a 75, 000 lbf prototype flight engine and by Los Alamos on a ~16,000 lbf "Small Engine" nuclear rocket design. The ground test history of NTP at the NRDS also offers many lessons learned on how best to setup, operate, emergency shutdown, and post-test examine NTP engines. The reactor and engine maintenance and disassembly facilities were used for assembly and inspection of radioactive engines after testing. Most reactor/ engines were run at test cell A or test cell C with open air exhaust. The Rover/NERVA program became aware of a new environmental regulation that would restrict the amount of radioactive particulates allowed to be release in open air and successfully demonstrated a scrubber concept with the NF-1. The ETS stand was the only one with a high altitude test chamber used for XE'. The ETS and other test cells showed the effects the engine's radiation had on the facility materials and instrumentation as well as side effects the ground test facility has back on the engine operation. The breakdown of Phoebus 1A at test cell C showed how the site was cleaned up and back to operation for five more engines before the program was cancelled.

Gerrish, Harold P.↗

Experimental physics characteristics of a heavy-metal-reflected fast-spectrum critical assembly

A zero-power critical assembly was designed, constructed, and operated for the purpose of conducting a series of benchmark experiments dealing with the physics characteristics of a UN-fueled, Li-7 cooled, Mo-reflected, drum-controlled compact fast reactor for use with a space-power electric conversion system. The experimental program consisted basically of measuring the differential neutron spectra and the changes in critical mass that accompanied the stepwise addition of (Li-7)3N, Hf, Ta, and W to a basic core fueled with U metal in a pin-type Ta honeycomb structure. In addition, experimental results were obtained on power distributions, control characteristics, neutron lifetime, and reactivity worths of numerous absorber, structural, and scattering materials.

Heneveld, W. H.↗

An analytical study of nitrogen oxides and carbon monoxide emissions in hydrocarbon combustion with added nitrogen, preliminary results

The effect of combustor operating conditions on the conversion of fuel-bound nitrogen (FBN) to nitrogen oxides NO sub x was analytically determined. The effect of FBN and of operating conditions on carbon monoxide (CO) formation was also studied. For these computations, the combustor was assumed to be a two stage, adiabatic, perfectly-stirred reactor. Propane-air was used as the combustible mixture and fuel-bound nitrogen was simulated by adding nitrogen atoms to the mixture. The oxidation of propane and formation of NO sub x and CO were modeled by a fifty-seven reaction chemical mechanism. The results for NO sub x and CO formation are given as functions of primary and secondary stage equivalence ratios and residence times.

Bittker, D. A.↗

Initial Operation of the Nuclear Thermal Rocket Element Environmental Simulator

The Nuclear Thermal Rocket Element Environmental Simulator (NTREES) facility is designed to perform realistic non-nuclear testing of nuclear thermal rocket (NTR) fuel elements and fuel materials. Although the NTREES facility cannot mimic the neutron and gamma environment of an operating NTR, it can simulate the thermal hydraulic environment within an NTR fuel element to provide critical information on material performance and compatibility. The NTREES facility has recently been upgraded such that the power capabilities of the facility have been increased significantly. At its present 1.2 MW power level, more prototypical fuel element temperatures nay now be reached. The new 1.2 MW induction heater consists of three physical units consisting of a transformer, rectifier, and inverter. This multiunit arrangement facilitated increasing the flexibility of the induction heater by more easily allowing variable frequency operation. Frequency ranges between 20 and 60 kHz can accommodated in the new induction heater allowing more representative power distributions to be generated within the test elements. The water cooling system was also upgraded to so as to be capable of removing 100% of the heat generated during testing In this new higher power configuration, NTREES will be capable of testing fuel elements and fuel materials at near-prototypic power densities. As checkout testing progressed and as higher power levels were achieved, several design deficiencies were discovered and fixed. Most of these design deficiencies were related to stray RF energy causing various components to encounter unexpected heating. Copper shielding around these components largely eliminated these problems. Other problems encountered involved unexpected movement in the coil due to electromagnetic forces and electrical arcing between the coil and a dummy test article. The coil movement and arcing which were encountered during the checkout testing effectively destroyed the induction coil in use at the time and resulted in NTREES being out of commission for a couple of months while a new stronger coil was procured. The new coil includes several additional pieces of support structure to prevent coil movement in the future. In addition, new insulating test article support components have been fabricated to prevent unexpected arcing to the test articles. Additional activities are also now underway to address ways in which the radial temperature profiles across test articles may be controlled such that they are more prototypical of what they would encounter in an operating nuclear engine. The causes of the temperature distribution problem are twofold. First, the fuel element test article is isolated in NTREES as opposed to being in the midst of many other mostly identical fuel elements in a nuclear engine. As a result, the fuel element heat flux boundary conditions in NTREES are far from adiabatic as would normally be the case in a reactor. Second, induction heating skews the power distribution such that power is preferentially deposited near the outside of the fuel element. Nuclear heating, conversely, deposits its power much more uniformly throughout the fuel element. Current studies are now looking at various schemes to adjust the amount of thermal radiation emitted from the fuel element surface so as to essentially vary the thermal boundary conditions on the test article. It is hoped that by properly adjusting the thermal boundary conditions on the fuel element test article, it may be possible to substantially correct for the inappropriate radial power distributions resulting from the induction heating so as to yield a more nearly correct temperature distribution throughout the fuel element.

