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Program solves thermal, hydraulic, or combined thermal and hydraulic problems. It can handle transient and steady-state hydraulic problems and combined thermal/hydraulic transient of steady-state problems. Physical system of interest is approximated to any degree of accuracy with lumped-parameter representation, using elements provided by program.
Thermal-hydraulics design and analyses of the Pellet Bed Reactor for nuclear thermal propulsion are performed using the nuclear propulsion thermal-hydraulic analysis model to determine the 2D steady-state temperature, pressure, and flow fields in the core and optimize the orificing in the hot-frit to avoid hot spots in the core at full power operation. Results show that by properly adjusting the axial porosity profile in the hot frit, hot spots in the core can be essentially eliminated during full power operation. This important accomplishment is achieved at the expense of slightly larger pressure losses in the core because of flow restriction at the hot frit. However, the overall pressure losses is only about 11 percent of the propellant inlet pressure.
This report reviews the state of the art of thermal-hydraulic analysis codes and presents a new code, Program ELM, for analysis of fuel elements. ELM is a concise computational tool for modeling the steady-state thermal-hydraulics of propellant flow through fuel element coolant channels in a nuclear thermal rocket reactor with axial coolant passages. The program was developed as a tool to swiftly evaluate various heat transfer coefficient and friction factor correlations generated for turbulent pipe flow with heat addition which have been used in previous programs. Thus, a consistent comparison of these correlations was performed, as well as a comparison with data from the NRX reactor experiments from the Nuclear Engine for Rocket Vehicle Applications (NERVA) project. This report describes the ELM Program algorithm, input/output, and validation efforts and provides a listing of the code.
ELM computer program is simple computational tool for modeling steady-state thermal hydraulics of flows of propellants through fuel-element-coolant channels in nuclear thermal rockets. Evaluates various heat-transfer-coefficient and friction-factor correlations available for turbulent pipe flow with addition of heat. Comparisons possible within one program. Machine-independent program written in FORTRAN 77.
Program solves both transient and steady-state thermal problems, steady-state hydraulic problems, and combined thermal, and hydraulic transient or steady-state problems.
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Thermal and hydraulic performance of prototype SNAP-8 mercury boiler - design, fabrication, and preliminary evaluation
Nuclear thermal propulsion (NTP), with its increased efficiency compared to traditional rocket engines, is a potentially enabling technology for space missions of higher complexity, such as delivering heavy machinery or crew to the moon or other planets. Thus, exploring modeling approaches that can effectively couple the thermal, fluid, and neutronic behaviors of NTP reactors is essential to mitigate cost and schedule challenges with hardware development. This paper details one such approach which utilizes Cardinal to unify reactor physics analysis performed using OpenMC, heat transfer analysis performed using the MOOSE framework, and thermal hydraulic analysis performed using the MOOSE Thermal Hydraulic Module. This approach is demonstrated with a proof of concept that focuses on a legacy reactor geometry from the historic NERVA program. This preliminary model found an overprediction in, when compared to the reference data, axial power peaking and the temperatures of both solids and fluids by roughly 20%. It is clear, however, that the model conserved a number of broad trends, and a few model refinements should reduce the observed error. This approach has definite merit and should enable the production of informative results for more modern NTP systems. This work is a first step in creating a transient reactor model that can be updated across a sequence of time steps based on continually fluctuating, time-dependent boundary conditions.
The conceptual design and analysis of a solar thermal free-piston Stirling hydraulic engine system designed to deliver 25 kWe when coupled to a 11 meter test bed concentrator is documented. A manufacturing cost assessment for 10,000 units per year was made. The design meets all program objectives including a 60,000 hr design life, dynamic balancing, fully automated control, more than 33.3 percent overall system efficiency, properly conditioned power, maximum utilization of annualized insolation, and projected production costs. The system incorporates a simple, rugged, reliable pool boiler reflux heat pipe to transfer heat from the solar receiver to the Stirling engine. The free-piston engine produces high pressure hydraulic flow which powers a commercial hydraulic motor that, in turn, drives a commercial rotary induction generator. The Stirling hydraulic engine uses hermetic bellows seals to separate helium working gas from hydraulic fluid which provides hydrodynamic lubrication to all moving parts. Maximum utilization of highly refined, field proven commercial components for electric power generation minimizes development cost and risk.
The identification of possible shorelines in the Martian northern plains suggests that the water discharged by the circum-Chryse outflow channels may have led to the formation of transient seas, or possibly even an ocean, covering as much as one-third of the planet. Speculations regarding the possible fate of this water have included local ponding and reinfiltration into the crust; freezing, sublimation, and eventual cold-trapping at higher latitudes; or the in situ survival of this now frozen water to the present day -- perhaps aided by burial beneath a protective cover of eolian sediment or lavas. Although neither cold-trapping at higher latitudes nor the subsequent freezing and burial of flood waters can be ruled out, thermal and hydraulic considerations effectively eliminate the possibility that any significant reassimilation of this water by local infiltration has occurred given climatic conditions resembling those of today. The arguments against the local infiltration of flood water into the northern plains are two-fold. First, given the climatic and geothermal conditions that are thought to have prevailed on Mars during the Late Hesperian (the period of peak outflow channel activity in the northern plains), the thickness of the cryosphere in Chryse Planitia is likely to have exceeded 1 km. A necessary precondition for the widespread occurrence of groundwater is that the thermodynamic sink represented by the cryosphere must already be saturated with ice. For this reason, the ice-saturated cryosphere acts as an impermeable barrier that effectively precludes the local resupply of subpermafrost groundwater by the infiltration of water discharged to the surface by catastraphic floods. Note that the problem of local infiltration is not significantly improved even if the cryosphere were initially dry, for as water attempts to infiltrate the cold, dry crust, it will quickly freeze, creating a seal that prevents any further infiltration from the ponded water above. The second argument against the local infiltration of flood water in the northern plains is based on hydraulic considerations. Repeated impacts have likely brecciated the Martian crust down to a depth of roughly 10 km. Given a value of permeability no greater than that inferred for the top 10 km of the Earth's crust (approximately 10(exp -2) darcies), a timescale as much as a billion years or more for the Martian groundwater system to achieve hydrostatic equilibrium, and the approximately 2-4 km elevation difference between the outflow channel source regions and the northern plains, the water confined beneath the frozen crust of the northern plains should have been under a significant hydraulic head. Thus, the existence of a hydraulic pathway between the ponded flood waters above the northern plains and the confined aquifer lying beneath it would not have led to the infiltration of flood water back into the crust, but rather the additional expulsion of groundwater onto the surface.
