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Sample records for pressure blasts steam

  1. Decontamination of Steam Generator tube using Abrasive Blasting Technology

    International Nuclear Information System (INIS)

    Min, B. Y.; Kim, G. N.; Choi, W. K.; Lee, K. W.; Kim, D. H.; Kim, K. H.; Kim, B. T.

    2010-01-01

    As a part of a technology development of volume reduction and self disposal for large metal waste project, We at KAERI and our Sunkwang Atomic Energy Safety (KAES) subcontractor colleagues are demonstrating radioactively contaminated steam generator tube by abrasive blasting technology at Kori-1 NPP. A steam generator is a crucial component in a PWR (pressurized Water Reactor). It is the crossing between the primary, contaminated, circuit and the secondary waste-steam circuit. The heat from the primary reactor coolant loop is transferred to the secondary side in thousands of small tubes. Due to several problems in the material of those tube, like SCC (Stress Corrosion Cracking), insufficient control in water chemistry, which can be cause of tube leakage, more and more steam generators are replaced today. Only in Korea, already 2 of them are replaced and will be replaced in the near future. The retired 300 ton heavy Steam generator was stored at the storage waste building of Kori NPP site. The steam generator waste has a large volume, so that it is necessary to reduce its volume by decontamination. A waste reduction effect can be obtained through decontamination of the inner surface of a steam generator. Therefore, it is necessary to develop an optimum method for decontamination of the inner surface of bundle tubes. The dry abrasive blasting is a very interesting technology for the realization of three-dimensional microstructures in brittle materials like glass or silicon. Dry abrasive blasting is applicable to most surface materials except those that might be shattered by the abrasive. It is most effective on flat surface and because the abrasive is sprayed and can also applicable on 'hard to reach' areas such as inner tube ceilings or behind equipment. Abrasive decontamination techniques have been applied in several countries, including Belgium, the CIS, France, Germany, Japan, the UK and the USA

  2. Blast wave parameters at diminished ambient pressure

    Science.gov (United States)

    Silnikov, M. V.; Chernyshov, M. V.; Mikhaylin, A. I.

    2015-04-01

    Relation between blast wave parameters resulted from a condensed high explosive (HE) charge detonation and a surrounding gas (air) pressure has been studied. Blast wave pressure and impulse differences at compression and rarefaction phases, which traditionally determine damage explosive effect, has been analyzed. An initial pressure effect on a post-explosion quasi-static component of the blast load has been investigated. The analysis is based on empirical relations between blast parameters and non-dimensional similarity criteria. The results can be directly applied to flying vehicle (aircraft or spacecraft) blast safety analysis.

  3. X-ray diffractometry of steam cured ordinary Portland and blast-furnace-slag cements

    International Nuclear Information System (INIS)

    Camarini, G.; Djanikian, J.G.

    1994-01-01

    This work studies some aspects of the phases produced by hydration of ordinary and blast-furnace-slag cements, at normal conditions and steam cured (60 and 95 0 C), using an X-ray diffraction technique. The blast-furnace-slag cement was a mixture of 50% of ordinary Portland cement and 50% of blast-furnace-slag (separately grinding). After curing the X-ray diffraction reveals that, in relation to ordinary Portland cement, the main phases in blast-furnace-slag cement are hydrated silicates and aluminates, hydro garnet, etringitte and mono sulphate. After steam curing the hydration of blast-furnace-slag cement proceeds. This is a result of the slag activation by the curing temperature. (author). 8 refs., 3 figs., 1 tab

  4. Flame emission spectroscopy measurement of a steam blast and air blast burner

    Directory of Open Access Journals (Sweden)

    Jozsa Viktor

    2017-01-01

    Full Text Available Control and online monitoring of combustion have become critical to meet the increasingly strict pollutant emission standards. For such a purpose, optical sensing methods, like flame emission spectrometry, seem to be the most feasible technique. Spectrometry is capable to provide information about the local equivalence ratio inside the flame through the chemiluminescence intensity ratio measurement of various radicals. In the present study, a 15 kW atmospheric burner was analyzed utilizing standard diesel fuel. Its plain jet type atomizer was operated with both air and steam atomizing mediums. Up to now, injection of steam into the reaction zone has attracted less scientific attention contrary to its practical importance. Spatial plots of OH*, CH*, and C2* excited radicals were analyzed at 0.35, 0.7, and 1 bar atomization gauge pressures, utilizing both atomizing mediums. The C2* was found to decrease strongly with increasing steam addition. The OH*/CH* and OH*/C2* chemiluminescence intensity ratios along the axis showed a divergent behavior in all the analyzed cases. Nevertheless, CH*/C2* chemiluminescence intensity ratio decreased only slightly, showing low sensitivity to the position of the spectrometer. The findings may be directly applied in steady operating combustion systems, i. e., gas turbines, boilers, and furnaces.

  5. Experimental Study and Engineering Practice of Pressured Water Coupling Blasting

    Directory of Open Access Journals (Sweden)

    J. X. Yang

    2017-01-01

    Full Text Available Overburden strata movement in large space stope is the major reason that induces the appearance of strong mining pressure. Presplitting blasting for hard coal rocks is crucial for the prevention and control of strong pressure in stope. In this study, pressured water coupling blasting technique was proposed. The process and effect of blasting were analyzed by orthogonal test and field practice. Results showed that the presence of pressure-bearing water and explosive cartridges in the drill are the main influence factors of the blasting effect of cement test block. The high load-transmitting performance of pore water and energy accumulation in explosive cartridges were analyzed. Noxious substances produced during the blasting process were properly controlled because of the moistening, cooling, and diluting effect of pore water. Not only the goal of safe and static rock fragmentation by high-explosive detonation but also a combination of superdynamic blast loading and static loading effect of the pressured water was achieved. Then the practice of blasting control of hard coal rocks in Datong coal mine was analyzed to determine reasonable parameters of pressured water coupling blasting. A good presplitting blasting control effect was achieved for the hard coal rocks.

  6. HTGR power plant hot reheat steam pressure control system

    International Nuclear Information System (INIS)

    Braytenbah, A.S.; Jaegtnes, K.O.

    1975-01-01

    A control system for a high temperature gas cooled reactor (HTGR) power plant is disclosed wherein such plant includes a plurality of steam generators. Dual turbine-generators are connected to the common steam headers, a high pressure element of each turbine receiving steam from the main steam header, and an intermediate-low pressure element of each turbine receiving steam from the hot reheat header. Associated with each high pressure element is a bypass line connected between the main steam header and a cold reheat header, which is commonly connected to the high pressure element exhausts. A control system governs the flow of steam through the first and second bypass lines to provide for a desired minimum steam flow through the steam generator reheater sections at times when the total steam flow through the turbines is less than such minimum, and to regulate the hot reheat header steam pressure to improve control of the auxiliary steam turbines and thereby improve control of the reactor coolant gas flow, particularly following a turbine trip. (U.S.)

  7. PRESSURE-IMPULSE DIAGRAM OF MULTI-LAYERED ALUMINUM FOAM PANELS UNDER BLAST PRESSURE

    Directory of Open Access Journals (Sweden)

    CHANG-SU SHIM

    2013-06-01

    Full Text Available Anti-terror engineering has increasing demand in construction industry, but basis of design (BOD is normally not clear for designers. Hardening of structures has limitations when design loads are not defined. Sacrificial foam claddings are one of the most efficient methods to protect blast pressure. Aluminum foam can have designed yield strength according to relative density and mitigate the blast pressure below a target transmitted pressure. In this paper, multi-layered aluminum foam panels were proposed to enhance the pressure mitigation by increasing effective range of blast pressure. Through explicit finite element analyses, the performance of blast pressure mitigation by the multi-layered foams was evaluated. Pressure-impulse diagrams for the foam panels were developed from extensive analyses. Combination of low and high strength foams showed better applicability in wider range of blast pressure.

  8. Numerical Analysis on Transient of Steam-gas Pressurizer

    International Nuclear Information System (INIS)

    Kim, Jong-Won; Lee, Yeon-Gun; Park, Goon-Cherl

    2008-01-01

    In nuclear reactors, various pressurizers are adopted to satisfy their characteristics and uses. The additional active systems such as heater, pressurizer cooler, spray and insulator are essential for a steam or a gas pressurizer. With a steam-gas pressurizer, additional systems are not required due to the use of steam and non-condensable gas as pressure-buffering materials. The steam-gas pressurizer in integrated small reactors experiences very complicated thermal-hydraulic phenomena. To ensure the integrity of this pressurizer type, the analysis on the transient behavior of the steam-gas pressure is indispensable. For this purpose, the steam-gas pressurizer model is introduced to predict the accurate system pressure. The proposed model includes bulk flashing, rainout, inter-region heat and mass transfer and wall condensation with non-condensable gas. However, the ideal gas law is not applied because of significant interaction at high pressure between steam and non-condensable gas. The results obtained from this proposed model agree with those from pressurizer tests. (authors)

  9. Modeling steam pressure under martian lava flows

    Science.gov (United States)

    Dundas, Colin M.; Keszthelyi, Laszlo P.

    2013-01-01

    Rootless cones on Mars are a valuable indicator of past interactions between lava and water. However, the details of the lava–water interactions are not fully understood, limiting the ability to use these features to infer new information about past water on Mars. We have developed a model for the pressurization of a dry layer of porous regolith by melting and boiling ground ice in the shallow subsurface. This model builds on previous models of lava cooling and melting of subsurface ice. We find that for reasonable regolith properties and ice depths of decimeters, explosive pressures can be reached. However, the energy stored within such lags is insufficient to excavate thick flows unless they draw steam from a broader region than the local eruption site. These results indicate that lag pressurization can drive rootless cone formation under favorable circumstances, but in other instances molten fuel–coolant interactions are probably required. We use the model results to consider a range of scenarios for rootless cone formation in Athabasca Valles. Pressure buildup by melting and boiling ice under a desiccated lag is possible in some locations, consistent with the expected distribution of ice implanted from atmospheric water vapor. However, it is uncertain whether such ice has existed in the vicinity of Athabasca Valles in recent history. Plausible alternative sources include surface snow or an aqueous flood shortly before the emplacement of the lava flow.

  10. Characterization of a steam plasma jet at atmospheric pressure

    International Nuclear Information System (INIS)

    Ni Guohua; Zhao Peng; Cheng Cheng; Song Ye; Meng Yuedong; Toyoda, Hirotaka

    2012-01-01

    An atmospheric steam plasma jet generated by an original dc water plasma torch is investigated using electrical and spectroscopic techniques. Because it directly uses the water used for cooling electrodes as the plasma-forming gas, the water plasma torch has high thermal efficiency and a compact structure. The operational features of the water plasma torch and the generation of the steam plasma jet are analyzed based on the temporal evolution of voltage, current and steam pressure in the arc chamber. The influence of the output characteristics of the power source, the fluctuation of the arc and current intensity on the unsteadiness of the steam plasma jet is studied. The restrike mode is identified as the fluctuation characteristic of the steam arc, which contributes significantly to the instabilities of the steam plasma jet. In addition, the emission spectroscopic technique is employed to diagnose the steam plasma. The axial distributions of plasma parameters in the steam plasma jet, such as gas temperature, excitation temperature and electron number density, are determined by the diatomic molecule OH fitting method, Boltzmann slope method and H β Stark broadening, respectively. The steam plasma jet at atmospheric pressure is found to be close to the local thermodynamic equilibrium (LTE) state by comparing the measured electron density with the threshold value of electron density for the LTE state. Moreover, based on the assumption of LTE, the axial distributions of reactive species in the steam plasma jet are estimated, which indicates that the steam plasma has high chemical activity.

