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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 127 records · Page 7
All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

WISE-CASING: Time Domain Reflectometry Data from Cymric Field, CA

The objective of this field test is to validate several technologies for non-invasive well integrity assessment using existing wells with a known completion. The tests were made at the Cymric oil field, which is a steam flood operation. The wells therefore undergo similar downhole conditions as geothermal wells. The Cymric field is mainly a cyclic steam operation where wells are 1000-15-00 ft in depth and the reservoir occupies the bottom 400ft. The maximum temperatures can exceed 500 degrees F and the well spacing is very close, often less than 50m. The field plan consisted of applying the Time Domain Reflectometry (TDR) method to the wells. The input voltages were set as 70 V shows the TDR responses at frequencies of 450 kHz, 2500 kHz, and 4500 kHz. There is a summary report will full information about the field tests.

15 GEOTHERMAL ENERGY↗
DOE OSTI2024

Numerical investigation of the AP1000 response following loss-of-coolant accident using PCTRAN and CFD to support R&D of SMRs

Small Modular Reactors (SMRs) present a promising solution for the future of sustainable energy, offering advantages such as reduced waste generation, advanced passive safety features, and potential cost efficiencies. Ongoing research focuses on the design and development of SMRs, addressing challenges through numerical simulations and experimental test data. Given that many next-generation reactors are first-of-a-kind (FOK), current operational reactors serve as valuable benchmarks for understanding various thermal-hydraulic phenomena during postulated design basis accidents (DBAs), including loss-of-coolant accidents (LOCA), main steam line breaks (MSLB), and steam generator tube ruptures (SGTR). This study employs the AP1000 model Personal Computer Transient Analyzer (PCTRAN) to simulate reactor responses to small-break LOCA scenarios, specifically analyzing break sizes of 2 and 10 inches. Additionally, ANSYS FLUENT software is utilized to assess containment responses to large-break LOCAs, focusing on the quantification of decay heat removal via natural convection. Verification of the PCTRAN results is achieved using Westinghouse data, with CFD results corroborating the findings. The outcomes demonstrate strong agreement with Westinghouse data, confirming the accuracy of the simulations.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Condition-Based Maintenance of a Circulating Water System of a Canadian Nuclear Power Plant using Machine Learning and Statistical Tools

Canada Deuterium Uranium pressurized-heavy-water reactors (PHWR) are a type of nuclear power plant that generate clean and reliable energy. The scope of this work is to automate data analysis methodologies to inform a condition-based maintenance strategy of a circulating water system (CWS) of a PHWR. The multiunit CWS provides a continuous supply of water to cool steam condensers, even during transient scenarios, thereby improving the thermal efficiency. This work aims to develop a machine learning (ML) based approach to detect anomalies in heterogeneous data of a CWS in a PHWR to help inform a predictive maintenance strategy. The heterogeneous data include textual and numeric time series data for a PHWR. Natural-language-processing (NLP)-based models are used to analyze textual data contained in work orders and operator logs and an event-timeseries correlation detection method is applied to assist anomalies diagnoses for CWS. An ML model Robust Linear Model (RLM) is also used to remove the seasonal variations in the system variable distributions based on distributions of environmental variables. A machine learning model, Density-Based Spatial Clustering of Applications with Noise (DBSCAN), trained on both original data and data without any seasonal variations will then be used to detect if an anomaly exists. Thus, by moving to an automated methodology to detect, classify, and forecast anomalies, the maintenance strategy would be based on component condition instead of a time-based schedule.

97 - MATHEMATICS AND COMPUTING↗
DOE OSTI2016

Exfoliation Propensity of Oxide Scale in Heat Exchangers Used for Supercritical CO2 Power Cycles

