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Results for “reactor pressure vessel”

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 19 records
All sourcesNASA NTRSDOE OSTI
NASA NTRS1973

Reactor moderator, pressure vessel, and heat rejection system of an open-cycle gas core nuclear rocket concept

A preliminary design study of a conceptual 6000-megawatt open-cycle gas-core nuclear rocket engine system was made. The engine has a thrust of 196,600 newtons (44,200 lb) and a specific impulse of 4400 seconds. The nuclear fuel is uranium-235 and the propellant is hydrogen. Critical fuel mass was calculated for several reactor configurations. Major components of the reactor (reflector, pressure vessel, and waste heat rejection system) were considered conceptually and were sized.

Taylor, M. F.↗
NASA NTRS1989

Energy efficient continuous flow ash lockhopper

The invention relates to an energy efficient continuous flow ash lockhopper, or other lockhopper for reactor product or byproduct. The invention includes an ash hopper at the outlet of a high temperature, high pressure reactor vessel containing heated high pressure gas, a fluidics control chamber having an input port connected to the ash hopper's output port and an output port connected to the input port of a pressure letdown means, and a control fluid supply for regulating the pressure in the control chamber to be equal to or greater than the internal gas pressure of the reactor vessel, whereby the reactor gas is contained while ash is permitted to continuously flow from the ash hopper's output port, impelled by gravity. The main novelty resides in the use of a control chamber to so control pressure under the lockhopper that gases will not exit from the reactor vessel, and to also regulate the ash flow rate. There is also novelty in the design of the ash lockhopper shown in two figures. The novelty there is the use of annular passages of progressively greater diameter, and rotating the center parts on a shaft, with the center part of each slightly offset from adjacent ones to better assure ash flow through the opening.

Collins, Earl R., Jr.↗
NASA NTRS2017

Non-Intrusive Velocity Measurements with MTV During DCC Event in the HTTF

Velocity profiles are measured using molecular tagging velocimetry (MTV) in the high temperature test facility (HTTF) at Oregon State University during a depressurized conduction cooldown (DCC) event. The HTTF is a quarter scale electrically heated nuclear reactor simulator designed to replicate various accident scenarios. During a DCC, a double ended guillotine break results in the reactor pressure vessel (RPV) depressurizing into the reactor cavity and ultimately leading to air ingress in the reactor core (lock-exchange and gas diffusion). It is critical to understand the resulting buoyancy-driven flow to characterize the reactor self-cooling capacity through natural circulation. During tests conducted at ambient pressure and temperature, the RPV containing helium is opened (via the hot and cold legs) to a large vessel filled with nitrogen to simulate the atmosphere. The velocity profile on the hot leg pipe centerline is recorded at 10 Hz with MTV based on NO tracers. The precision of the velocimetry was measured to be 0.02 m/s in quiescent flow prior to the tests. A helium flow from the RPV is initially observed in the top quarter of the pipe. During the first 20 seconds of the event, helium flows out of the RPV with a maximum velocity below 2 m/s. The velocity profile transitions from parabolic to linear in character and decays slowly over the rest of the recording; peak velocities of 0.2 m/s are observed after 30 min. A counter-flow of nitrogen is also observed intermittently, which occurs at lower velocities (>0.1 m/s).

Andre, M. A.↗
NASA NTRS1972

Discrete ordinates-Monte Carlo coupling: A comparison of techniques in NERVA radiation analysis

In the radiation analysis of the NERVA nuclear rocket system, two-dimensional discrete ordinates calculations are sufficient to provide detail in the pressure vessel and reactor assembly. Other parts of the system, however, require three-dimensional Monte Carlo analyses. To use these two methods in a single analysis, a means of coupling was developed whereby the results of a discrete ordinates calculation can be used to produce source data for a Monte Carlo calculation. Several techniques for producing source detail were investigated. Results of calculations on the NERVA system are compared and limitations and advantages of the coupling techniques discussed.

Lindstrom, D. G.↗
NASA NTRS1971

The open-cycle gas-core nuclear rocket engine - Some engineering considerations.