  11. Pressure effects on high temperature steam oxidation of Zircaloy-4

    International Nuclear Information System (INIS)

    Park, Kwangheon; Kim, Kwangpyo; Ryu, Taegeun

    2000-01-01

    The pressure effects on Zircaloy-4 (Zry-4) cladding in high temperature steam have been analyzed. A double layer autoclave was made for the high pressure, high temperature oxidation tests. The experimental test temperature range was 700 - 900 deg C, and pressures were 0.1 - 15 MPa. Steam partial pressure turns out to be an important one rather than total pressure. Steam pressure enhances the oxidation rate of Zry-4 exponentially. The enhancement depends on the temperature, and the maximum exists between 750 - 800 deg C. Pre-existing oxide layer decreases the enhancement about 40 - 60%. The acceleration of oxidation rate by high pressure team seems to be originated from the formation of cracks by abrupt transformation of tetragonal phase in oxide, where the un-stability of tetragonal phase comes from the reduction of surface energy by steam. (author)

  12. Ice drilling for blasting boreholes in deep seismic surveys (JARE-43 by steam type drilling system

    Directory of Open Access Journals (Sweden)

    Atsushi Watanabe

    2003-03-01

    Full Text Available A seismic exploration was accomplished in the austral summer of 2001-2002 by the 43rd Japanese Antarctic Research Expedition (JARE-43 along a profile oblique to that held by JARE-41 on the Mizuho Plateau, East Antarctica. We used a steam type drilling system to obtain seven blasting boreholes. We spent 7 to 8 hours to make an enough depth of the hole for one shot point. The holes were 35 to 40 cm in diameter and 23.5 to 28.7 m in depth. The average drilling speed was 3.25 m/hr.

  13. Steam plant for pressurized water reactors

    International Nuclear Information System (INIS)

    Anon.

    1983-01-01

    This book discusses the research and development organisations and users to highlight those aspects of the steam turbine and associated plant which are particularly related to the PWR system. The contents include: Characteristics of the steam system (including feed train, dump system and safety aspects); overall design aspects of high and half speed turbines; design aspects of the steam generator and seismic considerations; moisture separators and reheaters; feed pumps and their drives; water treatment; safety related valves; operational experience; availability and performance

  14. Pressure drop, steam content and turbulent cross exchange in water/steam flows

    International Nuclear Information System (INIS)

    Teichel, H.

    1978-01-01

    For describing the behaviour of two-phase flows of water and steam with the help of calculating patterns, a number of empirical correlations are required. - In this article, correlations for the friction pressure drop in water/steam flows are compared, as well as for the steam mass and the volumetric steam content with each other and with the test results on simple geometries. As the mutual effect between cooling chanels plays an important part at the longitudinal flow through bar bundles, the appertaining equations are evaluated, in addition. (orig.) 891 HP [de

  15. Pressurizer and steam-generator behavior under PWR transient conditions

    International Nuclear Information System (INIS)

    Wahba, A.B.; Berta, V.T.; Pointner, W.

    1983-01-01

    Experiments have been conducted in the Loss-of-Fluid Test (LOFT) pressurized water reactor (PWR), at the Idaho National Engineering Laboratory, in which transient phenomena arising from accident events with and without reactor scram were studied. The main purpose of the LOFT facility is to provide data for the development of computer codes for PWR transient analyses. Significant thermal-hydraulic differences have been observed between the measured and calculated results for those transients in which the pressurizer and steam generator strongly influence the dominant transient phenomena. Pressurizer and steam generator phenomena that occurred during four specific PWR transients in the LOFT facility are discussed. Two transients were accompanied by pressurizer inflow and a reduction of the heat transfer in the steam generator to a very small value. The other two transients were accompanied by pressurizer outflow while the steam generator behavior was controlled

  16. 46 CFR 54.01-10 - Steam-generating pressure vessels (modifies U-1(g)).

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 2 2010-10-01 2010-10-01 false Steam-generating pressure vessels (modifies U-1(g)). 54... ENGINEERING PRESSURE VESSELS General Requirements § 54.01-10 Steam-generating pressure vessels (modifies U-1(g)). (a) Pressure vessels in which steam is generated are classed as “Unfired Steam Boilers” except as...

  17. Steam chugging in pressure suppression containment

    International Nuclear Information System (INIS)

    Lee, C.K.B.; Chan, C.K.

    1978-01-01

    The condensation of steam flow in subcooled water was studied by injecting a quasi-steady stream of saturated steam into a pool water at different temperature. From the movies, it was observed that chugging occurred at a frequency on the order of 1 to 2 times a second. In between each chug over a period of approximately half a second, a few bubbles formed and collapsed at the exit of the downcomer. At a mass flow rate of approximately 5.02 Kg/m 2 sec., the chugging process is found to be strongly affected by the bubble formation. At pool temperatures below 50 0 C, the chugging process is dominated by internal chugging which is characterized by high water slug exit velocity, detached steam bubble and lhigh chugging level. Above 50 0 C, the external chugging mode is dominant. The external chugging mode is characterized by pancake bubble shape, low water slug exit velocity, and low chugging level. (author)

  18. Multi-layer casing of a steam turbine for high steam pressures and temperatures

    International Nuclear Information System (INIS)

    Remberg, A.

    1978-01-01

    In previous turbine casings there is no sealing provided between the inner layer and the outer layer, so that the steam pressure acts fully on the casing top and on the shaft seal housing situated there. To reduce the displacement which occurs there due to pressure differences in the various steam spaces, the normal inner casing is made with the shaft sealing housing in an inner layer, which cannot be divided in the axial direction. The inner layer can be inserted from the high pressure side into the unit outer casing. A horizontal section through the turbine in the attached drawing makes the construction and operation of the invention clear. (GL) [de

  19. Steam generator for pressurized-water reactors

    International Nuclear Information System (INIS)

    Michel, E.

    1971-01-01

    In the steam generator for a PWR the central fall space of a U-tube bundel heat exchanger is used as a preliminary cyclon separator. The steam escaping upwards, which is largely free of water, can flow through the residual heating surface, i.e. the U-tube turns. In this way substantial drying and less superheating by the heat still added becomes possible. In its upper part the central fall space for the water separated in the preliminary separator, enclosed by a cylindrical guide wall and the U-tube bundle, is provided with tangential inlet slots. Through these, the water-steam mixture steams out of the section of the vertical legs of the U-tube bundle into the fall space. Above the inlet slots the rising space is closed by means of a turn-round plate. At the lower end of the guide wall outlet, slots are provided for the water flowing downwards and radially outwards into the unfilled space. (DG/PB) [de

  20. Study on steam pressure characteristics in various types of nozzles

    Science.gov (United States)

    Firman; Anshar, Muhammad

    2018-03-01

    Steam Jet Refrigeration (SJR) is one of the most widely applied technologies in the industry. The SJR system was utilizes residual steam from the steam generator and then flowed through the nozzle to a tank that was containing liquid. The nozzle converts the pressure energy into kinetic energy. Thus, it can evaporate the liquid briefly and release it to the condenser. The chilled water, was produced from the condenser, can be used to cool the product through a heat transfer process. This research aims to study the characteristics of vapor pressure in different types of nozzles using a simulation. The Simulation was performed using ANSYS FLUENT software for nozzle types such as convergent, convrgent-parallel, and convergent-divergent. The results of this study was presented the visualization of pressure in nozzles and was been validated with experiment data.

  1. The steam pressure effect on high temperature corrosion of zircaloy-4

    International Nuclear Information System (INIS)

    Kim, K. P.; Park, G. H.

    1998-01-01

    To find the effect of pressure on the high temperature oxidation of zircaloy-4, an autoclave capable of measuring the degree of oxidation at high temperatures and high pressure was manufactured. The degree of high temperature oxidation of zircaloy-4 was measured at three different conditions, high pressure steam, high pressure Ar gas with small amount of steam, and 1 atm steam. All the measurements were done at 750 deg C. The oxide thickness is much thicker in high pressure steam, comparing to that in the 1 atm steam. And, the higher is the steam pressure, the thicker becomes the oxide. No effect was observed in the case of high pressure Ar containing small amount of steam. Many cracks exist on the surface of specimens oxidized at high pressure steam, which come from the enhanced tetragonal to monoclinic phase transformation due to high pressure steam. The enhanced oxidation seems to oxide cracking

  2. Effect of pressurized steam on AA1050 aluminium

    DEFF Research Database (Denmark)

    Jariyaboon, Manthana; Møller, Per; Ambat, Rajan

    2012-01-01

    Purpose - The purpose of this paper is to understand the effect of pressurized steam on surface changes, structures of intermetallic particles and corrosion behavior of AA1050 aluminium. Design/methodology/approach - Industrially pure aluminium (AA1050, 99.5 per cent) surfaces were exposed...... reactivities was observed due to the formation of the compact oxide layer. Originality/value - This paper reveals a detailed investigation of how pressurized steam can affect the corrosion behaviour of AA1050 aluminium and the structure of Fe-containing intermetallic particles....

  3. Steam generator for a pressurized-water coolant nuclear reactor

    International Nuclear Information System (INIS)

    Schroeder, H.J.; Berger, W.

    1975-01-01

    A description is given of a steam generator which has a vertical cylindrical housing having a steam output outlet, a horizontal tube sheet closing the lower end of this housing, and an inverted U-shaped tube bundle inside of the housing and having vertical inlet and outlet legs with their ends mounted in the tube sheet. Beneath the tube sheet there are inlet and outlet manifolds for the respective ends of the tube bundle so that pressurized-water coolant from a pressurized-water coolant nuclear reactor can be circulated through the tube bundle

  4. Muzzle Blast Pressure Loadings upon Aircraft Surfaces

    Science.gov (United States)

    1984-02-01

    0.02 m Aluminum plate. The plate is instrumented with a linear array of ten piezoelectric pressure transducers, Kistler Model 201B5. The plate is...maintained to insure compatibility of the various sets of 7. E. M. Schmidt, E. J. Gion, and D. D. Shear, "Acoustic Thermometric Measurements of...s -’-’»VA»’ ^I*T^^,*^^T^7.^ MVV-V-VV VV-V-V’-T-^ *7»T REFERENCES 7. E. M. Schmidt, E. J. Gion, and D. D. Shear, "Acoustic Thermometric

  5. Dependence of steam generator vibrations on feedwater pressure

    International Nuclear Information System (INIS)

    Sadilek, J.

    1989-01-01

    Vibration sensors are attached to the bottom of the steam generator jacket between the input and output primary circuit collectors. The effective vibration value is recorded daily. Several times higher vibrations were observed at irregular intervals; their causes were sought, and the relation between the steam generator vibrations measured at the bottom of its vessel and the feedwater pressure was established. The source of the vibrations was found to be in the feedwater tract of the steam generator. The feedwater tract is described and its hydraulic characteristics are given. Vibrations were measured on the S02 valve. It is concluded that vibrations can be eliminated by reducing the water pressure before the control valves and by replacing the control valves with ones with more suitable control characteristics. (E.J.). 3 figs., 1 tab., 3 refs

  6. Pressurized-water coolant nuclear reactor steam generator

    International Nuclear Information System (INIS)

    Mayer, H.; Schroder, H.J.

    1975-01-01

    A description is given of a pressurized-water coolant nuclear reactor steam generator having a vertical housing for the steam generating water and containing an upstanding heat exchanger to which the pressurized-water coolant passes and which is radially surrounded by a guide jacket supporting a water separator on its top. By thermosiphon action the steam generating water flows upward through and around the heat exchanger within the guide chamber to the latter's top from which it flows radially outwardly and downwardly through a down draft space formed between the outside of the jacket and the housing. The water separator discharges separated water downwardly. The housing has a feedwater inlet opening adjacent to the lower portion of the heat exchanger, providing preheating of the introduced feedwater. This preheated feedwater is conveyed by a duct upwardly to a location where it mixes with the water discharged from the water separator

  7. Detection of steam generator tube leaks in pressurized water reactors

    International Nuclear Information System (INIS)

    Roach, W.H.

    1985-01-01

    This report addresses the early detection of small steam generator tube leaks in pressurized water reactors. It discusses the third, and final, year's work on an NRC-funded project examining diagnostic instrumentation in water reactors. The first two years were broad in coverage, concentrating on anticipatory measurements for detection of potential problems in both pressurized- and boiling-water reactors, with recommendations for areas of further study. One of these areas, the early detection of small steam tube leaks in PWRs, formed the basis of study for the last year of the project. Four tasks are addressed in this study of the detection of steam tube leaks. (1) Determination of which physical parameters indicate the onset of steam generator tube leaks. (2) Establishing performance goals for diagnostic instruments which could be used for early detection of steam generator tube leaks. (3) Defining the diagnostic instrumentation and their location which satisfy Items 1 and 2 above. (4) Assessing the need for diagnostic data processing and display. Parameters are identified, performance goals established, and sensor types and locations are specified in the report, with emphasis on the use of existing instrumentation with a minimum of retrofitting. A simple algorithm is developed which yields the leak rate as a function of known or measurable quantities. The conclusion is that leak rates of less than one-tenth gram per second should be detectable with existing instrumentation. (orig./HP)

  8. Effects of blast wave to main steam piping under high energy line break condition by TNT model

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seung Hyun; Lee, Eung Seok; Chang, Yoon Suk [Kyung Hee University, Yongin (Korea, Republic of)

    2016-05-15

    The aim of this study is to examine effect of the blast wave according to pipe break position through FE (Finite Element) analyses. If HELB (High Energy Line Break) accident occurs in nuclear power plants, not only environmental effect such as release of radioactive material but also secondary structural defects should be considered. Sudden pipe rupture causes ejection of high temperature and pressure fluid, which acts as a blast wave around the break location. The blast wave caused by the HELB has a possibility to induce structural defects around the components such as safe-related injection pipes and other structures.

