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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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Altitude Testing of the J2-X engine at 100,000 feet (start capability). Chemical Steam Generation for providing vacuum. Project Started Mar 07. Test Stand Activation around Late 2010. J-2X Testing around early 2011.
This viewgraph presentation reviews computational fluid dynamics as a tool for modelling the dispersion of carbon monoxide at the Stennis Space Center's A3 Test Stand. The contents include: 1) Constellation Program; 2) Constellation Launch Vehicles; 3) J2X Engine; 4) A-3 Test Stand; 5) Chemical Steam Generators; 6) Emission Estimates; 7) Located in Existing Test Complex; 8) Computational Fluid Dynamics; 9) Computational Tools; 10) CO Modeling; 11) CO Model results; and 12) Next steps.
John C. Stennis Space Center is embarking on a very ambitious era in its rocket engine propulsion test history. The first new large rocket engine test stand to be built at Stennis Space Center in over 40 years is under construction. The new A3 Test Stand is designed to test very large (294,000 Ibf thrust) cryogenic propellant rocket engines at a simulated altitude of 100,000 feet. A3 Test Stand will have an engine testing chamber where the engine will be fired after the air in the chamber has been evacuated to a pressure at the simulated altitude of less than 0.16 PSIA. This will result in a very unique environment with extremely low pressures inside a very large chamber and ambient pressures outside this chamber. The test chamber is evacuated of air using a 2-stage diffuser / ejector system powered by 5000 lb/sec of steam produced by 27 chemical steam generators. This large amount of power and flow during an engine test will result in a significant acoustic and vibrational environment in and around A3 Test Stand.
A pilot operational ISHM capability has been implemented for the E-2 Rocket Engine Test Stand (RETS) and a Chemical Steam Generator (CSG) test article at NASA Stennis Space Center. The implementation currently includes an ISHM computer and a large display in the control room. The paper will address the overall approach, tools, and requirements. It will also address the infrastructure and architecture. Specific anomaly detection algorithms will be discussed regarding leak detection and diagnostics, valve validation, and sensor validation. It will also describe development and use of a Health Assessment Database System (HADS) as a repository for measurements, health, configuration, and knowledge related to a system with ISHM capability. It will conclude with a discussion of user interfaces, and a description of the operation of the ISHM system prior, during, and after testing.
NASA Stennis Space Center has implemented a pilot operational Integrated System Health Management (ISHM) capability. The implementation was done for the E-2 Rocket Engine Test Stand and a Chemical Steam Generator (CSG) test article; and validated during operational testing. The CSG test program is a risk mitigation activity to support building of the new A-3 Test Stand, which will be a highly complex facility for testing of engines in high altitude conditions. The foundation of the ISHM capability are knowledge-based integrated domain models for the test stand and CSG, with physical and model-based elements represented by objects the domain models enable modular and evolutionary ISHM functionality.
During the Fall of 2010, I spent my time at NASA White Sands Test Facility in Las Cruces, NM as an Undergraduate Student Research Program (USRP) Intern. During that time, I was given three projects to work on: Large Altitude Simulation System (LASS) basket strainer, log books, and the design of a case for touch screen monitors used for simulations. I spent most of my time on the LASS basket strainer. The LASS system has a water feed line with a basket strainer that filters out rust. In 2009, there were three misfires which cost approximately $27,000 and about 8% of the allotted time. The strainer was getting a large change in pressure that would result in a shutdown of the system. I have designed a new basket that will eliminate the large pressure change and it can be used with the old basket strainer housing. The LASS system has three steam generators (modules). Documents pertaining to these modules are stored electronically, and the majority of the documents are not able to be searched with keywords, so they have to be gone through one by one. I have come up with an idea on how to organize these files so that the Propulsion Department may efficiently search through the documents needed. Propulsion also has a LASS simulator that incorporates two touch screen monitors. Currently these monitors are in six foot by two foot metal cabinet on wheels. During simulation these monitors are used in the block house and need to be taken out of the block house when not in use. I have designed different options for hand held cases for storing and transporting the monitors in and out of the block house. The three projects previously mentioned demonstrate my contributions to the Propulsion Department and have taught me real world experience that is essential in becoming a productive engineer.
A thermal sleeve is mechanically attached to the bore of a surge nozzle of a pressurizer for the primary circuit of a pressurized water reactor steam generating system. The thermal sleeve is attached with a series of keys and slots which maintain the thermal sleeve centered in the nozzle while permitting thermal growth and restricting flow between the sleeve and the interior wall of the nozzle.