A preliminary design study of a conceptual 6000-MW open-cycle gas-core nuclear rocket engine system was made. The engine has a thrust of 44,200 lb and a specific impulse of 4400 sec. The nuclear fuel is uranium-235 and the propellant is hydrogen. Critical fuel mass was calculated for several reactor configurations. Major components of the reactor (reflector, pressure vessel) and the waste heat rejection system were considered conceptually and were sized.

Taylor, M. F.↗
NASA NTRS1989

Energy-Efficient, Continuous-Flow Ash Lockhopper

Pressure balance in control gas prevents loss of reactor gas. Energy efficiency of continuous-flow ash lockhopper increased by preventing hot gases from flowing out of reactor vessel through ash-hopper outlet and carrying away heat energy. Stopping loss of reactor gases also important for reasons other than energy efficiency; desired reaction product toxic or contained to prevent pollution. In improved continuous-flow ash lockhopper, pressure-driven loss of hot gas from reactor vessel through ash-hopper outlet prevented by using control gas in fluidic flow-control device to equalize pressure in reactor vessel. Also enables reactor to attain highest possible product yield with continuous processing while permitting controllable, continuous flow of ash.

Collins, Earl R., Jr.↗
NASA NTRS1971

A mini-cavity probe reactor.

The mini-cavity reactor is a rocket engine concept which combines the high specific impulse from a central gaseous fueled cavity (0.6 m diam) and NERVA type fuel elements in a driver region that is external to a moderator-reflector zone to produce a compact light weight reactor. The overall dimension including a pressure vessel that is located outside of the spherical reactor is approximately 1.21 m in diameter. Specific impulses up to 2000 sec are obtainable for 220 to 890 N of thrust with pressures less than 1000 atm. Powerplant weights including a radiator for disposing of the power in the driver region are between 4600 and 32,000 kg - less than payloads of the shuttle. This reactor could also be used as a test reactor for gas-core, MHD, breeding and materials research.

Hyland, R. E.↗
NASA NTRS1983

Ultrasonic Monitor

MicroUltrasonics PLR-1000 is a refined microprocessor-controlled version (usable on bolts, plates, liquids and gases) of the P2L2 developed by Langley Research Center. New technique is for nondestructive measurement of residual stress in various types of structures, for example, nuclear pressure vessels, pipes in nuclear reactors, offshore platforms, bridges, railroad tracks and wheels, aircraft wings, and engines. The instrument produces sound tone pulses that travel through a test specimen. PLR-1000 precisely measures speeds as stress increases speed of sound, tone changes, making precise measurements possible.

Source record↗
NASA NTRS1973

Nuclear characteristics of a fissioning uranium plasma test reactor with light-water cooling

An analytical study was performed to determine a design configuration for a cavity test reactor. Test section criteria were that an average flux of 10 to the 15th power neutrons/sq cm/sec (E less than or equal to 0.12 eV) be supplied to a 61-cm-diameter spherical cavity at 200-atm pressure. Design objectives were to minimize required driver power, to use existing fuel-element technology, and to obtain fuel-element life of 10 to 100 full-power hours. Parameter calculations were made on moderator region size and material, driver fuel arrangement, control system, and structure in order to determine a feasible configuration. Although not optimized, a configuration was selected which would meet design criteria. The driver fuel region was a cylindrical annular region, one element thick, of 33 MTR-type H2O-cooled elements (Al-U fuel plate configuration), each 101 cm long. The region between the spherical test cavity and the cylindrical driver fuel region was Be (10 vol. % H2O coolant) with a midplane dimension of 8 cm. Exterior to the driver fuel, the 25-cm-thick cylindrical and axial reflectors were also Be with 10 vol. % H2O coolant. The entire reactor was contained in a 10-cm-thick steel pressure vessel, and the 200-atm cavity pressure was equalized throughout the driver reactor. Fuel-element life was 50 hr at the required driver power of 200 MW. Reactor control would be achieved with rotating poison drums located in the cylindrical reflector region. A control range of about 18 percent delta k/k was required for reactor operation.