  9. Response of steam-water mixtures to pressure transients

    International Nuclear Information System (INIS)

    Hull, L.M.

    1985-01-01

    During the transition phase of a hypothetical core-disruptive accident in a liquid-metal fast breeder reactor, melting fuel-steel mixtures may begin to boil, resulting in a two-phase mixture of molten reactor fuel and steel vapor. Dispersal of this mixture by pressure transients may prevent recriticality of the fuel material. This paper describes the results of a series of experiments that investigated the response of two-phase mixtures to pressure transients. Simulant fluids (steam/water) were used in a transparent 10.2-cm-dia, 63.5-cm-long acrylic tube. The pressure transient was provided by releasing pressurized nitrogen from a supply tank. The data obtained are in the form of pressure-time records and high-speed movies. The varied parameters are initial void fraction (10% and 40%) and transient pressure magnitude (3.45 and 310 kPa)

  10. Characteristics of fluctuating pressure generated in BWR main steam lines

    International Nuclear Information System (INIS)

    Takahashi, Shiro; Okuyama, Keita; Tamura, Akinori

    2009-01-01

    The BWR-3 steam dryer in the Quad Cities Unit 2 Nuclear Power Plant was damaged by high cycle fatigue due to acoustic-induced vibration. The dryer failure was as attributed to flow-induced acoustic resonance at the stub pipes of safety relief valves (SRVs) in the main steam lines (MSLs). The acoustic resonance was considered to be generated by interaction between the sound field and an unstable shear layer across the closed side branches with SRV stub pipes. We have started a research program on BWR dryers to develop their loading evaluation methods. Moreover, it has been necessary to evaluate the dryer integrity of BWR-5 plants which are the main type of BWR in Japan. In the present study, we used 1/10-scale BWR tests and analyses to investigate the flow-induced acoustic resonance and acoustic characteristics in MSLs. The test apparatus consisted of a steam dryer, a steam dome and 4 MSLs with 20 SRV stub pipes. A finite element method (FEM) was applied for the calculation of three-dimensional wave equations in acoustic analysis. We demonstrated that remarkable fluctuating pressures occurred in high and low frequency regions. High frequency fluctuating pressures was generated by the flow-induced acoustic resonance in the SRV stub pipes. Low frequency fluctuating pressure was generated in an MSL with the dead leg. The frequency of the latter almost coincided with the natural frequency of the MSL with the dead leg. The amplitude of the fluctuating pressures in the multiple stub pipes became more intense because of interaction between them compared with that in the single stub pipe. Acoustic analysis results showed that the multiple stub pipes caused several natural frequencies in the vicinity of the natural frequency of the single stub pipe and several modes of the standing wave in the MSLs. (author)

  11. Detection of steam generator tube leaks in pressurized water reactors

    International Nuclear Information System (INIS)

    Roach, W.H.

    1984-11-01

    This report addresses the early detection of small steam generator tube leaks in pressurized water reactors. It identifies physical parameters, establishes instrumentation performance goals, and specifies sensor types and locations. It presents a simple algorithm that yields the leak rate as a function of known or measurable quantities. Leak rates of less than one-tenth gram per second should be detectable with existing instrumentation

  12. Research on axial total pressure distributions of sonic steam jet in subcooled water

    International Nuclear Information System (INIS)

    Wu Xinzhuang; Li Wenjun; Yan Junjie

    2012-01-01

    The axial total pressure distributions of sonic steam jet in subcooled water were experimentally investigated for three different nozzle diameters (6.0 mm, 8.0 mm and 10.0 mm). The inlet steam pressure, and pool subcooling subcooled water temperature were in the range of 0.2-0.6 MPa and 420-860 ℃, respectively. The effect of steam pressure, subcooling water temperature and nozzle size on the axial pressure distributions were obtained, and also the characteristics of the maximum pressure and its position were studied. The results indicated that the characteristics of the maximum pressure were influenced by the nozzle size for low steam pressure, but the influence could be ignored for high steam pressure. Moreover, a correlation was given to correlate the position of the maximum pressure based on steam pressure and subcooling water temperature, and the discrepancies of predictions and experiments are within ±15%. (authors)

  13. Disintegration of liquid metals by low pressure water blasting

    International Nuclear Information System (INIS)

    Heshmatpour, B.; Copeland, G.L.

    1981-01-01

    The feasibility of disintegrating metals by a low cost system and subsequently incorporating them into grout mixtures has been demonstrated. A low pressure water blasting technique consisting of multiple nozzles and a converging-line jet stream was developed to disintegrate liquid metals and produce coarse metal powder and shot. Molten iron resulted in spherical shot, while copper, aluminum, and tin produced irregular shaped particles. The particle size was between 0.05 and 3 mm (0.002 and 0.1 in.), and about half the particles were smaller than 1 mm (0.04 in.) in all cases. The water consumption was rather low, while the production rate was relatively high. The method proved to be simple and reliable. The coarse metal powders were suspendable in grout fluids, indicating that they are probably disposable by the shale hydrofracture technique

  14. The accelerated oxidation of zircaloy-4 at 700∼900 .deg. C in high pressure steam

    International Nuclear Information System (INIS)

    Kim, K. P.; Park, K. H.

    1999-01-01

    To find the effect of pressure on the high temperature oxidation of zircaloy-4, an autoclave capable of measuring the degree of oxidation at high temperatures and high pressure was manufactured. The specimens used in experiments are commercially available Zircaloy-4 used in Kori nuclear power plants. All the measurements were done at 700∼900 .deg. C in steam. Pressure effects were noticed. The oxide thickness was much thicker in high pressure steam, comparing to that in the 1 atm steam. And, the higher is the steam pressure, the thicker becomes the oxide. The enhancement of oxide growth rate at 700∼900 .deg. C in high pressure steam is approximately propotion to the power of 1.0∼1.6 of the ratio of experimental steam pressure to critical steam pressure. There is a critical steam pressure above that the oxidation rate enhances. The critical steam pressure was measured as 30∼40 bar. The enhanced oxidation seems from the oxide cracking due to the tetragonal to monoclinic phase transformation at high pressure steam

  15. A study on impulsive sound attenuation for a high-pressure blast flow field

    International Nuclear Information System (INIS)

    Kang, Kuk Jeong; Ko, Sung Ho; Lee, Dong Soo

    2008-01-01

    The present work addresses a numerical study on impulsive sound attenuation for a complex high-pressure blast flow field; these characteristics are generated by a supersonic propellant gas flow through a shock tube into an ambient environment. A numerical solver for analyzing the high pressure blast flow field is developed in this study. From numerical simulations, wave dynamic processes (which include a first precursor shock wave, a second main propellant shock wave, and interactions in the muzzle blasts) are simulated and discussed. The pressure variation of the blast flow field is analyzed to evaluate the effect of a silencer. A live firing test is also performed to evaluate four different silencers. The results of this study will be helpful in understanding blast wave and in designing silencers

  16. A study on impulsive sound attenuation for a high-pressure blast flow field

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Kuk Jeong [Agency for Defence Development, Daejeon (Korea, Republic of); Ko, Sung Ho; Lee, Dong Soo [Chungnam National University, Daejeon (Korea, Republic of)

    2008-01-15

    The present work addresses a numerical study on impulsive sound attenuation for a complex high-pressure blast flow field; these characteristics are generated by a supersonic propellant gas flow through a shock tube into an ambient environment. A numerical solver for analyzing the high pressure blast flow field is developed in this study. From numerical simulations, wave dynamic processes (which include a first precursor shock wave, a second main propellant shock wave, and interactions in the muzzle blasts) are simulated and discussed. The pressure variation of the blast flow field is analyzed to evaluate the effect of a silencer. A live firing test is also performed to evaluate four different silencers. The results of this study will be helpful in understanding blast wave and in designing silencers

  17. Seismic-safe conditions of blasting near pressure pipe-lines during power installation construction

    International Nuclear Information System (INIS)

    Smolij, N.I.; Nikitin, A.S.

    1980-01-01

    Seismic-safe conditions for performing drill-blasting operations in the vicinity of underground gas pipelines when constructing thermal- or nuclear power plants are discussed. It is shown that, for the determination of seismic-safe parameters, of drill-blasting operations, the maximum permissible level of seismic loads should be specified taking into account the mechanical properties of the pipeline.metal, structural parameters of the gas pipeline and the pressure of the medium transported. Besides, the seismic effect of the blast should be considered with regard to particular conditions of blasting and rock properties. The equations and diagrams used in the calculation are given

  18. Nb effect on Zr-alloy oxidation under high pressure steam at high temperatures

    International Nuclear Information System (INIS)

    Park, Kwangheon; Yang, Sungwoo; Kim, Kyutae

    2005-01-01

    The high-pressure steam effects on the oxidation of Zircaloy-4 (Zry-4) and Zirlo (Zry-1%Nb) claddings at high temperature have been analyzed. Test temperature range was 700-900degC, and pressures were 1-150 bars. High pressure-steam enhances oxidation of Zry-4, and the dependency of enhancement looks exponential to steam pressure. The origin of the oxidation enhancement turned out to be the formation of cracks in oxide. The loss of tetragonal phase by high-pressure steam seems related to the crack formation. Addition of Nb as an alloying element to Zr alloy reduces significantly the steam pressure effects on oxidation. The higher compressive stresses and the smaller fraction of tetragonal oxides in Zry-1%Nb seem to be the diminished effect of high-pressure steam on oxidation. (author)

  19. Development of main steam safety valve set pressure evaluating system

    International Nuclear Information System (INIS)

    Oketani, Koichiro; Manabe, Yoshihisa.

    1991-01-01

    A main steam safety valve set pressure test is conducted for all valves during every refueling outage in Japan's PWRs. Almost all operations of the test are manually conducted by a skilled worker. In order to obtain further reliability and reduce the test time, an automatic test system using a personnel computer has been developed in accordance with system concept. Quality assurance was investigated to fix system specifications. The prototype of the system was manufactured to confirm the system reliability. The results revealed that this system had high accuracy measurement and no adverse influence on the safety valve. This system was concluded to be applicable for actual use. (author)

  20. Twin header bore welded steam generator for pressurized water reactors

    International Nuclear Information System (INIS)

    Davies, R.J.; Hirst, B.

    1979-01-01

    A description is given of a pressurized water reactor (PWR) steam generator concept, several examples of which have been in service for up to fourteen years. Details are given of the highly successful service record of this equipment and the features which have been incorporated to minimize corrosion and deposition pockets. The design employs a vertical U tube bundle carried off two horizontal headers to which the tubes are welded by the Foster Wheeler Power Products (FWPP) bore welding process. The factors to be considered in uprating the design to meet the current operating conditions for a 1000 MW unit are discussed. (author)

  1. SURGTANK, Steam Pressure, Saturation Temperature or Reactor Surge Tank

    International Nuclear Information System (INIS)

    Gorman, D.J.; Gupta, R.K.

    2001-01-01

    1 - Description of problem or function: SURGTANK generates the steam pressure, saturation temperature, and ambient temperature history for a nuclear reactor steam surge tank (pressurizer) in a state of thermodynamic equilibrium subjected to a liquid insurge described by a specified time history of liquid levels. It is capable also of providing the pressure and saturation temperature history, starting from thermodynamic equilibrium conditions, for the same tank subjected to an out-surge described by a time history of liquid levels. Both operations are available for light- or heavy- water nuclear reactor systems. The tank is assumed to have perfect thermal insulation on its outer wall surfaces. 2 - Method of solution: Surge tank geometry and initial liquid level and saturation pressure are provided as input for the out-surge problem, along with the prescribed time-sequence level history. SURGTANK assumes a reduced pressure for the end of the first change in liquid level and determines the associated change of entropy for the closed system. The assumed pressure is adjusted and the associated change in entropy recalculated until a pressure is attained for which no change occurs. This pressure is recorded and used as the beginning pressure for the next level increment. The system is then re-defined to exclude the small amount of liquid which has left the tank, and a solution for the pressure at the end of the second level increment is obtained. The procedure is terminated when the pressure at the end of the final increment has been determined. Surge tank geometry, thermal conductivity, specific heat, and density of tank walls, initial liquid level, and saturation pressure are provided as input for the insurge problem, along with the prescribed time-sequence level history. SURGTANK assumes a slightly in- creased pressure for the end of the first level, the inner tank sur- face is assumed to follow saturation temperature, linearly with time, throughout the interval, and

  2. Superheated steam annealing of pressurized water reactor vessel

    International Nuclear Information System (INIS)

    Porowski, J.S.