A series of reduced-gravity two-phase flow experiments has been conducted with a boiler/condenser apparatus in the NASA KC-135 aircraft in order to obtain basic thermal-hydraulic data applicable to analytical design tools. Several test points from the KC-135 tests were selected for simulation by means of the COBRA/TRAC two-fluid, three-field thermal-hydraulic computer code; the points were chosen for a 25-90 percent void-fraction range. The possible causes for the lack of agreement noted between simulations and experiments are explored, with attention to the physical characteristics of two-phase flow in one-G and near-zero-G conditions.
An adequate understanding is sought of flow and heat transfer behavior in reduced and zero gravity conditions. Microgravity thermal-hydraulic analysis capabilities were developed for application to space nuclear power systems. A series of reduced gravity two phase flow experiments using the NASA KC-135 were performed. The objective was to supply basic thermal hydraulic information that could be used in development of analytical tools for design of space power systems. The experiments are described. Two main conclusions were drawn. First, the tests demonstrate that the KC-135 is a suitable test environment for obtaining two phase flow and heat transfer data in reduced gravity conditions. Second, the behavior of two phase flow in low gravity is sufficiently different from that obtained in 1 g to warrant intensive investigation of the phenomenon if adequate analytical tools are to be developed for microgravity conditions.
Condensing heat transfer coefficients and two- phase pressure change measurement for forced convection condensation of potassium in vertical downflow inside single tubes
Interrupted flow, impingement cooling, and axial power distribution are employed to enhance the heat-transfer configuration of a solid-core nuclear thermal rocket engine. Impingement cooling is introduced to increase the local heat-transfer coefficients between the reactor material and the coolants. Increased fuel loading is used at the inlet end of the reactor to enhance heat-transfer capability where the temperature differences are the greatest. A thermal-hydraulics computer program for an unfueled NERVA reactor core is employed to analyze the proposed configuration with attention given to uniform fuel loading, number of channels through the impingement wafers, fuel-element length, mass-flow rate, and wafer gap. The impingement wafer concept (IWC) is shown to have heat-transfer characteristics that are better than those of the NERVA-derived reactor at 2500 K. The IWC concept is argued to be an effective heat-transfer configuration for solid-core nuclear thermal rocket engines.
The authors summarize the results of detailed neutronic and thermal-hydraulic evaluations of the heat pipe cooled thermionic (HPTI) reactor design, identify its key design attributes, and quantify its performance characteristics. The HPTI core uses modular, liquid-metal core heat transfer assemblies to replace the liquid-metal heat transport loop employed by in-core thermionic reactor designs of the past. The nuclear fuel, power conversion, heat transport, and heat rejection functions are all combined into a single modular unit. The reactor/converter assembly uses UN fuel pins to obtain a critical core configuration with in-core safety rods and reflector controls added to complete the subassembly. By thermally bonding the core heat transfer assemblies during the reactor core is coupled neutronically, thermally, and electrically into a modular assembly of individual power sources with cross-tied architecture. A forward-facing heat pipe radiator assembly extends from the reactor head in the shape of a frustum of a cone on the opposite side of the power system from the payload. Important virtues of the concept are the absence of any single-point failures and the ability of the core to effectively transfer the TFE waste heat load laterally to other in-core heat transfer assemblies in the event of multiple failures in either in-core and radiator heat pipes.
This paper proposes a mathematical model to simulate the cool-down behavior of a warm transfer line using liquid helium (including He II) in a microgravity environment. The thermal-hydraulic characteristics of different possible flow regimes which may occur during the cool-down process are discussed. Correlations for predicting the heat and momentum transfer rates in each flow regime are presented.
SP-100 Control Systems modeling was done using a thermal hydraulic transient analysis model called ARIES-S. The ARIES-S Computer Simulation provides a basis for design, integration and analysis of the reactor including the control and protection systems. It is a modular digital computer simulation written in FORTRAN that operates interactively in real time on a VAX minicomputer.
Reduced gravity two phase flow boiling and condensing experiments were conducted in the NASA KC-135. In an attempt to gain a better understanding of two phase flow behavior in reduced gravity, the individual test runs were critically examined to determine their suitability to be taken as representative of two phase flow in microgravity. Selected runs were simulated using the two fluid thermal-hydraulic COBRA/TRAC computer code. The comparisons of these sophisticated codes with the observed flow behavior illustrate the generic shortcomings of current two phase modeling capabilities in application to reduced gravity conditions.