High chromium, nickel-base Alloy 690 exhibits an increased resistance to stress corrosion cracking (SCC) in pressurized water reactor (PWR) primary water environments over lower chromium alloy 600. As a result, Alloy 690 has been used to replace Alloy 600 for steam generator tubing, reactor pressure vessel nozzles and other pressure boundary components. However, recent laboratory crack-growth testing has revealed that heavily cold-worked Alloy 690 materials can become susceptible to SCC. To evaluate reasons for this increased SCC susceptibility, detailed characterizations have been performed on as-received and cold-worked Alloy 690 materials using electron backscatter diffraction (EBSD) and Vickers hardness measurements. Examinations were performed on cross sections of compact tension specimens that were used for SCC crack growth rate testing in simulated PWR primary water. Hardness and the EBSD integrated misorientation density could both be related to the degree of cold work for materials of similar grain size. However, a microstructural dependence was observed for strain correlations using EBSD and hardness which should be considered if this technique is to be used for gaining insight on SCC growth rates
A relatively simple method using the nominal constant average stress information and the creep rupture model is developed to predict the creep-fatigue lifetime of Alloy 617, in terms of time to rupture. The nominal constant average stress is computed using the stress relaxation curve. The predicted time to rupture can be converted to number of cycles to failure using the strain range, the strain rate during each cycle, and the hold time information. The predicted creep-fatigue lifetime is validated against the experimental measurements of the creep-fatigue lifetime collected using conventional laboratory creep-fatigue tests. High temperature creep-fatigue tests of Alloy 617 were conducted in air at 950°C with a tensile hold period of up to 1800s in a cycle at total strain ranges of 0.3% and 0.6%. It was observed that the proposed method is conservative in that the predicted lifetime is less than the experimentally determined values. The approach would be relevant to calculate the remaining useful life to a component like a steam generator that might fail by the creep-fatigue mechanism.
The bootstrap start-up system (42) achieves an efficient start-up of the power plant (10) that minimizes formation of soot within a reformed hydrogen rich fuel. A burner (48) receives un-reformed fuel directly from the fuel supply (30) and combusts the fuel to heat cathode air which then heats an electrolyte (24) within the fuel cell (12). A dilute hydrogen forming gas (68) cycles through a sealed heat-cycling loop (66) to transfer heat and generated steam from an anode side (32) of the electrolyte (24) through fuel processing system (36) components (38, 40) and back to an anode flow field (26) until fuel processing system components (38, 40) achieve predetermined optimal temperatures and steam content. Then, the heat-cycling loop (66) is unsealed and the un-reformed fuel is admitted into the fuel processing system (36) and anode flow (26) field to commence ordinary operation of the power plant (10).
Beginning in 2015, a project was initiated to update and modernize the probabilistic safety assessment (PSA) of the GE-Hitachi PRISM sodium fast reactor. This project is a collaboration between GE-Hitachi and Argonne National Laboratory (Argonne), and funded in part by the U.S. Department of Energy. Specifically, the role of Argonne is to assess the reliability of passive safety systems, complete a mechanistic source term calculation, and provide component reliability estimates. The assessment of passive system reliability focused on the performance of the Reactor Vessel Auxiliary Cooling System (RVACS) and the inherent reactivity feedback mechanisms of the metal fuel core. The mechanistic source term assessment attempted to provide a sequence specific source term evaluation to quantify offsite consequences. Lastly, the reliability assessment focused on components specific to the sodium fast reactor, including electromagnetic pumps, intermediate heat exchangers, the steam generator, and sodium valves and piping.
Alloy 800H is the leading candidate for the steam generator in Next Generation Nuclear Plant (NGNP) Project. It is necessary to extend the allowable stresses for Alloy 800H to 900 degrees C and 600,000 hours. This also requires data on welds in order to extend code acceptance. The focus of this presentation is on the creep-rupture properties of Alloy 800H cross welds with Alloy 617 filler. The goal is to understand the effect of an overmatched weld on creep-rupture properties.
A consequence of operating steam cycles under time varying loads, is the variability of the feedwater temperature entering the bottom of the steam generators. This variability is a direct result of changes in turbine impulse pressure and subsequently turbine tap pressures in response to the load changes, with a reduction in feed temperature during periods of low turbine demand. These fluctuations over time will increase thermal cycling and threaten the integrity of the tube bundle as well as other components in the feed system. Sensible heat, TES systems have been shown to be an effective load management strategy allowing nuclear reactor systems to operate at effectively 100% full power while storing excess thermal energy for recovery later. A byproduct of the TES system is the availability of large amounts of low grade waste heat that can be used for ancillary applications, including auxiliary feed heating during periods of low turbine demand. The addition of an auxiliary feed heater at the end of the traditional feed train can minimize feed temperature variations during periods of low turbine demand, reducing the effects of thermal cycling and increasing overall thermodynamic efficiency. This paper discusses the design, operation and control strategies for an auxiliary feed water system as an integral component of a coupled Thermal Energy Storage system and Integral Pressurized Water Reactor. The impact on system performance, particularly as it affects reactor dynamic response and reductions in thermal cycling will be addressed. In addition to the improved thermal efficiency, a benefit of this approach is the relocation of thermal stressors from the reactor coolant system to more easily managed Balance of Plant systems
Chemical looping technology is advancing toward large pilot-scale demonstrations. When scaling chemical looping reactors and chemical looping system integrated with steam generation issues of fuel distribution, thermal integration and modular operation must be considered in addition to sparing philosophy. The keynote addressed discusses a number of scaling issues unique to chemical looping systems.