    1993-01-01

    Thermal annealing of an embrittled Reactor Pressure Shell is the only recognized means for recovering material properties lost due to long-term exposure of the reactor walls to radiation. Reduced toughness of the material during operation is a major concern in evaluations of structural integrity of older reactors. Extensive studies performed within programs related to life extension of nuclear plants have confirmed that the thermal treatment of 850 deg. F for 168 hours on irradiated material essentially recovers material properties lost due to neutron exposure. Dry and wet annealing methods have been considered. Wet annealing involves operating the reactor at near design temperatures and pressures. Since the temperature of wet annealing must be limited to vessel design temperature of 650 deg. F, only partial recovery of the lost properties is achieved. Thus dry annealing was selected as an alternative for future development and industrial implementation to extend the safe life of reactors. Dry thermal annealing consists of heating portions of the reactor vessel at a specific temperature for a given period of time using a high temperature heat source. The use of spent fuel assemblies, induction heating and resistance heating elements as well as the circulation of heated fluid were investigated as potential candidate methods. To date the use of resistance heating elements which are lowered into a dry empty reactor was considered to be the preferred method. In-depth research in the United States and practical applications of such a method in Russia have confirmed feasibility of the method. The method of using circulating superheated steam to anneal the vessel at 850 deg. F without complete removal of the reactor internals is described herein. After removing the reactor head and fuel, the core barrel along with the upper and lower core in PWRs is lifted to open an annular space between the reactor shell flange and the core barrel flange. The thermal shield can remain

  3. An ultra-fast fiber optic pressure sensor for blast event measurements

    International Nuclear Information System (INIS)

    Wu, Nan; Tian, Ye; Wang, Xingwei; Zou, Xiaotian; Fitek, John; Maffeo, Michael; Niezrecki, Christopher; Chen, Julie

    2012-01-01

    Soldiers who are exposed to explosions are at risk of suffering traumatic brain injury (TBI). Since the causal relationship between a blast and TBI is poorly understood, it is critical to have sensors that can accurately quantify the blast dynamics and resulting wave propagation through a helmet and skull that are imparted onto and inside the brain. To help quantify the cause of TBI, it is important to record transient pressure data during a blast event. However, very few sensors feature the capabilities of tracking the dynamic pressure transients due to the rapid change of the pressure during blast events, while not interfering with the physical material layers or wave propagation. In order to measure the pressure transients efficiently, a pressure sensor should have a high resonant frequency and a high spatial resolution. This paper describes an ultra-fast fiber optic pressure sensor based on the Fabry–Perot principle for the application of measuring the rapid pressure changes in a blast event. A shock tube experiment performed in US Army Natick Soldier Research, Development and Engineering Center has demonstrated that the resonant frequency of the sensor is 4.12 MHz, which is relatively close to the designed theoretical value of 4.113 MHz. Moreover, the experiment illustrated that the sensor has a rise time of 120 ns, which demonstrates that the sensor is capable of observing the dynamics of the pressure transient during a blast event. (paper)

  4. Correction of Pressure Drop in Steam and Water System in Performance Test of Boiler

    Science.gov (United States)

    Liu, Jinglong; Zhao, Xianqiao; Hou, Fanjun; Wu, Xiaowu; Wang, Feng; Hu, Zhihong; Yang, Xinsen

    2018-01-01

    Steam and water pressure drop is one of the most important characteristics in the boiler performance test. As the measuring points are not in the guaranteed position and the test condition fluctuation exsits, the pressure drop test of steam and water system has the deviation of measuring point position and the deviation of test running parameter. In order to get accurate pressure drop of steam and water system, the corresponding correction should be carried out. This paper introduces the correction method of steam and water pressure drop in boiler performance test.

  5. Design of a Simple Blast Pressure Gauge Based on a Heterodyne Velocimetry Measuring Technique

    Science.gov (United States)

    2016-08-01

    intensity of the blast being measured. For relatively low-pressure fields, such as that generated by release of compressed air from a standard shop ...unlimited. 13 4. References 1. Walter PL. Air-blast and the science of dynamic pressure measurements. Depew (NY): PCB Piezotronics; Fort Worth (TX...ALEGRA: An arbitrary Lagrangian-Eulerian multimaterial, multiphysics code. 46th AIAA Aerospace Sciences Meeting and Exhibit; 2008 Jan 7–10; Reno (NV

  6. Attenuation of blast pressure behind ballistic protective vests.

    Science.gov (United States)

    Wood, Garrett W; Panzer, Matthew B; Shridharani, Jay K; Matthews, Kyle A; Capehart, Bruce P; Myers, Barry S; Bass, Cameron R

    2013-02-01

    Clinical studies increasingly report brain injury and not pulmonary injury following blast exposures, despite the increased frequency of exposure to explosive devices. The goal of this study was to determine the effect of personal body armour use on the potential for primary blast injury and to determine the risk of brain and pulmonary injury following a blast and its impact on the clinical care of patients with a history of blast exposure. A shock tube was used to generate blast overpressures on soft ballistic protective vests (NIJ Level-2) and hard protective vests (NIJ Level-4) while overpressure was recorded behind the vest. Both types of vest were found to significantly decrease pulmonary injury risk following a blast for a wide range of conditions. At the highest tested blast overpressure, the soft vest decreased the behind armour overpressure by a factor of 14.2, and the hard vest decreased behind armour overpressure by a factor of 56.8. Addition of body armour increased the 50th percentile pulmonary death tolerance of both vests to higher levels than the 50th percentile for brain injury. These results suggest that ballistic protective body armour vests, especially hard body armour plates, provide substantial chest protection in primary blasts and explain the increased frequency of head injuries, without the presence of pulmonary injuries, in protected subjects reporting a history of blast exposure. These results suggest increased clinical suspicion for mild to severe brain injury is warranted in persons wearing body armour exposed to a blast with or without pulmonary injury.

  7. Large scale steam flow test: Pressure drop data and calculated pressure loss coefficients

    International Nuclear Information System (INIS)

    Meadows, J.B.; Spears, J.R.; Feder, A.R.; Moore, B.P.; Young, C.E.

    1993-12-01

    This report presents the result of large scale steam flow testing, 3 million to 7 million lbs/hr., conducted at approximate steam qualities of 25, 45, 70 and 100 percent (dry, saturated). It is concluded from the test data that reasonable estimates of piping component pressure loss coefficients for single phase flow in complex piping geometries can be calculated using available engineering literature. This includes the effects of nearby upstream and downstream components, compressibility, and internal obstructions, such as splitters, and ladder rungs on individual piping components. Despite expected uncertainties in the data resulting from the complexity of the piping geometry and two-phase flow, the test data support the conclusion that the predicted dry steam K-factors are accurate and provide useful insight into the effect of entrained liquid on the flow resistance. The K-factors calculated from the wet steam test data were compared to two-phase K-factors based on the Martinelli-Nelson pressure drop correlations. This comparison supports the concept of a two-phase multiplier for estimating the resistance of piping with liquid entrained into the flow. The test data in general appears to be reasonably consistent with the shape of a curve based on the Martinelli-Nelson correlation over the tested range of steam quality

  8. Development of Steam Pressure Control Logic for SMART

    International Nuclear Information System (INIS)

    Sneed, Todd; Godbole, Sadashiva; Streaath, Drew

    2011-01-01

    The objective of this work is to develop the Steam Pressure Control Logic(SPCL) in order to satisfy the performance requirements of SMART plant control systems. The final conceptual SPCL described in this report has been developed after studying the SMART system together with the secondary system, patents of control systems for fossil and unclear power plants featuring OTSG's, and familiarity and experience with power plant control systems. The logic represents a combination of good features from various control concepts to make SPCL effective in controlling relevant SMART secondary system parameter. The SPCL includes some new features, such as use of FCV position setpoint to control pump speed, relaxation of the measure RC hot-leg temperature to accelerate load-following response, and use of steam flow as feedwater flow setpoint. This report describes performance evaluation results for the SMART SPCL MMS model. The evaluation is based on three transients under three control modes. Three transients include load-following ramp at normal ramp rate to reduce load and then to increase load, turbin trip, and loss of external electrical load resulting in only house load as the remaining load. The three control modes include coordinated, turbin-following, and feedwater-following. It seems that any of the three control modes is viable for SMART control

  9. Imaging optical probe for pressurized steam-water environment

    International Nuclear Information System (INIS)

    Donaldson, M.R.; Pulfrey, R.E.

    1979-01-01

    An air-cooled imaging optical probe, with an outside diameter of 25.4 mm, has been developed to provide high resolution viewing of flow regimes in a steam-water environment at 343 0 C and 15.2 MPa. The design study considered a 3-m length probe. A 0.3-m length probe prototype was fabricated and tested. The optical probe consists of a 3.5-mm diameter optics train surrounded by two coaxial coolant flow channels and two coaxial insulating dead air spaces. With air flowing through the probe at 5.7 g/s, thermal analysis shows that no part of the optics train will exceed 93 0 C when a 3-m length probe is immersed in a 343 0 C environment. Computer stress analysis plus actual tests show that the probe can operate successfully with conservative safety factors. The imaging optical probe was tested five times in the design environment at the semiscale facility at the INEL. Two-phase flow regimes in the high temperature, high pressure, steam-water blowdown and reflood experiments were recorded on video tape for the first time with the imaging optical probe

  10. A single-stage high pressure steam injector for next generation reactors: test results and analysis

    International Nuclear Information System (INIS)

    Cattadori, G.; Galbiati, L.; Mazzocchi, L.; Vanini, P.

    1995-01-01

    Steam injectors can be used in advanced light water reactors (ALWRs) for high pressure makeup water supply; this solution seems to be very attractive because of the ''passive'' features of steam injectors, that would take advantage of the available energy from primary steam without the introduction of any rotating machinery. The reference application considered in this work is a high pressure safety injection system for a BWR; a water flow rate of about 60 kg/s to be delivered against primary pressures covering a quite wide range up to 9 MPa is required. Nevertheless, steam driven water injectors with similar characteristics could be used to satisfy the high pressure core coolant makeup requirements of next generation PWRs. With regard to BWR application, an instrumented steam injector prototype with a flow rate scaling factor of about 1:6 has been built and tested. The tested steam injector operates at a constant inlet water pressure (about 0.2 MPa) and inlet water temperature ranging from 15 to 37 o C, with steam pressure ranging from 2.5 to 8.7 MPa, always fulfilling the discharge pressure target (10% higher than steam pressure). To achieve these results an original double-overflow flow rate-control/startup system has been developed. (Author)

  11. Flow Instabilities and Main Steam Line Vibrations in a Pressurized Water Reactor

    International Nuclear Information System (INIS)

    Henriksson, Mats; Westin, Johan; Granhall, Tord; Andersson, Lars; Bjerke, Lars-Erik

    2002-01-01

    Severe vibrational problems occurred in the main steam system of a PWR nuclear power plant, about 18 months after a steam generator replacement had been carried out. The magnitude of the vibrations reached levels at which the operators had to reduce power in order to stay within the operating limits imposed by the nuclear inspectorate. To solve the problem the following analyses methods were employed: - Testing the influence on vibration level from different modes of plant operation; - Analyses of plant measurement data; - Calculations of: hydraulic behaviour of the system, structural dynamic behaviour of the system, flow at the steam generator outlet. Scale model testing of the steam generator outlet region. Hydraulic flow disturbances in the main steam system were measured using pressure and strain gauges, which made it possible to track individual pressure pulses propagating through the main steam system. Analyses showed that the pressure pulses causing the vibration originated from the vicinity of the steam generator outlet. By using computer codes for network fluid flow analyses the pressure pulses found in the measurement traces could be generated in calculations. Careful studies of the flow at the steam generator outlet region, using model testing in a 1:3 scale model as well as transient 3D CFD calculations, gave clear indications that flow separation occurred at the steam generator outlet nozzle and at the first bend. Finally, by substituting the outlet nozzle for a different design with a multi-port nozzle, the steam line vibration problem has been solved. (authors)

  12. A single point of pressure approach as input for injury models with respect to complex blast loading conditions

    NARCIS (Netherlands)

    Teland, J.A.; Doormaal, J.C.A.M. van; Horst, M.J. van der; Svinsås, E.

