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Evaluation of burst probability for tubes by Weibull distributions

Description: The investigations of candidate distributions that best describe the burst pressure failure probability characteristics of nuclear power steam generator tubes has been continued. To date it has been found that the Weibull distribution provides an acceptable fit for the available data from both the statistical and physical viewpoints. The reasons for the acceptability of the Weibull distribution are stated together with the results of tests for the suitability of fit. In exploring the acceptability of the Weibull distribution for the fitting, a graphical method to be called the ''density-gram'' is employed instead of the usual histogram. With this method a more sensible graphical observation on the empirical density may be made for cases where the available data is very limited. Based on these methods estimates of failure pressure are made for the left-tail probabilities.
Date: October 1, 1975
Creator: Kao, S.
Partner: UNT Libraries Government Documents Department

Flow-induced vibration in LMFBR steam generators: a state-of-the-art review

Description: This state-of-the-art review identifies and discusses existing methods of flow-induced vibration analysis applicable to steam generators, their limitations, and base-technology needs. Also included are discussions of five different LMFBR steam-generator configurations and important design considerations, failure experiences, possible flow-induced excitation mechanisms, vibration testing, and available methods of vibration analysis. The objectives are to aid LMFBR steam-generator designers in making the best possible evaluation of potential vibration in steam-generator internals, and to provide the basis for development of design guidelines to avoid detrimental flow-induced vibration. (auth)
Date: May 1, 1975
Creator: Shin, Y.S. & Wambsganss, M.W.
Partner: UNT Libraries Government Documents Department

Receiver System: Lessons Learned From Solar Two

Description: The Boeing Company fabricated the Solar Two receiver as a subcontractor for the Solar Two project. The receiver absorbed sunlight reflected from the heliostat field. A molten-nitrate-salt heat transfer fluid was pumped from a storage tank at grade level, heated from 290 to 565 C by the receiver mounted on top of a tower, then flowed back down into another storage tank. To make electricity, the hot salt was pumped through a steam generator to produce steam that powered a conventional Rankine steam turbine/generator. This evaluation identifies the most significant Solar Two receiver system lessons learned from the Mechanical Design, Instrumentation and Control, Panel Fabrication, Site Construction, Receiver System Operation, and Management from the perspective of the receiver designer/manufacturer. The lessons learned on the receiver system described here consist of two parts: the Problem and one or more identified Solutions. The appendix summarizes an inspection of the advanced receiver panel developed by Boeing that was installed and operated in the Solar Two receiver.
Date: March 1, 2002
Partner: UNT Libraries Government Documents Department

Development of the Radiation Stabilized Distributed Flux Burner, Phase II Final Report

Description: This report covers progress made during Phase 2 of a three-phase DOE-sponsored project to develop and demonstrate the Radiation Stabilized Distributed Flux burner (also referred to as the Radiation Stabilized Burner, or RSB) for use in industrial watertube boilers and process heaters. The goal of the DOE-sponsored work is to demonstrate an industrial boiler burner with NOx emissions below 9 ppm and CO emissions below 50 ppm (corrected to 3% stack oxygen). To be commercially successful, these very low levels of NOx and CO must be achievable without significantly affecting other measures of burner performance such as reliability, turndown, and thermal efficiency. Phase 1 of the project demonstrated that sub-9 ppm NOx emissions and sub-50 ppm CO emissions (corrected to 3% oxygen) could be achieved with the RSB in a 3 million Btu/Hr laboratory boiler using several methods of NOx reduction. The RSB was also tested in a 60 million Btu/hr steam generator used by Chevron for Thermally Enhanced Oil Recovery (TEOR). In the larger scale tests, fuel staging was demonstrated, with the RSB consistently achieving sub-20 ppm NOx and as low as 10 ppm NOx. Large-scale steam generator tests also demonstrated that flue gas recirculation (FGR) provided a more predictable and reliable method of achieving sub-9 ppm NOx levels. Based on the results of tests at San Francisco Thermal and Chevron, the near-term approach selected by Alzeta for achieving low NOx is to use FGR. This decision was based on a number of factors, with the most important being that FGR has proved to be an easier approach to transfer to different facilities and boiler designs. In addition, staging has proved difficult to implement in a way that allows good combustion and emissions performance in a fully modulating system. In Phase 3 of the project, the RSB will be ...
Date: June 1, 1997
Creator: Webb, A. & Sullivan, J.D.
Partner: UNT Libraries Government Documents Department

Compact Single-Stage Fuel Processor for PEM Fuel Cells. Final report

Description: Based on observations during the steam reforming of ethanol, the authors conclude that carbon was forming in the steam generator due to the thermal decomposition of ethanol. Since ethanol is being thermally decomposed, they were operating the steam generator at too high of a temperature. The thermal degradation of ethanol was confirmed by using a GC with a flame ionization detector. They observed trace amounts of additional hydrocarbons other than methane in the effluent which we assume maybe ethane and ethylene. We identified the operating conditions that allowed us to steam reform ethanol for an acceptable amount of time. These conditions were a steam temperature of 200 C and a wall temperature of 400 C at the center of the reactor. The calculated ratios of CO{sub 2}/CO indicate that we can lower the potential for carbon deposition from the Boudouard further by reducing the pressure.
Date: January 2000
Creator: Rhine, Wendell E. & Ye, Neng
Partner: UNT Libraries Government Documents Department