Supercritical CO2 (sCO2) Brayton cycle systems offer the possibility of improved efficiency in future fossil energy power generation plants operating at temperatures of 650 C and above. As there are few data on the oxidation/corrosion behavior of structural alloys in sCO2 at these temperatures, modeling to predict the propensity for oxide exfoliation is not well developed, thus hindering materials selection for these novel cycles. The ultimate goal of this effort is to provide needed data on scale exfoliation behavior in sCO2 for confident alloy selection. To date, a model developed by ORNL and EPRI for the exfoliation of oxide scales formed on boiler tubes in high-temperature, high-pressure steam has proven useful for managing exfoliation in conventional steam plants. A major input provided by the model is the ability to predict the likelihood of scale failure and loss based on understanding of the evolution of the oxide morphologies and the conditions that result in susceptibility to exfoliation. This paper describes initial steps taken to extend the existing model for exfoliation of steam-side oxide scales to sCO2 conditions. The main differences between high-temperature, high-pressure steam and sCO2 that impact the model involve (i) significant geometrical differences in the heat exchangers, ranging from standard pressurized tubes seen typically in steam-producing boilers to designs for sCO2 that employ variously-curved thin walls to create shaped flow paths for extended heat transfer area and small channel cross-sections to promote thermal convection and support pressure loads; (ii) changed operating characteristics with sCO2 due to the differences in physical and thermal properties compared to steam; and (iii) possible modification of the scale morphologies, hence properties that influence exfoliation behavior, due to reaction with carbon species from sCO2. The numerical simulations conducted were based on an assumed sCO2 operating schedule and several generic heat exchanger channel shapes and cross-sectional areas. Implications for the evolution of stresses in the oxide scales formed on sCO2 heat exchangers, and ensuing critical oxide thicknesses for exfoliation, were derived and compared with expectations for an equivalent conventional tubular heat exchanger in a steam cycle (for a given alloy).

Sabau, Adrian S↗
DOE OSTI2023

Plan for a Data Base Approach to Evaluating LWR Plant Integration with Hydrogen and Industrial Facilities

The purpose of this report is to provide an outline and scope of a database tool that could be used to identify the most beneficial locations to site a hydrogen, chemical facility or other industrial facility that would be powered by nuclear energy. This energy may be in the form of electricity produced by the plant, heat in the form of steam or both. In addition, this siting tool should consider factors such as economics, geography, industries, residences, geological hydrogen storage, and key infrastructure. The infrastructure to be considered could include pipelines or other storage and transportation facilities to carry the created products from the production facility to the end user. Also considered should be infrastructure that could deliver needed chemicals or gasses to allow the production of desired products such as ammonia or biofuels. This data base tool should allow a screening study of all U.S. nuclear plants to identify opportunities and barriers to creating hydrogen hubs around nuclear power plants. For this effort, it has been decided that the Siting Tool for Advanced Nuclear Development (STAND) would be used to conduct this screening study which is a follow-on research project to this present effort. Currently, plans are being developed to modify the STAND tool to provide the capability necessary to perform screening of each US commercial nuclear plant to identify the best locations to site hydrogen or chemical production facilities. Many features of the current STAND tool will be employed to meet this objective, but additional infrastructure data (such as the location of pipelines) may need to be added to allow these locations to be fully vetted. In addition, new analysis capabilities may need to be added so economic considerations can be understood

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Strength and rupture geometry of un-irradiated C26M FeCrAl under LOCA burst testing conditions

Ferritic iron-chromium-aluminum (FeCrAl) alloys are an accident tolerant fuel candidate to replace the incumbent Zr-based claddings. Nuclear grade FeCrAl alloys are marked by superior high temperature mechanical behavior and exceptional steam oxidation resistance, both of which increase safety margins during accident scenarios. In the present study, the loss of coolant accident (LOCA) burst behaviors of three un-irradiated commercially fabricated cladding materials in a simulated LOCA environment were compared: (1) T35Y2, a 1 st generation nuclear grade FeCrAl, (2) C26M, a 2 nd generation nuclear grade FeCrAl, and (3) Zircaloy-2. Both FeCrAl alloys showed improved mechanical strength and steam oxidation resistance compared to Zircaloy-2. C26M claddings burst at significantly higher temperatures for all tested engineering hoop stresses, had limited ballooning, and demonstrated preferential fuel retention behavior in terms of burst opening area and length. High temperature tensile data for C26M is also presented. For both un-irradiated FeCrAl alloys, it was found that a distinct “threshold” burst stress signified the transition between small and large openings. Finally, higher threshold hoop stresses were associated with higher uniaxial strength for the FeCrAl alloys, indicating that tensile data, rather than creep data, could be useful for predicting rupture size and assessing fuel dispersal concerns.