    2010-01-01

    Blast injury models, like Axelsson and Stuhmiller, require four pressure signals as input. Those pressure signals must be acquired by a Blast Test Device (BTD) that has four pressure transducers placed in a horizontal plane at intervals of 90 degrees. This can be either in a physical test setup or

  13. A demonstration experiment of steam-driven, high-pressure melt ejection

    International Nuclear Information System (INIS)

    Allen, M.D.; Pitch, M.; Nichols, R.T.

    1990-08-01

    A steam blowdown test was performed at the Surtsey Direct Heating Test Facility to test the steam supply system and burst diaphragm arrangement that will be used in subsequent Surtsey Direct Containment Heating (DCH) experiments. Following successful completion of the steam blowdown test, the HIPS-10S (High-Pressure Melt Streaming) experiment was conducted to demonstrate that the technology to perform steam-driven, high-pressure melt ejection (HPME) experiments has been successfully developed. In addition, the HIPS-10S experiment was used to assess techniques and instrumentation design to create the proper timing of events in HPME experiments. This document discusses the results of this test

  14. Mitigation of steam generator tube rupture in a pressurized water reactor with passive safety systems

    Science.gov (United States)

    McDermott, Daniel J.; Schrader, Kenneth J.; Schulz, Terry L.

    1994-01-01

    The effects of steam generator tube ruptures in a pressurized water reactor are mitigated by reducing the pressure in the primary loop by diverting reactor coolant through the heat exchanger of a passive heat removal system immersed in the in containment refueling water storage tank in response to a high feed water level in the steam generator. Reactor coolant inventory is maintained by also in response to high steam generator level introducing coolant into the primary loop from core make-up tanks at the pressure in the reactor coolant system pressurizer. The high steam generator level is also used to isolate the start-up feed water system and the chemical and volume control system to prevent flooding into the steam header. 2 figures.

  15. Development of a higher capacity, lower pressure drop steam/water separator with reduced primary-to-secondary spacing

    International Nuclear Information System (INIS)

    Pruster, W.P.; Kidwell, J.H.; Eaton, A.M.; Wall, J.R.

    1985-01-01

    The goal of this development effort was to double the steam flow capacity of an existing module steam/water separator design without significantly increasing the pressure drop while simultaneously minimizing the vertical distance (spacing) between the primary and secondary separation stages. The development work included extensive air/water and steam/water testing. The steam/water tests were performed at a common pressure of 300 psia (2.1 MPa) with comparable water and steam flows

  16. Imitative modeling automatic system Control of steam pressure in the main steam collector with the influence on the main Servomotor steam turbine

    Science.gov (United States)

    Andriushin, A. V.; Zverkov, V. P.; Kuzishchin, V. F.; Ryzhkov, O. S.; Sabanin, V. R.

    2017-11-01

    The research and setting results of steam pressure in the main steam collector “Do itself” automatic control system (ACS) with high-speed feedback on steam pressure in the turbine regulating stage are presented. The ACS setup is performed on the simulation model of the controlled object developed for this purpose with load-dependent static and dynamic characteristics and a non-linear control algorithm with pulse control of the turbine main servomotor. A method for tuning nonlinear ACS with a numerical algorithm for multiparametric optimization and a procedure for separate dynamic adjustment of control devices in a two-loop ACS are proposed and implemented. It is shown that the nonlinear ACS adjusted with the proposed method with the regulators constant parameters ensures reliable and high-quality operation without the occurrence of oscillations in the transient processes the operating range of the turbine loads.

  17. Tube micro-fouling, boiling and steam pressure after chemical cleaning

    International Nuclear Information System (INIS)

    Hu, M.H.

    1998-01-01

    This paper presents steam pressure trends after chemical cleaning of steam generator tubes at four plants. The paper also presents tube fouling factor that serves as an objective parameter to assess tubing boiling conditions for understanding the steam pressure trend. Available water chemistry data helps substantiate the concept of tube micro-fouling, its effect on tubing boiling, and its impact on steam pressure. All four plants experienced a first mode of decreasing steam pressure in the post-cleaning operation. After 3 to 4 months of operation, the decreasing trend stopped for three plants and then restored to a pre-cleaning value or better. The fourth plant is soil in decreasing trend after 12 months of operation. Dissolved chemicals, such as silica, titanium can precipitate on tube surface. The precipitate micro-fouling can deactivate or eliminate boiling nucleation sites. Therefore, the first phase of the post-cleaning operation suffered a decrease in steam pressure or an increase in fouling factor. It appears that micro fouling by magnetite deposit can activate or create more bubble nucleation sites. Therefore, the magnetite deposit micro-fouling results in a decrease in fouling factor, and a recovery in steam pressure. Fully understanding the boiling characteristics of the tubing at brand new, fouled and cleaned conditions requires further study of tubing surface conditions. Such study should include boiling heat transfer tests and scanning electronic microscope examination. (author)

  18. Study on the behavior of moisture droplets in low pressure steam turbines

    International Nuclear Information System (INIS)

    Kimura, Y.; Kuramoto, Y.; Yoshida, K.; Etsu, M.

    1978-01-01

    Low pressure stages of fossil turbines and almost all stages of nuclear and geothermal turbines operate on wet steam. Turbine operating on wet steam have the following two disadvantages: decrease of efficiency and erosion of blades. Decrease of efficiency results from an increase in profile loss caused by water films on the blade surface; loss of steam energy in breaking up the films and accelerating moisture droplets; undercooling and condensation shocks associated with it; velocity difference between water and steam phases and consequent decelerating action of moisture droplets in the rotating blades, etc. Impingement of moisture droplets on the rotating blades also causes quick erosion of the blades. In this paper, the behavior of moisture droplets in wet steam flow is described and the correlation between their behavior and the abovementioned two disadvantages of turbines operating on wet steam is clarified. (author)

  19. High pressure oxidation of sponge-Zr in steam/hydrogen mixtures

    International Nuclear Information System (INIS)

    Kim, Y.S.

    1997-01-01

    A thermogravimetric apparatus for operation in 1 and 70 atm steam-hydrogen or steam-helium mixtures was used to investigate the oxidation kinetics of sponge-Zr containing 215 ppm Fe. Weight-gain rates, reflecting both oxygen and hydrogen uptake, were measured in the temperature range 350-400 C. The specimens consisted of thin sponge-Zr layers metallurgically bonded to a Zircaloy disk. The edges of the disk specimens were coated with a thin layer of pure gold to avoid the deleterious effect of corners. Following each experiment, the specimens were examined metallographically to reveal the morphology of the oxide and/or hydride formed. Two types of oxide, one black and uniform and the other white and nodular, were observed on sponge-Zr surfaces oxidized in steam environments at 70 atm. The oxidation rate when white-nodular oxide formed was a factor of two higher than that of black-uniform oxide at 400 C for steam contents above 1 mol%. The oxidation rate was independent of total pressure, the carrier gas (H 2 or He) and steam content above ∝1 mol%. The oxidation kinetics of sponge-Zr follows a linear law for maximum reaction times up to ∝6 days. The oxidation rate in steam-hydrogen mixtures at 70 atm total pressure decreases when the steam content approaches the steam-starved region (∝0.5 mol% steam at 400 C and ∝0.02 mol% steam at 350 C). Lower steam concentrations cause massive hydriding of the specimens. Even at steam concentrations above the critical value, direct hydrogen absorption from the gas was manifest by hydrogen pickup fractions greater than unity. (orig.)

  20. Numerical simulations of pressure fluctuations at branch piping in BWR main steam line

    International Nuclear Information System (INIS)

    Morita, Ryo; Inada, Fumio; Yoshikawa, Kazuhiro; Takahashi, Shiro

    2009-01-01

    The power uprating of a nuclear power plant may increase/accelerate degradation phenomena such as flow-induced vibration and wall thinking. A steam dryer was damaged by a high cycle fatigue due to an acoustic-induced vibration at the branch piping of safety relief valves (SRVs) in main steam lines. In this study, we conducted the numerical simulations of steam/air flow around a simplified branch piping to clarify the basic characteristics of resonance. LES simulations were conducted in ordinary pressure/temperature air and steam under BWR plant conditions. In both cases, the excitation of the pressure fluctuations at the branch was observed under some inlet velocity conditions. These fluctuations and inlet conditions were normalized and the obtained results were compared. The normalized results showed that the range and maximum amplitude of pressure fluctuations were almost the same in low-pressure/temperature air and high-pressure/temperature steam. We found that ordinary pressure/temperature air experiments and simulations can possibly clarify the characteristics of the resonance in high-pressure/temperature steam. (author)

  1. Accelerated growth of oxide film on aluminium alloys under steam: Part II: Effects of alloy chemistry and steam vapour pressure on corrosion and adhesion performance

    DEFF Research Database (Denmark)

    Din, Rameez Ud; Bordo, Kirill; Jellesen, Morten Stendahl

    2015-01-01

    The steam treatment of aluminium alloys with varying vapour pressure of steamresulted in the growth of aluminium oxyhydroxide films of thickness range between 450 - 825nm. The surface composition, corrosion resistance, and adhesion of the produced films was characterised by XPS, potentiodynamic p...... of the vapour pressure of the steam. The accelerated corrosion and adhesion tests on steam generated oxide films with commercial powder coating verified that the performance of the oxide coating is highly dependent on the vapour pressure of the steam....... polarization, acetic acid salt spray, filiform corrosion test, and tape test. The oxide films formed by steam treatment showed good corrosion resistance in NaCl solution by significantly reducing anodic and cathodic activities. The pitting potential of the surface treated with steam was a function...

  2. Lower pressure heating steam is practical for the distributed dry dilute sulfuric acid pretreatment.

    Science.gov (United States)

    Shao, Shuai; Zhang, Jian; Hou, Weiliang; Qureshi, Abdul Sattar; Bao, Jie

    2017-08-01

    Most studies paid more attention to the pretreatment temperature and the resulted pretreatment efficiency, while ignored the heating media and their scalability to an industry scale. This study aimed to use a relative low pressure heating steam easily provided by steam boiler to meet the requirement of distributed dry dilute acid pretreatment. The results showed that the physical properties of the pretreated corn stover were maintained stable using the steam pressure varying from 1.5, 1.7, 1.9 to 2.1MPa. Enzymatic hydrolysis and high solids loading simultaneous saccharification and fermentation (SSF) results were also satisfying. CFD simulation indicated that the high injection velocity of the low pressure steam resulted in a high steam holdup and made the mixing time of steam and solid corn stover during pretreatment much shorter in comparison with the higher pressure steam. This study provides a design basis for the boiler requirement in distributed pretreatment concept. Copyright © 2017 Elsevier Ltd. All rights reserved.

  3. X-ray diffractometry of steam cured ordinary Portland and blast-furnace-slag cements; Difratometria de raios X de pastas de cimento Portland comum e de alto-forno submetidas a cura termica

    Energy Technology Data Exchange (ETDEWEB)

    Camarini, G [Universidade Estadual de Campinas, SP (Brazil). Faculdade de Engenharia; Djanikian, J G [Sao Paulo Univ., SP (Brazil). Escola Politecnica

    1994-12-31

    This work studies some aspects of the phases produced by hydration of ordinary and blast-furnace-slag cements, at normal conditions and steam cured (60 and 95{sup 0} C), using an X-ray diffraction technique. The blast-furnace-slag cement was a mixture of 50% of ordinary Portland cement and 50% of blast-furnace-slag (separately grinding). After curing the X-ray diffraction reveals that, in relation to ordinary Portland cement, the main phases in blast-furnace-slag cement are hydrated silicates and aluminates, hydro garnet, etringitte and mono sulphate. After steam curing the hydration of blast-furnace-slag cement proceeds. This is a result of the slag activation by the curing temperature. (author). 8 refs., 3 figs., 1 tab.

  4. Critical review of use of high pressure saturated steam turbine economizers in nuclear power plants

    International Nuclear Information System (INIS)

    Urbanek, J.