Emrich, William J., Jr.↗

Circulation system for flowing uranium hexafluoride cavity reactor experiments

Research related to determining the feasibility of producing continuous power from fissile fuel in the gaseous state is presented. The development of three laboratory-scale flow systems for handling gaseous UF6 at temperatures up to 500 K, pressure up to approximately 40 atm, and continuous flow rates up to approximately 50g/s is presented. A UF6 handling system fabricated for static critical tests currently being conducted is described. The system was designed to supply UF6 to a double-walled aluminum core canister assembly at temperatures between 300 K and 400 K and pressure up to 4 atm. A second UF6 handling system designed to provide a circulating flow of up to 50g/s of gaseous UF6 in a closed-loop through a double-walled aluminum core canister with controlled temperature and pressure is described. Data from flow tests using UF6 and UF6/He mixtures with this system at flow rates up to approximately 12g/s and pressure up to 4 atm are presented. A third UF6 handling system fabricated to provide a continuous flow of UF6 at flow rates up to 5g/s and at pressures up to 40 atm for use in rf-heated, uranium plasma confinement experiments is described.

Jaminet, J. F.↗

Results of Uranium Dioxide-Tungsten Irradiation Test and Post-Test Examination

A uranium dioxide (UO2) fueled capsule was fabricated and irradiated in the NASA Plum Brook Reactor Facility. The capsule consisted of two bulk UO2 specimens clad with chemically vapor deposited tungsten (CVD W) 0.762 and 0.1016 cm (0.030-and 0.040-in.) thick, respectively. The second specimen with 0.1016-cm (0.040-in.) thick cladding was irradiated at temperature for 2607 hours, corresponding to an average burnup of 1.516 x 10 to the 20th power fissions/cu cm. Postirradiation examination showed distortion in the bottom end cap, failure of the weld joint, and fracture of the central vent tube. Diametral growth was 1.3 percent. No evidence of gross interaction between CVD tungsten or arc-cast tungsten cladding and the UO2 fuel was observed. Some of the fission gases passed from the fuel cavity to the gas surrounding the fuel specimen via the vent tube and possibly the end-cap weld failure. Whether the UO2 loss rates through the vent tube were within acceptable limits could not be determined in view of the end-cap weld failure.

Collins, J. F.↗

Applying design principles to fusion reactor configurations for propulsion in space

The application of fusion power to space propulsion requires rethinking the engineering-design solution to controlled-fusion energy. Whereas the unit cost of electricity (COE) drives the engineering-design solution for utility-based fusion reactor configurations; initial mass to low earth orbit (IMLEO), specific jet power (kW(thrust)/kg(engine)), and reusability drive the engineering-design solution for successful application of fusion power to space propulsion. We applied three design principles (DP's) to adapt and optimize three candidate-terrestrial-fusion-reactor configurations for propulsion in space. The three design principles are: provide maximum direct access to space for waste radiation, operate components as passive radiators to minimize cooling-system mass, and optimize the plasma fuel, fuel mix, and temperature for best specific jet power. The three candidate terrestrial fusion reactor configurations are: the thermal barrier tandem mirror (TBTM), field reversed mirror (FRM), and levitated dipole field (LDF). The resulting three candidate space fusion propulsion systems have their IMLEO minimized and their specific jet power and reusability maximized. We performed a preliminary rating of these configurations and concluded that the leading engineering-design solution to space fusion propulsion is a modified TBTM that we call the Mirror Fusion Propulsion System (MFPS).