Nuclear power plants (NPPs) are becoming cost prohibitive to operate. This is due to historically low natural gas prices and wind and solar prices often being less than the marginal cost of operating baseload coal and NPPs, which sometimes forces NPPs to pay for the electricity they create. Thus, the feasibility of using part of the heat from an NPP to create other commodities such as hydrogen is being evaluated; operating in this way is called flexible power operation (FPO). To change the NPP design and operation, a license review must happen. Part of the review is the Probabilistic Risk Assessment (PRA) which proves the safety of the design and operation by identifying sequences of events that result in plant failure and estimating their likelihood of occurrence. A Level 1 PRA model that evaluates the risk of removing heat from the secondary side of NPP was created. The model was developed using Systems Analysis Program for Hands-On Integrated Reliability Evaluations (SAPHIRE 8) software to calculate core damage frequency (CDF). It is a fault tree that represents the possible pathways that could lead to a loss of steam inventory in the secondary loop. A loss of steam inventory can induce steam generator tube rupture or other events that activates response systems to avoid core damage. This is the main concern for the Nuclear Regulatory Commission (NRC), the group in the United States that licenses NPPs. This model will calculate a total CDF of any of these pathways occurring which will be compared to the baseline CDF in a generic NPP PRA for licensing support.
Solar Vaporization using heat localization under solar irradiation has the potential to serve as an inexpensive and viable component of seawater desalination processes, industrial wastewater treatment, hydroponic greenhouse humidifiers, and steam generation systems. In this paper we tested different porosities of carbon foam (CF) and combinations of CF as interface media at 1.2 sun solar concentration, in search of a suitable candidate for use with distillation process. We developed an experimental condition to analyze how the heat localization is achieved in each case and investigated vapor temperatures, surface temperatures with single CF as interface. The rate of evaporation, temperature distribution in each combination is studied and suitable combination is selected to use in further testing of desalination model.
Electron beam welding (EBW) is commonly employed by industry today to join relatively thin section components. Industrial sectors which utilize the technology include; automotive, aircraft, aerospace, defense, medical, oil & gas, etc. The ability to produce single pass, thick section (up to 6 inches, 150mm) welds has also been available for more than 40 years; however, the technology has not been deployed for civil nuclear manufacturing and fabrication such as reactor pressure vessels, steam generators, or pressurizers. Instead, thick section joining is performed much in the same manner it has been for decades using submerged arc or gas tungsten arc welding processes. Recent research performed under U.S. Department of Energy (DOE) project DE-NE0008629 has demonstrated that the overall welding time to produce one major reactor pressure vessel (RPV) girth weld can be reduced from weeks to less than 90 minutes, while maintaining exceptional quality and inspection characteristics. The subject project (DE-NE0008846) investigates a modular in-chamber (MIC) EBW approach which allows various RPV sections to be joined within a “modular” chamber. The MIC-EBW approach utilizes individual modules that can be stacked and de-stacked to various heights to enable production girth welds to be performed at a given elevation. This report describes the various processing methodology necessary to complete major girth weldments for the NuScale Power small modular reactor including welding development, manufacturing, production sequence, inspection, and cladding.
Pool boiling heat transfer is one of the most effective heat transfer methos and crucial for several energy conversion processes such as electronic cooling, steam generation, water purification and thermal management. Enhancing boiling heat transfer around the tubes enables higher heat flux, which can improve the efficiency and lead to compact heat exchangers. Delaying the onset of critical heat flux (CHF) allows for higher CHFs and protects equipment during anomalous transient conditions. Growing interest in process intensification and delayed CHF has motivated the current study. An experimental approach has been deployed to investigate a potential technique to enhance the pool boiling performance around small diameter tubes by open cell metal foams. The findings for nucleate pool boiling of Novec HFE 7000 on the externally enhanced and smooth tubes are reported. The enhanced tube comprised of compressed aluminum foam brazed around. All the samples were horizontally oriented, and the experiments were performed under ambient pressure conditions. The heat transfer measurements for the smooth tubes were compared with several existing correlations and established the baseline. The preliminary results showed that the aluminum foam effectively enhanced the heat transfer coefficient and delayed the onset of critical heat flux. In addition, the pool boiling process around the round tube sample was visualized from both side and front views to comprehend the heat transfer mechanisms underlying substantial enhancement observed during experiments.