    1981-01-01

    In the high-pressure part of the turbine drops of moisture condensate, which causes erosion and has negative impact on the service-life of the turbine and on its thermodynamic efficiency. Various designs have been put forward to eliminate moisture. A good combination is moisture separation combined with the offtake of steam for the regeneration of feed water or for the steam re-heater. As concerns the high-pressure component of the turbine it is best to offtake steam for the feed water heater and for heating the steam between the high- and low-pressure components of the turbine. The connections of the heater and re-heater in diagrams of various manufacturers are evaluated and compared. It appears to be uneconomical to use the heater in cases where feed water would be heated to temperature considerably below its optimal value. (M.D.)

  5. Testing of acoustic emission method during pressure tests of WWER-440 steam generators and pressurizers

    International Nuclear Information System (INIS)

    Wuerfl, K.; Crha, J.

    1987-01-01

    The results are discussed of measuring acoustic emission in output pressure testing of steam generators and pressurizers for WWER-440 reactors. The objective of the measurements was to test the reproducibility of measurements and to find the criterion which would be used in assessing the condition of the components during manufacture and in operation. The acoustic emission was measured using a single-channel Dunegan/Endevco apparatus and a 16-channel LOCAMAT system. The results showed that after the first assembly, during a repeat dismantle of the lids and during seal replacement, processes due to seal contacts and bolt and washer deformations were the main source of acoustic emission. A procedure was defined of how to exclude new acoustic emission sources in such cases. The acoustic emission method can be used for the diagnostics of plastic deformation processes or of crack production and propagation in components during service. (Z.M.)

  6. Worldwide assessment of steam-generator problems in pressurized-water-reactor nuclear power plants

    International Nuclear Information System (INIS)

    Woo, H.H.; Lu, S.C.

    1981-01-01

    Objective is to assess the reliability of steam generators of pressurized water reactor (PWR) power plants in the United States and abroad. The assessment is based on operation experience of both domestic and foreign PWR plants. The approach taken is to collect and review papers and reports available from the literature as well as information obtained by contacting research institutes both here and abroad. This report presents the results of the assessment. It contains a general background of PWR plant operations, plant types, and materials used in PWR plants. A review of the worldwide distribution of PWR plants is also given. The report describes in detail the degradation problems discovered in PWR steam generators: their causes, their impacts on the performance of steam generators, and the actions to mitigate and avoid them. One chapter is devoted to operating experience of PWR steam generators in foreign countries. Another discusses the improvements in future steam generator design

  7. Design study on steam generator integration into the VVER reactor pressure vessel

    International Nuclear Information System (INIS)

    Hort, J.; Matal, O.

    2004-01-01

    The primary circuit of VVER (PWR) units is arranged into loops where the heat generated by the reactor is removed by means of main circulating pumps, loop pipelines and steam generators, all located outside the reactor pressure vessel. If the primary circuit and reactor core were integrated into one pressure vessel, as proposed, e.g., within the IRIS project (WEC), a LOCA situation would be limited by the reactor pressure vessel integrity only. The aim of this design study regarding the integration of the steam generator into the reactor pressure vessel was to identify the feasibility limits and some issues. Fuel elements and the reactor pressure vessel as used in the Temelin NPP were considered for the analysis. From among the variants analyzed, the variant with steam generators located above the core and vertically oriented circulating pumps at the RPV lower bottom seems to be very promising for future applications

  8. MEDEA, Steady-State Pressure and Temperature Distribution in He H2O Steam Generator

    International Nuclear Information System (INIS)

    Hansen, Ulf

    1976-01-01

    1 - Nature of physical problem solved: MEDEA calculates the time-independent pressure and temperature distribution in a helium-water steam generator. The changing material properties of the fluids with pressure and temperature are treated exactly. The steam generator may consist of economizer, evaporator, superheater and reheater in variable flow patterns. In case of reheating the high-pressure turbine is taken into account. The main control circuits influencing the behaviour of the system are simulated. These are water spraying of the hot steam, load-dependent control of steam pressure at the HP-turbine inlet and valves before the LP-turbine to ensure constant pressure in the reheater section. Investigations of hydrodynamic flow stability in single tubes can be performed. 2 - Method of solution: The steam generator is calculated as a 1-dimensional model, (i.e. all parallel tubes working under equal conditions) and is divided into small heat exchanger elements with helium and water in ideal parallel or counter flow. The material and thermodynamic properties are kept constant within one element. The calculations start at the cold end of the steam generator and proceed stepwise along the water flow pattern to produce pressure and temperature distributions of helium and water. The gas outlet temperature is changed until convergence is reached with a continuous temperature profile on the gas side. MEDEA chooses the iteration scheme according to flow pattern and other special arrangements in the steam generator. The hydrodynamic stability is calculated for a single tube assuming that all tubes are exposed to the same gas temperature profile and changing the water flow in a single tube will not influence the conditions on the gas side. Varying the water flow by keeping gas temperature constant and repeating the steam generator calculations yield pressure drop and steam temperature as a function of flow rate. 3 - Restrictions on the complexity of the problem: Maximum

  9. Investigation of shock waves in explosive blasts using fibre optic pressure sensors

    Energy Technology Data Exchange (ETDEWEB)

    Watson, S [School of Engineering and Physical Sciences, Heriot-Watt University, Edinburgh EH14 4AS (United Kingdom); MacPherson, W N [School of Engineering and Physical Sciences, Heriot-Watt University, Edinburgh EH14 4AS (United Kingdom); Barton, J S [School of Engineering and Physical Sciences, Heriot-Watt University, Edinburgh EH14 4AS (United Kingdom); Jones, J D C [School of Engineering and Physical Sciences, Heriot-Watt University, Edinburgh EH14 4AS (United Kingdom); Tyas, A [Department of Civil and Structural Engineering, University of Sheffield, Sheffield S1 3JD (United Kingdom); Pichugin, A V [Department of Civil and Structural Engineering, University of Sheffield, Sheffield S1 3JD (United Kingdom); Hindle, A [Department of Civil and Structural Engineering, University of Sheffield, Sheffield S1 3JD (United Kingdom); Parkes, W [Scottish Microelectronics Centre, Kings Buildings, West Mains Road, Edinburgh EH9 3JF (United Kingdom); Dunare, C [Scottish Microelectronics Centre, Kings Buildings, West Mains Road, Edinburgh EH9 3JF (United Kingdom); Stevenson, T [Scottish Microelectronics Centre, Kings Buildings, West Mains Road, Edinburgh EH9 3JF (United Kingdom)

    2005-01-01

    We describe miniature all-optical pressure sensors, fabricated by wafer etching techniques, less than 1mm{sup 2} in overall cross-section with rise times in the {mu}s regime and pressure ranges typically 600 kPa. Their performance is suitable for experimental studies of the pressure-time history for test models exposed to shocks initiated by an explosive charge. The small size and fast response of the sensors promises higher quality data than has been previously available from conventional electrical sensors, with potential improvements to numerical models of blast effects. Provisional results from blast tests will be presented in which up to 6 sensors were multiplexed, embedded within test models in a range of orientations relative to the shock front.

  10. Investigation of shock waves in explosive blasts using fibre optic pressure sensors

    International Nuclear Information System (INIS)

    Watson, S; MacPherson, W N; Barton, J S; Jones, J D C; Tyas, A; Pichugin, A V; Hindle, A; Parkes, W; Dunare, C; Stevenson, T

    2005-01-01

    We describe miniature all-optical pressure sensors, fabricated by wafer etching techniques, less than 1mm 2 in overall cross-section with rise times in the μs regime and pressure ranges typically 600 kPa. Their performance is suitable for experimental studies of the pressure-time history for test models exposed to shocks initiated by an explosive charge. The small size and fast response of the sensors promises higher quality data than has been previously available from conventional electrical sensors, with potential improvements to numerical models of blast effects. Provisional results from blast tests will be presented in which up to 6 sensors were multiplexed, embedded within test models in a range of orientations relative to the shock front

  11. Heat exchanging tube behaviour in steam generators of pressurized water reactors

    International Nuclear Information System (INIS)

    Pastor, D.; Oertel, K.

    1979-01-01

    Based on a comprehensive failure statistics, materials corrosion chemistry and thermohydraulics problems of the tubings of steam generators are considered. A historical review of failures in the tubings of steam generators in pressurized water reactors reflects the often successless measures by designers, manufacturers and operating organizations for preventing failures, especially with regard to materials selection and water regime. It is stated that laboratory tests could not give sufficient information about safe and stable operation of nuclear steam generators unless real constructive, hydrodynamic, thermodynamical and chemical conditions of operation had been taken into account. (author)

  12. Evaluation of anticipatory signal to steam generator pressure control program for 700 MWe Indian pressurized heavy water reactor

    International Nuclear Information System (INIS)

    Pahari, S.; Hajela, S.; Rammohan, H. P.; Malhotra, P. K.; Ghadge, S. G.

    2012-01-01

    700 MWe Indian Pressurized Heavy Water Reactor (IPHWR) is horizontal channel type reactor with partial boiling at channel outlet. Due to boiling, it has a large volume of vapor present in the primary loops. It has two primary loops connected with the help of pressurizer surge line. The pressurizer has a large capacity and is partly filled by liquid and partly by vapor. Large vapor volume improves compressibility of the system. During turbine trip or load rejection, pressure builds up in Steam Generator (SG). This leads to pressurization of Primary Heat Transport System (PHTS). To control pressurization of SG and PHTS, around 70% of the steam generated in SG is dumped into the condenser by opening Condenser Steam Dump Valves (CSDVs) and rest of the steam is released to the atmosphere by opening Atmospheric Steam Discharge Valves (ASDVs) immediately after sensing the event. This is accomplished by adding anticipatory signal to the output of SG pressure controller. Anticipatory signal is proportional to the thermal power of reactor and the proportionality constant is set so that SG pressure controller's output jacks up to ASDV opening range when operating at 100% FP. To simulate this behavior for 700 MWe IPHWR, Primary and secondary heat transport system is modeled. SG pressure control and other process control program have also been modeled to capture overall plant dynamics. Analysis has been carried out with 3-D neutron kinetics coupled thermal hydraulic computer code ATMIKA.T to evaluate the effect of the anticipatory signal on PHT pressure and over all plant dynamics during turbine trip in 700 MWe IPHWR. This paper brings out the results of the analysis with and without considering anticipatory signal in SG pressure control program during turbine trip. (authors)

  13. Influence of the Steam Addition on Premixed Methane Air Combustion at Atmospheric Pressure

    Directory of Open Access Journals (Sweden)

    Mao Li

    2017-07-01

    Full Text Available Steam-diluted combustion in gas turbine systems is an effective approach to control pollutant emissions and improve the gas turbine efficiency. The primary purpose of the present research is to analyze the influence of steam dilution on the combustion stability, flame structures, and CO emissions of a swirl-stabilized gas turbine model combustor under atmospheric pressure conditions. The premixed methane/air/steam flame was investigated with three preheating temperatures (384 K/434 K/484 K and the equivalence ratio was varied from stoichiometric conditions to the flammability limits where the flame was physically blown out from the combustor. In order to represent the steam dilution intensity, the steam fraction Ω defined as the steam to air mass flow rate ratio was used in this work. Exhaust gases were sampled with a water-cooled emission probe which was mounted at the combustor exit. A 120 mm length quartz liner was used which enabled the flame visualization and optical measurement. Time-averaged CH chemiluminescence imaging was conducted to characterize the flame location and it was further analyzed with the inverse Abel transform method. Chemical kinetics calculation was conducted to support and analyze the experimental results. It was found that the LBO (lean blowout limits were increased with steam fraction. CH chemiluminescence imaging showed that with a high steam fraction, the flame length was elongated, but the flame structure was not altered. CO emissions were mapped as a function of the steam fraction, inlet air temperature, and equivalence ratios. Stable combustion with low CO emission can be achieved with an appropriate steam fraction operation range.