Carpenter, Scott A.↗

An unmanned mission to Mars with sample collection and in-situ resource utilization

The design for the Mars Analysis and Return Vehicle with In-Situ Resource Utilization (MARVIN) project is outlined. The MARVIN mission is designed to collect samples of the Martian environment; to produce fuel from local Martian resources; and to use the fuel produced to return the samples to earth. It uses only existing technologies. Exploratory Technologies' mission-design efforts have focused on methods of orbit determination, sample collection, fuel production, power, communications, control, and structural design. Lambert Targeting provided Delta-V's, launch dates, and travel times. The landing site is the Tharsis Plateau, to the southeast of Olympus Mons, chosen for its substantial scientific value. Samples of soil, dust, and atmosphere are collected with lander-based collection devices: the soil sample, with a robotic arm similar to those used in the Viking missions; the atmospheric sample, from a bleed line to the compressor in the fuel-production facility; a dust sample, from the dust-collection container in the fuel-production facility; and a redundant dust sample, with a with a passive filter system, which relies upon neither a power source nor other collection methods. The sample-return capsule (SRC) houses these samples, which are triply contained to prevent contamination. Proven technology can be used to produce methane and oxygen for fuel with relative ease at the landing site: the Sabatier reactor produces methane and water by combining carbon dioxide and hydrogen (brought from earth); the Reverse Water-Gas Shift unit combines carbon dioxide and hydrogen to form carbon monoxide and water; a water-electrolysis unit splits the water into hydrogen and oxygen. The Mars-lander vehicle (MLV) transports the equipment from earth to Mars. The Mars-ascent vehicle (MAV) contains the SRC and the engine, which is the same for both the MLV and the MAV. All equipment that is unnecessary for the Mars-Earth trajectory remains on Mars. This report presents detailed sizing information, for which a spreadsheet has been developed. The trends suggest possibilities for expansion, and suggestions for future work in these areas are offered.

Source record↗

Nuclear Thermal Rocket Element Environmental Simulator (NTREES) Upgrade Activities

A key technology element in Nuclear Thermal Propulsion is the development of fuel materials and components which can withstand extremely high temperatures while being exposed to flowing hydrogen. NTREES provides a cost effective method for rapidly screening of candidate fuel components with regard to their viability for use in NTR systems. The NTREES is designed to mimic the conditions (minus the radiation) to which nuclear rocket fuel elements and other components would be subjected to during reactor operation. The NTREES consists of a water cooled ASME code stamped pressure vessel and its associated control hardware and instrumentation coupled with inductive heaters to simulate the heat provided by the fission process. The NTREES has been designed to safely allow hydrogen gas to be injected into internal flow passages of an inductively heated test article mounted in the chamber.

Emrich, William↗

Cryogenic Fluid Management Technology and Nuclear Thermal Propulsion

Cryogenic fluid management (CFM) is critical to the success of future nuclear thermal propulsion powered vehicles. While this is an issue for any propulsion system utilizing cryogenic propellants, this is made more challenging by the radiation flux produced by the reactor in a nuclear thermal rocket (NTR). Managing the cryogenic fuel to prevent propellant loss to boil off and leakage is needed to limit the required quantity of propellant to a reasonable level. Analysis shows deposition of energy into liquid hydrogen fuel tanks in the vicinity of the nuclear thermal engine. This is on top of ambient environment sources of heat. Investments in cryogenic/thermal management systems (some of which are ongoing at various organizations) are needed in parallel to nuclear thermal engine development in order to one day see the successful operation of an entire stage. High durability, low thermal conductivity insulation is one developmental need. Light weight cryocoolers capable of removing heat from large fluid volumes at temperatures as low as approx. 20 K are needed to remove heat leak from the propellant of an NTR. Valve leakage is an additional CFM issue of great importance. Leakage rates of state of the art, launch vehicle size valves (which is approximately the size valves needed for a Mars transfer vehicle) are quite high and would result in large quantities of lost propellant over a long duration mission. Additionally, the liquid acquisition system inside the propellant tank must deliver properly conditioned propellant to the feed line for successful engine operation and avoid intake of warm or gaseous propellant. Analysis of the thermal environment and the CFM technology development are discussed in the accompanying presentation.