36 MATERIALS SCIENCE↗
DOE OSTI2016

The MELCOR Analysis of Chinshan Nuclear Power Plant Spent Fuel Pool for Fukushima-like Accident

The safety analysis of the nuclear power plant (NPP) is very important work in the NPP safety. Especially after the Fukushima NPP accident occurred, the importance of NPP safety analysis has been raised and there is more concern for the safety of NPPs in Taiwan. Because the earthquake and tsunami occurred, the cooling system of spent fuel pool failed and the safety issue of spent fuel pool generated in Japan's Fukushima NPP. Chinshan NPP is the first NPP in Taiwan which is BWR/4 plant and OLTP (Original Licensed Thermal Power) for each unit is 1775 MWt. Chinshan NPP finished SPU (stretch power uprate) and the operating power is 103.66% of the OLTP, which is 1840 MWt now. After Fukushima NPP accident occurred, in order to concern the safety of Chinshan NPP spent fuel pool, we performed the safety analysis of spent fuel pool by using MELCOR/SNAP which also assumed the cooling system of spent fuel pool failed. The geometry of Chinshan NPP spent fuel pool is 12.17 m x 7.87 m x 11.61 m and the initial condition is 60 deg. (water temperature) / 1.013 x 10{sup 5} Pa. And, the total power of the fuels is roughly 8.9 MWt initially. MELCOR is developed by Sandia National Laboratories. MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor NPPs, including containment response and leakage of activity to the environment. A broad spectrum of severe accident phenomena in both BWRs and PWRs is treated in MELCOR in a unified framework. These include thermalhydraulic response in the reactor coolant system, reactor cavity, containment, and confinement buildings; core heatup, degradation, and relocation; core-concrete attack; hydrogen production, transport, and combustion; fission product release and transport behavior. Current uses of MELCOR include estimation of severe accident source terms and their sensitivities and uncertainties in a variety of applications. According to Dr. Carbajo's paper, MELCOR was used to perform the study of spent fuel pool of Fukushima Daiichi Unit 4. Niina E. Koenoenen used MELCOR to establish the spent fuel pool model of a Nordic BWR. The loss-of-pool-cooling accidents was simulated and analyzed by this model. The above studies indicate that MELCOR is capable of handling the simulation of the spent fuel pool. An increasing number of researchers are using MELCOR code to analyze test facilities and nuclear power plants. Jun Wang et al. used MELCOR to perform core degradation simulation and MELCOR results were compared with CORA experimental data. Through this work, they reviewed the performance of MELCOR COR package in detail. Tae Woon Kim et al. developed MELCOR model of APR1400 (an advanced pressurized water reactor). They used this model to perform the LBLOCA transient simulation and sensitivity study. Longze Li et al. established the MELCOR model of CPR1000 (a Chinese pressurized reactor 1000-MW power plant). The station blackout (SBO) with failure of the steam generator (SG) safety relief valve (SRV) transient was simulated and analyzed by this model. SNAP is a graphic user interface program that processes the inputs and outputs of MELCOR. In addition, there is an animation function in SNAP which can present the animation of analysis results. There are two main steps in this research. The first step was the establishment of Chinshan NPP spent fuel pool MELCOR/SNAP model. And the transient analysis of Chinshan NPP spent fuel pool MELCOR/SNAP model under the cooling system failure condition was performed. The next step was the comparison among the MELCOR, TRACE and CFD data. TRACE and CFD data were from INER reports. In addition, the animation model of Chinshan NPP spent fuel pool was presented by using the animation function of SNAP with MELCOR/SNAP analysis results. This study has developed MELCOR/SNAP model of Chinshan NPP spent fuel pool successfully. By using the above model, the safety analysis of the spent fuel pool was performed under the cooling system of spent fuel pool failed condition. The analysis results of MELCOR, TRACE and CFD were similar in this case. It indicated that there was a respectable accuracy in MELCOR/SNAP model. The analysis results depicted that the uncovered of the fuels occurred at 2.6 day and the metal-water reaction of fuels occurred roughly at 3.7 day after the cooling system failed. The above results indicated that the failure of cladding occurred after 3.7 day. This study's results can help to evaluate the safety issue of Chinshan NPP spent fuel pool. In addition, the cooling system and some mitigation measures of spent fuel pool is not simulated in the above model. However, we will do it in the future. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

AOI 3 Life Modelling of Critical Steam Cycle Components in Coal-Fueled Power Plants