  14. Influence of steam leakage through vane, gland, and shaft seals on rotordynamics of high-pressure rotor of a 1,000 MW ultra-supercritical steam turbine

    Energy Technology Data Exchange (ETDEWEB)

    Jiang, P.N. [Shanghai Jiao Tong University, Key Laboratory of Power Machinery and Engineering, Ministry of Education, School of Mechanical Engineering, Shanghai (China); Shanghai Turbine Company, Department of R and D, Shanghai (China); Wang, W.Z.; Liu, Y.Z. [Shanghai Jiao Tong University, Key Laboratory of Power Machinery and Engineering, Ministry of Education, School of Mechanical Engineering, Shanghai (China); Meng, G. [Shanghai Jiao Tong University, State Key Laboratory of Mechanical System and Vibration, School of Mechanical Engineering, Shanghai (China)

    2012-02-15

    A comparative analysis of the influence of steam leakage through vane, gland, and shaft seals on the rotordynamics of the high-pressure rotor of a 1,000 MW ultra-supercritical steam turbine was performed using numerical calculations. The rotordynamic coefficients associated with steam leakage through the three labyrinth seals were calculated using the control-volume method and perturbation analysis. A stability analysis of the rotor system subject to the steam forcing induced by the leakage flow was performed using the finite element method. An analysis of the influence of the labyrinth seal forcing on the rotordynamics was carried out by varying the geometrical parameters pertaining to the tooth number, seal clearance, and inner diameter of the labyrinth seals, along with the thermal parameters with respect to pressures and temperatures. The results demonstrated that the steam forcing with an increase in the length of the blade for the vane seal significantly influences the rotordynamic coefficients. Furthermore, the contribution of steam forcing to the instability of the rotor is decreased and increased with increases in the seal clearance and tooth number, respectively. The comparison of the rotordynamic coefficients associated with steam leakage through the vane seal, gland seal, and shaft seal convincingly disclosed that, although the steam forcing attenuates the stability of the rotor system, the steam turbine is still operating under safe conditions. (orig.)

  15. Energy Analysis of Cascade Heating with High Back-Pressure Large-Scale Steam Turbine

    Directory of Open Access Journals (Sweden)

    Zhihua Ge

    2018-01-01

    Full Text Available To reduce the exergy loss that is caused by the high-grade extraction steam of traditional heating mode of combined heat and power (CHP generating unit, a high back-pressure cascade heating technology for two jointly constructed large-scale steam turbine power generating units is proposed. The Unit 1 makes full use of the exhaust steam heat from high back-pressure turbine, and the Unit 2 uses the original heating mode of extracting steam condensation, which significantly reduces the flow rate of high-grade extraction steam. The typical 2 × 350 MW supercritical CHP units in northern China were selected as object. The boundary conditions for heating were determined based on the actual climatic conditions and heating demands. A model to analyze the performance of the high back-pressure cascade heating supply units for off-design operating conditions was developed. The load distributions between high back-pressure exhaust steam direct supply and extraction steam heating supply were described under various conditions, based on which, the heating efficiency of the CHP units with the high back-pressure cascade heating system was analyzed. The design heating load and maximum heating supply load were determined as well. The results indicate that the average coal consumption rate during the heating season is 205.46 g/kWh for the design heating load after the retrofit, which is about 51.99 g/kWh lower than that of the traditional heating mode. The coal consumption rate of 199.07 g/kWh can be achieved for the maximum heating load. Significant energy saving and CO2 emission reduction are obtained.

  16. Pulsed high-pressure (PHP) drain-down of steam generating system

    International Nuclear Information System (INIS)

    Petrusek, R.A.

    1991-01-01

    This patent describes an improved method of draining down contained reactor-coolant water from the inverted vertical U-tubes of at least one vertical-type steam generator in which the upper inverted U-shaped ends of the tubes are closed and the lower ends thereof are open, the steam generator having a channel head at its lower end including a vertical dividing wall defining a primary water inlet side and a primary water outlet side of the generator, the steam generator having chemical volume control system means and residual heat removal system means, and the steam generator being part of a nuclear-powered steam generating system wherein the reactor-coolant water is normally circulated from and back into the reactor via a loop comprising the steam generator and inlet and outlet conduits connected to the lower end of the steam generator, and the reactor being in communication with pressurizer means and comprising the steps of introducing a gas which is inert to the system and which is under pressure above atmospheric pressure into at least one of the downwardly facing open ends of each of the U-tubes from below the tubesheet in which the open ends of the U-tubes are mounted adjacent the lower end of the steam generator while permitting the water to flow out from the open ends of the U-tubes, the improvement in combination therewith for substantially increasing the effectiveness and efficiency of such water removal from the tubes. It includes determining the parameters effecting a first average volumetric rate of removal for a predetermined period of time, infra, of the reactor-coolant water from the inverted vertical U-tubes, the specific unit for the first average volumetric rate expressing properties identical with the properties expressed in a second average volumetric rate maintained in a later mentioned step

  17. A system for regulating the pressure of resuperheated steam in high temperature gas-cooled reactor power stations

    International Nuclear Information System (INIS)

    Braytenbah, A.S.; Jaegines, K.O.

    1975-01-01

    The invention relates to a system for regulating steam-pressure in the re-superheating portion of a steam-boiler receiving heat from a gas-cooled high temperature nuclear reactor, provided with gas distributing pumps driven by steam-turbines. The system comprises means for generating a pressure signal of desired magnitude for the re-superheating portion, and means for providing a real pressure in the re-superheating portion, means (including a by-passing device) for generating steam-flow rate signal of desired magnitude, a turbine by-pass device comprising a by-pass tapping means for regulating the steam-flow-rate in said turbine according to the desired steam-flow rate signal and means for controlling said by-pass tapping means according to said desired steam-flow-rate signal [fr

  18. Frictional pressure drop of high pressure steam-water two-phase flow in internally helical ribbed tubes

    International Nuclear Information System (INIS)

    Tingkuan, C.; Xuanzheng, C.

    1987-01-01

    It is well known that the internally helical ribbed tubes are effective in suppressing the dry-out in boiling tubes at high pressures, so they are widely used as furnace water wall tubes in modern large steam power boilers. Design of the boilers requires the data on frictional pressure drop characteristics of the ribbed tubes, but they are not sufficient now. This paper describes the experimental results on the adiabatic frictional pressure drop in both horizontal ribbed tubes with measured mean inside diameter of 11.69 mm and 35.42 mm at high pressure from 10 to 21 MPa, mass flow rate from 350 to 3800 kg/m/sup 2/s and steam quality from 0 to 1 in our high pressure electrically heated water loop. Simultaneously, both smooth tubes under the same conditions for comparison. Based on the tests the correlation for determining the frictional pressure drop of internally ribbed tubes are proposed

  19. Effect of Low Pressure End Conditions on Steam Power Plant Performance

    Directory of Open Access Journals (Sweden)

    Ali Syed Haider

    2014-07-01

    Full Text Available Most of the electricity produced throughout the world today is from steam power plants and improving the performance of power plants is crucial to minimize the greenhouse gas emissions and fuel consumption. Energy efficiency of a thermal power plant strongly depends on its boiler-condenser operating conditions. The low pressure end conditions of a condenser have influence on the power output, steam consumption and efficiency of a plant. Hence, the objective this paper is to study the effect of the low pressure end conditions on a steam power plant performance. For the study each component was modelled thermodynamically. Simulation was done and the results showed that performance of the condenser is highly a function of its pressure which in turn depends on the flow rate and temperature of the cooling water. Furthermore, when the condenser pressure increases both net power output and plant efficiency decrease whereas the steam consumption increases. The results can be used to run a steam power cycle at optimum conditions.

  20. Type GQS-1 high pressure steam manifold water level monitoring system

    International Nuclear Information System (INIS)

    Li Nianzu; Li Beicheng; Jia Shengming

    1993-10-01

    The GQS-1 high pressure steam manifold water level monitoring system is an advanced nuclear gauge that is suitable for on-line detecting and monitor in high pressure steam manifold water level. The physical variable of water level is transformed into electrical pulses by the nuclear sensor. A computer is equipped for data acquisition, analysis and processing and the results are displayed on a 14 inch color monitor. In addition, a 4 ∼ 20 mA output current is used for the recording and regulation of water level. The main application of this gauge is for on-line measurement of high pressure steam manifold water level in fossil-fired power plant and other industries

  1. Steam explosions of molten iron oxide drops: easier initiation at small pressurizations

    International Nuclear Information System (INIS)

    Nelson, L.S.; Duda, P.M.

    1982-01-01

    Steam explosions caused by hot molten materials contacting liquid water following a possible light water nuclear reactor core overheat have been investigated by releasing single drops of a core melt simulant, molten iron oxide, into liquid water. Small steam explosions were triggered shortly afterwards by applying a pressure pulse to the water. The threshold peak pulse level above which an explosion always occurs was studied at ambient pressures between 0.083 and 1.12 MPa. It was found that the threshold decreased to a minimum in the range 0.2 - 0.8 MPa and then increased again. The effect of easier initiation as ambient pressure increases may have an important role in the triggering and propagation of a large scale steam explosion through a coarsely premixed dispersion of melt in water. (U.K.)

  2. Zirconium metal-water oxidation kinetics. V. Oxidation of Zircaloy in high pressure steam

    International Nuclear Information System (INIS)

    Pawel, R.E.; Cathcart, J.V.; Campbell, J.J.; Jury, S.H.

    1977-12-01

    A series of scoping tests to determine the influence of steam pressure on the isothermal oxidation kinetics of Zircaloy-4 PWR tubing was undertaken. The oxidation experiments were conducted in flowing steam at 3.45, 6.90, and 10.34 MPa (500, 1000, and 1500 psi) at 905 0 C (1661 0 F), and at 3.45 and 6.90 MPa at 1101 0 C (2014 0 F). A comparison of the results of these experiments with those obtained for oxidation in steam at atmospheric pressure under similar conditions indicated that measurable enhancement of the oxidation rate occurred with increasing pressure at 905 0 C, but not at 1100 0 C

  3. A standing pressure wave hypothesis of oscillating forces generated during a steam line break

    International Nuclear Information System (INIS)

    Tinoco, H.

    2001-01-01

    A rapid glance at the figure depicting the net forces acting on the reactor vessel and internals, as obtained through a CFD simulation of a BWR steam line break, reveals an amazing oscillating regularity of these forces which is in glaring contrast to the chaotic behaviour of the steam pressure field in the steam annulus. Assuming that the decompression process excites and maintains standing pressure waves in the annular cylindrical region constituted by the steam annulus, it is possible to reconstruct the net forces acting on the reactor vessel and internals through the contribution of almost only the first dispersive mode. If a Neumann boundary condition is assumed at the section connecting the steam annulus to the steam dome, the frequency predicted is approximately % 5.9 higher than that of the CFD simulations. However, this connecting section allows wave transmission, and a more appropriate boundary condition should be one of the Robin type. Therefore, this section is modelled as an absorbing wall, and the corresponding normal impedance is calculated using the CFD simulations. Week non-linear effects can also be observed in the calculated forces through the presence of the first subharmonic. By the methodology described above, an estimate of the forces acting on the reactor vessel and internals of unit 3 of Forsmark Nuclear Power Plant has been obtained. (author)

  4. Steam gasification of coal using a pressurized circulating fluidized bed

    International Nuclear Information System (INIS)

    Werner, K.F.J.

    1989-09-01

    Subject of this investigation is the process engineering of a coal gasification using nuclear heat. A special aspect is the efficiency. To this purpose a new method for calculating the kinetics of hard coal steam gasification in a fluidized bed is presented. It is used for evaluations of gasification kinetics in a large-scale process on the basis of laboratory-scale experiments. The method is verified by experimental data from a large-scale gasifier. The investment costs and the operating costs of the designed process are estimated. (orig.) [de

  5. The nuclear physical method for high pressure steam manifold water level gauging and its error

    International Nuclear Information System (INIS)

    Li Nianzu; Li Beicheng; Jia Shengming

    1993-10-01

    A new method, which is non-contact on measured water level, for measuring high pressure steam manifold water level with nuclear detection technique is introduced. This method overcomes the inherent drawback of previous water level gauges based on other principles. This method can realize full range real time monitoring on the continuous water level of high pressure steam manifold from the start to full load of boiler, and the actual value of water level can be obtained. The measuring errors were analysed on site. Errors from practical operation in Tianjin Junliangcheng Power Plant and in laboratory are also presented

  6. Analytical treatment of large leak pressure behavior in LMFBR steam generators

    International Nuclear Information System (INIS)

    Hori, Masao; Miyake, Osamu

    1980-07-01

    Simplified analytical methods applicable to the estimation of initial pressure spike in case of a large leak accident in LMFBR steam generators were devised as follows; (i) Estimation of the initial water leak rate by the centered rarefaction wave method, (ii) Estimation of the initial pressure spike by the one-dimensional compressible method with either the columnar bubble growth model or the spherical bubble growth model. These methods were compared with relevant experimental data or other more elaborate analyses and validated to be usable in simple geometry and limited time span cases. Application of these methods to an actual steam generator case was explained and demonstrated. (author)

  7. New safety experiments in decommissioned superheated steam reactor at Karlstein

    International Nuclear Information System (INIS)

    Koerting, K.