Taylor, Brian D.↗

Energetic closed-cycle gas core reactors for orbit raising

Closed-cycle gas core reactor power plants can be of two types. In the 'mixed flow' type, the gaseous nuclear fuel is intimately mixed with the working gas in the cavity. In the 'light bulb' type the fissioning plasma is enclosed in a transparent tube, and energy transfer to the separate working gas occurs by thermal radiation. The potentials of high temperature gas core reactors in terrestrial electric power generator applications have been considered, and a number of civilian power-beaming applications for gaseous fuel nuclear-MHD power plants in space have been suggested. Major conclusions of investigations related to the design of space power systems are discussed. Attention is given to options for conversion cycles, the power system specific mass, and research and technology issues.

Rosa, R. J.↗

Gas core reactors for actinide transmutation and breeder applications

This work consists of design power plant studies for four types of reactor systems: uranium plasma core breeder, uranium plasma core actinide transmuter, UF6 breeder and UF6 actinide transmuter. The plasma core systems can be coupled to MHD generators to obtain high efficiency electrical power generation. A 1074 MWt UF6 breeder reactor was designed with a breeding ratio of 1.002 to guard against diversion of fuel. Using molten salt technology and a superheated steam cycle, an efficiency of 39.2% was obtained for the plant and the U233 inventory in the core and heat exchangers was limited to 105 Kg. It was found that the UF6 reactor can produce high fluxes (10 to the 14th power n/sq cm-sec) necessary for efficient burnup of actinide. However, the buildup of fissile isotopes posed severe heat transfer problems. Therefore, the flux in the actinide region must be decreased with time. Consequently, only beginning-of-life conditions were considered for the power plant design. A 577 MWt UF6 actinide transmutation reactor power plant was designed to operate with 39.3% efficiency and 102 Kg of U233 in the core and heat exchanger for beginning-of-life conditions.

Clement, J. D.↗

Gas core reactor concepts and technology - Issues and baseline strategy

Results of a research program including phenomenological studies, conceptual design, and systems analysis of a series of gaseous/vapor fissile fuel driven engines for space power platforms and for thermal and electric propulsion are reviewed. It is noted that gas and vapor phase reactors provide the path for minimum mass in orbit and trip times, with a specific impulse from 1020 sec at the lowest technololgical risk to 5200 sec at the highest technological risk. The discussion covers various configurations of gas core reactors and critical technologies and the nuclear vapor thermal rocket engine.

Diaz, Nils J.↗

Heat Pipe Powered Stirling Conversion for the Demonstration Using Flattop Fission (DUFF) Test

Design concepts for small Fission Power Systems (FPS) have shown that heat pipe cooled reactors provide a passive, redundant, and lower mass option to transfer heat from the fuel to the power conversion system, as opposed to pumped loop designs typically associated with larger FPS. Although many systems have been conceptually designed and a few making it to electrically heated testing, none have been coupled to a real nuclear reactor. A demonstration test named DUFF Demonstration Using Flattop Fission, was planned by the Los Alamos National Lab (LANL) to use an existing criticality experiment named Flattop to provide the nuclear heat source. A team from the NASA Glenn Research Center designed, built, and tested a heat pipe and power conversion system to couple to Flattop with the end goal of making electrical power. This paper will focus on the design and testing performed in preparation for the DUFF test.

Fission↗

Compact reactor for onboard hydrogen generation

Hydrogen, chemically stored as methanol, is promising internal-combustion fuel. Methanol is readily obtainable from natural products such as wood, compost, or various organic wastes. Steam reformation of methanol as source for hydrogen is relatively simple operation.

Brabbs, T. A.↗