Microstructural damage accumulation models have been used to produce calibrated life estimation models for a DR22/P22 steel wye-block welds, and a Jethete stainless steel turbine blade (bucket). The calibrated life estimation models will aid the power plant operator in determining optimal maintenance and operation schedules based upon historical operational data as well as current, or future operation schemes. The impact of this project will enable existing coal-fueled power plants to operate safely for longer periods of time and at higher efficiencies, thereby reducing the economic and environmental impact of the existing coal power plant fleet. Testing, characterization, and modelling indicates that the operating life of P22 pipelines and their welds are dominated by fatigue damage mechanisms. Specifically, fatigue is of no concern in these materials when operating under realistic conditions manifesting in the main steam piping of coal-fueled power plants. However, if a low-temperature overload ever occurs during operation, fatigue will manifest as a damage mechanism of interest. Primary impact provided by the completion of this work is in the manifestation of a detailed ABAQUS solid model providing accurate boundary conditions to enable the prediction of operational stresses and strains. The completed plant solid model, in conjunction with the calibrated fatigue and creep life models provide definitive confirmation that creep is the dominant damage mechanism during operational conditions. Jethete life modelling has been completed by the manifestation of a material-specific, and temperature-specific, Kitagowa diagram. The Kitagowa diagram provides maintenance and operation decision making with scientifically-based go/no-go support based upon the crack-like features that have been identified by use of inspection.

01 COAL, LIGNITE, AND PEAT↗
DOE OSTI2016

Large Eddy Simulations for a Helical Coil Steam Generator

The desire for scalable power plants in remote locations and specialized applications has brought forth the small modular reactor (SMR). Now, smaller light water reactors are seen as viable means for meeting global energy requirements. Generally, straight or U-shape tubes are used to exchange heat from the primary to secondary coolant; however this is not ideal when size is a criterion. Helical coil steam generators (HCSGs) are advantageous for SMRs due to their enhanced heat transfer in compact spaces. Still, there is the concern of safety as steam generator tubes could rupture due to vibrations and fatigue on the material. Understanding flow characteristics and quantifying pressures for these tubes are important components for the design of safety and efficiency in these steam generators. The current study uses the spectral element computational fluid dynamic (CFD) code, Nek5000, to develop large eddy simulations (LES) for validation against experiments that will be conducted at Texas A and M University. Time-averaged pressure data will be compared with pressure tap readings, and velocity profile with particle image velocimetry (PIV) measurements. The presented preliminary results investigate the differences between 5.-, 7.-, and 9.-order polynomial simulations. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

GOOML - Real World Applications of Machine Learning in Geothermal Operations

GOOML (Geothermal Operational Optimization with Machine Learning) is a machine-learning based framework that enables geothermal power plant operators to explore optimization opportunities for their assets in an efficient and robust digital environment. Backed by real-world data sources, thermodynamic constraints and steamfield intelligence, the GOOML environment provides new tools to explore how to best operate steamfields as well as test new scenarios and configurations prior to implementation in the field. To prove the effectiveness of GOOML, we have undertaken optimization experiments using reinforcement learning (RL) to generate operational suggestions using a balance of mass-take targets, sustainability considerations and net generation. Our experiments use the GOOML construct to explore different field parameters and perform multiple reinforcement learning experiments. Like a comprehensive laboratory workbench, we can change out components of a steamfield to perform testing under a variety of conditions (restrict mass, increase pressure, reroute steam, etc.). This flexibility allows us to explore conditions that would require significant infrastructure changes in a real-world setting at a fraction of the cost and time in a digital environment. The results highlight the benefits of using digital twins and advanced data analytics for the geothermal industry.

forecasting↗
NASA NTRS1981

The small community solar thermal power experiment. Parabolic dish technology for industrial process heat application

Aspects of incorporating a thermal energy transport system (ETS) into a field of parabolic dish collectors for industrial process heat (IPH) applications were investigated. Specific objectives are to: (1) verify the mathematical optimization of pipe diameters and insulation thicknesses calculated by a computer code; (2) verify the cost model for pipe network costs using conventional pipe network construction; (3) develop a design and the associated production costs for incorporating risers and downcomers on a low cost concentrator (LCC); (4) investigate the cost reduction of using unconventional pipe construction technology. The pipe network design and costs for a particular IPH application, specifically solar thermally enhanced oil recovery (STEOR) are analyzed. The application involves the hybrid operation of a solar powered steam generator in conjunction with a steam generator using fossil fuels to generate STEOR steam for wells. It is concluded that the STEOR application provides a baseline pipe network geometry used for optimization studies of pipe diameter and insulation thickness, and for development of comparative cost data, and operating parameters for the design of riser/downcomer modifications to the low cost concentrator.