Description: Advanced ferritic/martensitic steels are being used extensively in fossil energy applications. New steels such as 2 1/4Cr-W-V (T23, T24), 3Cr-W-V, 9Cr-Mo-V (T91), 7Cr-W-V, 9Cr-W-V (T92 and T911), and 12Cr-W-V (T122, SAVE 12, and NF12) are examples of tubing being used in boilers and heat recovery steam generators (1). Other products for these new steels include piping, plates, and forgings. There is concern about the high-temperature performance of the advanced steels for several reasons. First, they exhibit a higher sensitivity to temperature than the 300 series stainless steels that they often replace. Second, they tend to be metallurgically unstable and undergo significant degradation at service temperatures in the creep range. Third, the experience base is limited in regard to duration. Fourth, they will be used for thick-section, high-pressure components that require high levels of integrity. To better understand the potential limitations of these steels, damage models are being developed that consider metallurgical factors as well as mechanical performance factors. Grade 91 steel was chosen as representative of these steels for evaluation of cumulative damage models since laboratory and service exposures of grade 91 exceed 100,000 hours.
Date: April 22, 2003
Creator: Swindeman, R.W.; Maziasz, P.J. & Swindeman, M.J.
Partner: UNT Libraries Government Documents Department

Analysis and testing of rupture of steam generator tubing with flaws.

Description: A high-temperature (300 C), high-pressure (18 MPa), and high-leak rate (1500 L/min) facility, and a room temperature, high-pressure (52 MPa) test facility were used to test flawed steam generator tubes. Single and multiple rectangular flaws were fabricated by electro-discharge machining on the outside surface of the tubes. This paper briefly reviews analytical methods for predicting ligament rupture and unstable burst of tubes with single and multiple rectangular flaws. Test data are presented to validate the failure models. The ligament rupture pressure of specimens with multiple flaws predicted by an equivalent rectangular crack method agree fairly well with measured data.
Date: February 8, 2001
Creator: Majumdar, S.; Kasza, K. S.; Park, J. Y. & Hanna, J. A.
Partner: UNT Libraries Government Documents Department

Development of Design Criteria for Fluid Induced Structural Vibrations in Steam Generators and Heat Exchangers

Description: Flow-induced vibration in heat exchangers has been a major cause of concern in the nuclear industry for several decades. Many incidents of failure of heat exchangers due to apparent flow-induced vibration have been reported through the USNRC incident reporting system. Almost all heat exchangers have to deal with this problem during their operation. The phenomenon has been studied since the 1970s and the database of experimental studies on flow-induced vibration is constantly updated with new findings and improved design criteria for heat exchangers.
Date: April 6, 2004
Creator: Catton, Uvan; Dhir, Vijay K.; Mitra, Deepanjan; Alquaddoomi, Omar & Adinolfi, Pierangelo
Partner: UNT Libraries Government Documents Department

Modeling a Helical-coil Steam Generator in RELAP5-3D for the Next Generation Nuclear Plant

Description: Options for the primary heat transport loop heat exchangers for the Next Generation Nuclear Plant are currently being evaluated. A helical-coil steam generator is one heat exchanger design under consideration. Safety is an integral part of the helical-coil steam generator evaluation. Transient analysis plays a key role in evaluation of the steam generators safety. Using RELAP5-3D to model the helical-coil steam generator, a loss of pressure in the primary side of the steam generator is simulated. This report details the development of the steam generator model, the loss of pressure transient, and the response of the steam generator primary and secondary systems to the loss of primary pressure. Back ground on High Temperature Gas-cooled reactors, steam generators, the Next Generation Nuclear Plant is provided to increase the readers understanding of the material presented.
Date: January 1, 2011
Creator: Hoffer, Nathan V.; Sabharwall, Piyush & Anderson, Nolan A.
Partner: UNT Libraries Government Documents Department

Experimental Evaluation of Tude Support Plate Crevice Chemistry

Description: A test methodology for measuring temperature, impedance, pH, and electrochemical potential distributions within a sludge-packed tube support plate crevice in a laboratory test is described. The method successfully showed that there were large concentration gradients between the tube and tube support plate sides of the crevice. The testing also showed that strong bases concentrated more effectively than strong acids, and that the crevice pH, when exposed to seawater-based solutions, increased with increasing superheat and decreasing bulk concentration. The large variations in the crevice chemistry observed under heat transfer were eliminated upon shutdown. These new test data suggest that it might be beneficial to evaluate the variation in the extent of stress corrosion cracking with tube support plate elevation found in some steam generators in light of local chemistry changes, as well as the variation in tubing temperature. Because of the large crevice chemistry gradients during boiling heat transfer and their subsequent homogenization upon test shutdown, the results suggest reassessing the use of hideout return measurements and tube deposit analyses in industry to infer the crevice chemistry under heat transfer conditions.
Date: May 8, 2001
Creator: Baum, Allen
Partner: UNT Libraries Government Documents Department

LOFT Monthly Progress Report for March 1980

Description: During March, evaluation of Test L3-2 continued and preparations began for the next tests, L6-5 and L3-7. Test L6-5 is the first test of the operational transient series and is a loss-of-feedwater incident. Test L3-7 is a small break, similar to L3-2, wherein the emergency coolant flow is adjusted to permit investigation of natural circulation modes. The controlling schedule items involve efforts to add instrumentation to the plant to better characterize system response. Specifically, a new hot-leg penetration will be installed and steam generator water level instrumentation will be improved. For March, costs to date agree very well with budget. Efforts are underway to develop a new baseline program for FY-80 and FY-81 based on a revised test plan and recent budget guidance for FY-81. These efforts are targeted for mid-year review in April.
Date: April 1, 1980
Creator: Kaufman, N. C.
Partner: UNT Libraries Government Documents Department