Whitmarsh, C. L., Jr.↗
NASA NTRS1978

Study on prestressed concrete reactor vessel structures. II-5: Crack analysis by three dimensional finite elements method of 1/20 multicavity type PCRV subjected to internal pressure

A three-dimensional finite elements analysis is reported of the nonlinear behavior of PCRV subjected to internal pressure by comparing calculated results with test results. As the first stage, an analysis considering the nonlinearity of cracking in concrete was attempted. As a result, it is found possible to make an analysis up to three times the design pressure (50 kg/sqcm), and calculated results agree well with test results.

Source record↗
NASA NTRS1992

Rotary device for removing particulates from a gas stream

A rotary particulate separator for removing particulates from a pressurized gas stream such as that emanating from a reactor vessel is disclosed which precharges the particles in the gas stream, and then utilizes the charge on the particles to induce them from the main flow path through an airblock and into the rotary particulate separator. The rotor of the rotary particulate separator has polarized plates which use a first charge opposite that on the charged particles to attract the particles as they enter the rotation chamber, and then use a second charge of the same polarity as the charge on the charged particles to release the particles into a control gas flow vortex which draws the particles radially inwardly into an exit aperture contained in the center of one of the rotor segments and out from the device. Pressure letdown devices are used to drop the pressure of both the control gas flow exiting the separator with the particles and the cleaned gas stream.

Collins, Jr., Earl R.↗
NASA NTRS2001

Creep-Fatigue Interaction Testing

Fatigue fives in metals are nominally time independent below 0.5 T(sub Melt). At higher temperatures, fatigue lives are altered due to time-dependent, thermally activated creep. Conversely, creep rates are altered by super. imposed fatigue loading. Creep and fatigue generally interact synergistically to reduce material lifetime. Their interaction, therefore, is of importance to structural durability of high-temperature structures such as nuclear reactors, reusable rocket engines, gas turbine engines, terrestrial steam turbines, pressure vessel and piping components, casting dies, molds for plastics, and pollution control devices. Safety and lifecycle costs force designers to quantify these interactions. Analytical and experimental approaches to creep-fatigue began in the era following World War II. In this article experimental and life prediction approaches are reviewed for assessing creep-fatigue interactions of metallic materials. Mechanistic models are also discussed briefly.

Halford, Gary R.↗
NASA NTRS2013

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↗
NASA NTRS1978

Heat and moisture flow in concrete as a function of temperature

Due to temperature, reactors in operation cause heat and moisture flows in the thick walled prestressed pressure vessels. These flows were studied in three beams of concrete made with crushed limestone aggregate, and in three beams made of crushed gravel/sand aggregate. The flow phenomena were related to the structural development of the concrete by determining the amount of non-evaporatable water, the total porosity, and the pore size distribution. Local temperature and moisture conditions also influenced the technical properties. Compressive strength, changes in length due to shrinkage and contraction, thermal expansion, and thermal conductivity were determined.

Hundt, J.↗
NASA NTRSDate not supplied

Coupled Reactor Multiphysics and Mass Scalability Assessment for Crewed Megawatt-Class NEP System Architectures

Nuclear Electric Propulsion (NEP) is an in-space propulsion technology capable of enabling opposition and conjunction class crewed Mars missions. NEP subsystems include the reactor for heat generation, a power conversion system (PCS), power management and distribution (PMAD), electric propulsion subsystem (EPS), and a primary heat rejection system. Specific mass, or αe (kg / kWe), is a key performance parameter (KPP) of the propulsion system which is directly scalable with the performance and mass estimates for individual components. To inform technology maturation planning, full system and component level parametric modeling is ongoing to explore the design trade space and illustrate the effect of subsystem design choices on the system KPPs. In this study, scaling of high-assay, low-enriched uranium (HALEU) reactor designs is assessed through coupled reactor physics and thermal hydraulics analyses. Scaling analyses evaluate the impact of system performance parameters (power level, interface temperatures) on mass for direct gas cooled, pumped liquid metal, and passively cooled heat pipe reactor concepts. Each concept requires specific geometries and working fluids to reach the performance goals of PCS interface conditions (temperature, pressure, flow rate) and system mass. The reactor assembly includes the active core (fuel, moderator, cladding, working fluid), axial and radial neutron reflectors, control drums, structural support / pressure vessel, and external radiation shielding. Each of these components are parametrically sized based on performance parameters for a megawatt-class power cycle. Results of this scaling analysis increase NEP propulsion system modeling fidelity and ultimately aim to support technology down-selection along with related technology development planning. The reactor and shield αe are a function of several PCS design choices, and reactor scaling with these parameters must be considered to enable an informed decision on reactor geometry and working fluid combination.