Polzien, R. E.↗
NASA NTRS1993

Mixing and transient interface condensation of a liquid hydrogen tank

Experiments were conducted to investigate the effect of axial jet-induced mixing on the pressure reduction of a thermally stratified liquid hydrogen tank. The tank was nearly cylindrical, having a volume of about 0.144 cu m with 0.559 m in diameter and 0.711 m long. A mixer/pump unit, which had a jet nozzle outlet of 0.0221 m in diameter was located 0.178 m from the tank bottom and was installed inside the tank to generate the axial jet mixing and tank fluid circulation. The liquid fill and jet flow rate ranged from 42 to 85 percent (by volume) and 0.409 to 2.43 cu m/hr, respectively. Mixing tests began with the tank pressure ranging from 187.5 to 238.5 kPa at which the thermal stratification results in 4.9 to 6.2 K liquid sub cooling. The mixing time and transient vapor condensation rate at the liquid-vapor interface are determined. Two mixing time correlations, based on the thermal equilibrium and pressure equilibrium, are developed. Both mixing time correlations are expressed as functions of system and buoyancy parameters and compared well with other experimental data. The steady state condensation rate correlation of Sonin et al. based on steam-water data is modified and expressed as a function of jet subcooling. The limited liquid hydrogen data of the present study shows that the modified steady state condensation rate correlation may be used to predict the transient condensation rate in a mixing process if the instantaneous values of jet sub cooling and turbulence intensity at the interface are employed.

Lin, C. S.↗
DOE OSTI2012

Effect of steam generator configuration in a loss of the RHR during mid-loop operation at PKL facility

The loss of the residual heat removal system in mid-loop conditions may occur with a non-negligible contribution to the plant risk, so the analysis of the accidental sequences and the actions to mitigate the accident are of great interest in shutdown conditions. In order to plan the appropriate measures to mitigate the accident is necessary to understand the thermal-hydraulic processes following the loss of the residual heat removal system during shutdown. Thus, transients of this kind have been simulated using best-estimate codes in different integral test facilities and compared with experimental data obtained in different facilities. In PKL (Primaerkreislauf-Versuchsanlage, primary coolant loop test facility) test facility different series of experiments have been undertaken to analyze the plant response in shutdown. In this context, the E3 and F2 series consist of analyzing the loss of the residual heat removal system with a reduced inventory in the primary system. In particular, the experiments were developed to investigate the influence of the steam generators secondary side configuration on the plant response, what involves the consideration of different number of steam generators filled with water and ready for activation, on the heat transfer mechanisms inside the steam generators U-tubes. This work presents the results of such experiments calculated using, RELAP5/Mod 3.3. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Dynamic Data Reconciliation and Validation of a Dynamic Model for Solvent-Based CO 2 Capture Using Pilot-Plant Data

This paper is focused on the development and validation of a dynamic model for an MEA-based CO 2 capture unit. Starting with a rigorous steady-state process model, the dynamic model is developed in Aspen Plus Dynamics ® using a modified Murphree-efficiency approach and is validated using dynamic data collected from the National Carbon Capture Center (NCCC) in Wilsonville, Alabama. The dynamic test runs were designed with due consideration of process excitation and process nonlinearities constrained by the available time to conduct the test runs and implement the experimental designs in the plant control system. The experimental data include solvent composition and loading. Since the experimental data were found to violate mass and energy balances, filtering algorithms coupled with dynamic data reconciliation techniques were used. Without adjusting any model parameters, the dynamic model satisfactorily predicted dynamic response of CO 2 capture due to step changes in the solvent flowrate, flue gas flowrate, and steam flowrates. Transient studies show that the process gain and time constants can considerably change depending on the direction of the disturbance or manipulated variables, indicating process nonlinearities. Due to a storage tank between the absorber and stripper, the time constant of the full plant can be considerably longer than the time constants of the absorber or stripper individually. Furthermore, impact of the control system on the process efficiency while following an optimal CO 2 capture schedule was also studied.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
NASA NTRS1980

Cogeneration Technology Alternatives Study (CTAS). Volume 3: Energy conversion system characteristics

Six current and thirty-six advanced energy conversion systems were defined and combined with appropriate balance of plant equipment. Twenty-six industrial processes were selected from among the high energy consuming industries to serve as a frame work for the study. Each conversion system was analyzed as a cogenerator with each industrial plant. Fuel consumption, costs, and environmental intrusion were evaluated and compared to corresponding traditional values. The advanced energy conversion technologies indicated reduced fuel consumption, costs, and emissions. Fuel energy savings of 10 to 25 percent were predicted compared to traditional on site furnaces and utility electricity. With the variety of industrial requirements, each advanced technology had attractive applications. Fuel cells indicated the greatest fuel energy savings and emission reductions. Gas turbines and combined cycles indicated high overall annual savings. Steam turbines and gas turbines produced high estimated returns. In some applications, diesels were most efficient. The advanced technologies used coal derived fuels, or coal with advanced fluid bed combustion or on site gasifications. Data and information for both current and advanced energy conversion technology are presented. Schematic and physical descriptions, performance data, equipment cost estimates, and predicted emissions are included. Technical developments which are needed to achieve commercialization in the 1985-2000 period are identified.

Source record↗
DOE OSTI2023

Coupling of CTF and RELAP5-3D Within an Enhanced Fidelity Nuclear Power Plant Simulator

A robust and accurate multiphysics engineering simulator is being developed to model the core behavior and system response of pressurized water reactors. This simulator relies on the NESTLE and CTF computer codes to model the neutronics and thermal hydraulics (TH), respectively, inside the core on a nodal scale and on the Reactor Excursion and Leak Analysis Program—Three Dimensional (RELAP5-3D) to model the entire nuclear steam supply system. The RELAP5-3D model includes highly detailed nodalization and multidimensional flow modeling throughout the vessel. Previously, pin-resolved data generated via the Virtual Environment for Reactor Analysis core simulator were used to improve the accuracy of the NESTLE core predictions. The engineering simulator being developed as part of this work uses the 3KEYMASTER platform to couple the enhanced NESTLE model to a nodal-fidelity CTF model to balance run time with accuracy; NESTLE provides node-dependent powers to CTF, and CTF provides node-dependent coolant densities and fuel temperatures to NESTLE.An overlapping domain approach is used for the core TH in which RELAP5-3D provides core boundary conditions based on the system response and CTF provides a node-dependent coolant heating rate to the RELAP5-3D core solution. In the preliminary TH demonstration discussed in this paper, CTF and RELAP5-3D provided similar steady-state core predictions, indicating the hydraulic compatibility between the codes, as well as reasonable and expected behavior under hypothetical transient conditions. This provides an initial step in ongoing efforts toward a robust, multiscale TH/neutronics engineering simulator capability.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRS1978

A study of trends and techniques for space base electronics

A sputtering system was developed to deposit aluminum and aluminum alloys by the dc sputtering technique. This system is designed for a high level of cleanliness and for monitoring the deposition parameters during film preparation. This system is now ready for studying the deposition and annealing parameters upon double-level metal preparation. A technique recently applied for semiconductor analysis, the finite element method, was studied for use in the computer modeling of two dimensional MOS transistor structures. It was concluded that the method has not been sufficiently well developed for confident use at this time. An algorithm was developed for confident use at this time. An algorithm was developed for implementing a computer study which is based upon the finite difference method. The program which was developed was modified and used to calculate redistribution data for boron and phosphorous which had been predeposited by ion implantation with range and straggle conditions. Data were generated for 111 oriented SOS films with redistribution in N2, dry O2 and steam ambients.

Trotter, J. D.↗
DOE OSTI2017

Using Dynamic Simulation to Evaluate Attemperator Operation in a Natural Gas Combined Cycle With Duct Burners in the Heat Recovery Steam Generator

A generic training simulator of a natural gas combined cycle was modified to match operations at a real plant. The objective was to use the simulator to analyze cycling operations of the plant. Initial operation of the simulator revealed the potential for saturation conditions in the final high pressure superheater as the attemperator tried to control temperature at the superheater outlet during gas turbine loading and unloading. Subsequent plant operational data confirmed simulation results. Multiple simulations were performed during loading and unloading of the gas turbine to determine operational strategies that prevented saturation and increased the approach to saturation temperature. The solutions included changes to the attemperator temperature control setpoints and strategic control of the steam turbine inlet pressure control valve.

20 FOSSIL-FUELED POWER PLANTS↗