Nuclear Electric Propulsion↗
NASA NTRSDate not supplied

Coupled Reactor Multiphysics and Mass Scalability Assessment for Crewed Megawatt-Class NEP System Architectures

Nuclear Electric Propulsion (NEP) is an in-space propulsion technology capable of enabling opposition and conjunction class crewed Mars missions. NEP subsystems include the reactor for heat generation, a power conversion system (PCS), power management and distribution, electric propulsion system, and heat rejection system. Specific mass, or α (kg/kWe), is a key performance parameter (KPP) of the propulsion system which is directly scalable with the performance and mass predictions for each individual component. To inform technology maturation planning activities, full system and component level parametric modeling is ongoing to explore the design trade space and illustrate the effect of subsystem design choices on the system KPPs. In this study, scaling of high-assay, low-enriched uranium reactor designs is assessed through coupled reactor physics and thermal hydraulics analyses. Scaling analyses evaluate the impact of system performance parameters (power level, interface temperatures) on mass for direct gas cooled, pumped liquid metal, and passively-cooled heat pipe reactor concepts. Each concept requires specific geometries, fluids, and power conversion interface conditions (temperature, pressure, flow rate) to meet desired performance and mass. The reactor assembly includes the active core (fuel, moderator, cladding, working fluid), axial and radial neutron reflectors, control drums, structural support / pressure vessel, and external radiation shielding. Each of these components are parametrically sized based on performance parameters for a megawatt-class power cycle. Results of this scaling analysis increase NEP propulsion system modeling fidelity and ultimately aim to support concept down-selection along with related technology development planning. The reactor and shield α are a function of several PCS and heat rejection system design choices, and reactor scaling with these parameters must be considered to enable an informed decision on an optimal reactor geometry and working fluid combination.

Nuclear Electric Propulsion↗
NASA NTRS1998

Nuclear modules for space electric propulsion

Analysis of interplanetary cargo and piloted missions requires calculations of the performances and masses of subsystems to be integrated in a final design. In a preliminary and scoping stage the designer needs to evaluate options iteratively by using fast computer simulations. The Oak Ridge National Laboratory (ORNL) has been involved in the development of models and calculational procedures for the analysis (neutronic and thermal hydraulic) of power sources for nuclear electric propulsion. The nuclear modules will be integrated into the whole simulation of the nuclear electric propulsion system. The vehicles use either a Brayton direct-conversion cycle, using the heated helium from a NERVA-type reactor, or a potassium Rankine cycle, with the working fluid heated on the secondary side of a heat exchanger and lithium on the primary side coming from a fast reactor. Given a set of input conditions, the codes calculate composition. dimensions, volumes, and masses of the core, reflector, control system, pressure vessel, neutron and gamma shields, as well as the thermal hydraulic conditions of the coolant, clad and fuel. Input conditions are power, core life, pressure and temperature of the coolant at the inlet of the core, either the temperature of the coolant at the outlet of the core or the coolant mass flow and the fluences and integrated doses at the cargo area. Using state-of-the-art neutron cross sections and transport codes, a database was created for the neutronic performance of both reactor designs. The free parameters of the models are the moderator/fuel mass ratio for the NERVA reactor and the enrichment and the pitch of the lattice for the fast reactor. Reactivity and energy balance equations are simultaneously solved to find the reactor design. Thermalhydraulic conditions are calculated by solving the one-dimensional versions of the equations of conservation of mass, energy, and momentum with compressible flow.

Difilippo, F. C.↗