    1986-01-01

    This article gives a concise summary of the Status Report of the Superheated Steam Reactor Safety Program (PHDR) Project, held at KfK on Dec. 5, 1985. The results discussed dealt with fire experiments, shock tests simulating airplane crashes, temperature shocks in the reactor pressure vessel, studies of crack detection in pressure vessels and blasting experiments associated with nuclear plant decommissioning

  8. Impact of pressure losses in small-sized parabolic-trough collectors for direct steam generation

    International Nuclear Information System (INIS)

    Lobón, David H.; Valenzuela, Loreto

    2013-01-01

    Using PTC (parabolic-trough solar collectors) for industrial thermal processes in the temperature range up to 300 °C is not new, but in recent years there is a boosted interest in this type of concentrating solar technology. One of the problems that arise when designing PTC solar fields is how to deal with the pressure losses which are critical when producing saturated steam directly in the collectors. Depending on the characteristics of the collector, mainly on the receiver diameter, and on the nominal process conditions defined, a solar field configuration can be feasible or not. This paper presents a sensitivity analysis done using a software tool developed to study the thermo-hydraulic behaviour of PTC systems using water-steam as heat transfer fluid. In the case study presented, a small-sized PTC designed for industrial process heat applications is considered, which has a focal length of 0.2 m, an aperture area of 2 m 2 , and its receiver pipe has an inner diameter of 15 mm. Varied process conditions are inlet water pressure, temperature, and mass flow rate, solar irradiance and incidence angle of solar radiation. Results show that working pressure definition is particularly critical to make feasible or not the direct steam generation in solar collectors. - Highlights: • DSG (Direct steam generation) in small-sized parabolic-trough collectors. • Thermo-hydraulic sensitivity analysis. • Influence of working pressure and receiver geometry in DSG process

  9. Manufacture of the 300 MW steam generator and pressure stabilizer for Qinshan Nuclear Power Station

    International Nuclear Information System (INIS)

    Qian Yi; Miao Deming.

    1989-01-01

    A brief description of the manufacturing process of the steam generator and pressure stabilizer for 300 MWe Qinshan Nuclear Power Station in Shanghai Boiler Works is presented, with special emphasis on fabrication facilities, test procedures and technological evaluations during the manufaturing process-imcluding deep driling of tubesheets, welding of tubes to tube-sheets and tube rolling tests

  10. Pressure drop-flow rate curves for single-phase steam in Combustion Engineering type steam generator U-tubes during severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Fynan, Douglas A.; Ahn, Kwang-Il, E-mail: kiahn@kaeri.re.kr

    2016-12-15

    Highlights: • Pressure drop-flow rate curves for superheated steam in U-tubes were generated. • Forward flow of hot steam is favored in the longer and taller U-tubes. • Reverse flow of cold steam is favored in short U-tubes. • Steam generator U-tube bundle geometry and tube diameter are important. • Need for correlation development for natural convention heat transfer coefficient. - Abstract: Characteristic pressure drop-flow rate curves are generated for all row numbers of the OPR1000 steam generators (SGs), representative of Combustion Engineering (CE) type SGs featuring square bend U-tubes. The pressure drop-flow rate curves are applicable to severe accident natural circulations of single-phase superheated steam during high pressure station blackout sequences with failed auxiliary feedwater and dry secondary side which are closely related to the thermally induced steam generator tube rupture event. The pressure drop-flow rate curves which determine the recirculation rate through the SG tubes are dependent on the tube bundle geometry and hydraulic diameter of the tubes. The larger CE type SGs have greater variation of tube length and height as a function of row number with forward flow of steam favored in the longer and taller high row number tubes and reverse flow favored in the short low row number tubes. Friction loss, natural convection heat transfer coefficients, and temperature differentials from the primary to secondary side are dominant parameters affecting the recirculation rate. The need for correlation development for natural convection heat transfer coefficients for external flow over tube bundles currently not modeled in system codes is discussed.

  11. Experimental study on steam gasification of coal using molten blast furnace slag as heat carrier for producing hydrogen-enriched syngas

    International Nuclear Information System (INIS)

    Duan, Wenjun; Yu, Qingbo; Wu, Tianwei; Yang, Fan; Qin, Qin

    2016-01-01

    Highlights: • New method for producing HRG by gasification using BFS as heat carrier was proposed. • The continuous experiment of steam gasification in molten BFS was conducted. • The hydrogen-enriched syngas was produced by this method. • The molten BFS waste heat was utilized effectively by steam gasification. • This method could be widely used in steam gasification of different types of coal. - Abstract: The new method for producing hydrogen-enriched syngas (HRG) by steam gasification of coal using molten blast furnace slag (BFS) as heat carrier was established. In order to achieve the HRG production, a gasification system using this method was proposed and constructed. The carbon gasification efficiency (CE), hydrogen yield (YH_2) and cold gasification efficiency (CGE) in the molten slag reactor were measured, and the effects of temperature, S/C (steam to coal) ratio and coal type on the reaction performance were accessed. The results indicated that the preferred temperature was 1350 °C, which ensured the miscibility of coal–steam–slag, the diffusion of reactant in molten BFS as well as recovering waste heat. The optimal S/C ratio was 1.5–2.0 for producing HRG. Under these conditions, the hydrogen fraction was higher than 63% and the gas yield reached to 1.89 Nm"3/kg. The CE and CGE were higher than 96% and 102%, respectively. The YH_2 also reached to 1.20 Nm"3/kg. Meanwhile, different types of coal were successfully gasified in molten BFS reactor for producing HRG. The proposed method enhanced the gasification efficiency of different types of coal, recovered the BFS waste heat effectively, and had important guidance for industrial manufacture.

  12. Steam Pressure-Reducing Station Safety and Energy Efficiency Improvement Project

    Energy Technology Data Exchange (ETDEWEB)

    Lower, Mark D [ORNL; Christopher, Timothy W [ORNL; Oland, C Barry [ORNL

    2011-06-01

    The Facilities and Operations (F&O) Directorate is sponsoring a continuous process improvement (CPI) program. Its purpose is to stimulate, promote, and sustain a culture of improvement throughout all levels of the organization. The CPI program ensures that a scientific and repeatable process exists for improving the delivery of F&O products and services in support of Oak Ridge National Laboratory (ORNL) Management Systems. Strategic objectives of the CPI program include achieving excellence in laboratory operations in the areas of safety, health, and the environment. Identifying and promoting opportunities for achieving the following critical outcomes are important business goals of the CPI program: improved safety performance; process focused on consumer needs; modern and secure campus; flexibility to respond to changing laboratory needs; bench strength for the future; and elimination of legacy issues. The Steam Pressure-Reducing Station (SPRS) Safety and Energy Efficiency Improvement Project, which is under the CPI program, focuses on maintaining and upgrading SPRSs that are part of the ORNL steam distribution network. This steam pipe network transports steam produced at the ORNL steam plant to many buildings in the main campus site. The SPRS Safety and Energy Efficiency Improvement Project promotes excellence in laboratory operations by (1) improving personnel safety, (2) decreasing fuel consumption through improved steam system energy efficiency, and (3) achieving compliance with applicable worker health and safety requirements. The SPRS Safety and Energy Efficiency Improvement Project being performed by F&O is helping ORNL improve both energy efficiency and worker safety by modifying, maintaining, and repairing SPRSs. Since work began in 2006, numerous energy-wasting steam leaks have been eliminated, heat losses from uninsulated steam pipe surfaces have been reduced, and deficient pressure retaining components have been replaced. These improvements helped ORNL

  13. Failure Pressure Estimates of Steam Generator Tubes Containing Wear-type Defects

    International Nuclear Information System (INIS)

    Yoon-Suk Chang; Jong-Min Kim; Nam-Su Huh; Young-Jin Kim; Seong Sik Hwang; Joung-Soo Kim

    2006-01-01

    It is commonly requested that steam generator tubes with defects exceeding 40% of wall thickness in depth should be plugged to sustain all postulated loads with appropriate margin. The critical defect dimensions have been determined based on the concept of plastic instability. This criterion, however, is known to be too conservative for some locations and types of defects. In this context, the accurate failure estimation for steam generator tubes with a defect draws increasing attention. Although several guidelines have been developed and are used for assessing the integrity of defected tubes, most of these guidelines are related to stress corrosion cracking or wall-thinning phenomena. As some of steam generator tubes are also failed due to fretting and so on, alternative failure estimation schemes for relevant defects are required. In this paper, three-dimensional finite element (FE) analyses are carried out under internal pressure condition to simulate the failure behavior of steam generator tubes with different defect configurations; elliptical wastage type, wear scar type and rectangular wastage type defects. Maximum pressures based on material strengths are obtained from more than a hundred FE results to predict the failure of the steam generator tube. After investigating the effect of key parameters such as wastage depth, wastage length and wrap angle, simplified failure estimation equations are proposed in relation to the equivalent stress at the deepest point in wastage region. Comparison of failure pressures predicted according to the proposed estimation scheme with some corresponding burst test data shows good agreement, which provides a confidence in the use of the proposed equations to assess the integrity of steam generator tubes with wear-type defects. (authors)

  14. Influence of particle size distribution on the blast pressure profile from explosives buried in saturated soils

    Science.gov (United States)

    Rigby, S. E.; Fay, S. D.; Tyas, A.; Clarke, S. D.; Reay, J. J.; Warren, J. A.; Gant, M.; Elgy, I.

    2017-06-01

    The spatial and temporal distribution of pressure and impulse from explosives buried in saturated cohesive and cohesionless soils has been measured experimentally for the first time. Ten experiments have been conducted at quarter-scale, where localised pressure loading was measured using an array of 17 Hopkinson pressure bars. The blast pressure measurements are used in conjunction with high-speed video filmed at 140,000 fps to investigate in detail the physical processes occurring at the loaded face. Two coarse cohesionless soils and one fine cohesive soil were tested: a relatively uniform sand, a well-graded sandy gravel, and a fine-grained clay. The results show that there is a single fundamental loading mechanism when explosives are detonated in saturated soil, invariant of particle size and soil cohesion. It is also shown that variability in localised loading is intrinsically linked to the particle size distribution of the surrounding soil.

  15. Influence of particle size distribution on the blast pressure profile from explosives buried in saturated soils

    Science.gov (United States)

    Rigby, S. E.; Fay, S. D.; Tyas, A.; Clarke, S. D.; Reay, J. J.; Warren, J. A.; Gant, M.; Elgy, I.

    2018-05-01

    The spatial and temporal distribution of pressure and impulse from explosives buried in saturated cohesive and cohesionless soils has been measured experimentally for the first time. Ten experiments have been conducted at quarter-scale, where localised pressure loading was measured using an array of 17 Hopkinson pressure bars. The blast pressure measurements are used in conjunction with high-speed video filmed at 140,000 fps to investigate in detail the physical processes occurring at the loaded face. Two coarse cohesionless soils and one fine cohesive soil were tested: a relatively uniform sand, a well-graded sandy gravel, and a fine-grained clay. The results show that there is a single fundamental loading mechanism when explosives are detonated in saturated soil, invariant of particle size and soil cohesion. It is also shown that variability in localised loading is intrinsically linked to the particle size distribution of the surrounding soil.

  16. Statistical evaluation of steam condensation loads in pressure suppression pool, (1)

    International Nuclear Information System (INIS)

    Kukita, Yutaka; Takeshita, Isao; Namatame, Ken; Shiba, Masayoshi; Kato, Masami; Moriya, Kumiaki.

    1981-10-01

    The LOCA steam condensation loads in the BWR pressure suppression pool was evaluated with use of the test data obtained in the first eight tests of the JAERI Full-Scale Mark II CRT Program. Through this evaluation, finite desynchronization between the vent pressures during the chugging and the condensation oscillation phases was identified and quantified. The characteristics of the pressure oscillation propagation through the vent pipe and in the pool water, the fluid-structure-interaction (FSI) effects on the pool pressure loads, and the characteristics of the vent lateral loads were also investigated. (author)

  17. The deformation of zircaloy PWR cladding with low internal pressures, under mainly convective cooling by steam

    International Nuclear Information System (INIS)

    Hindle, E.D.; Mann, C.A.; Reynolds, A.E.

    1981-01-01

    The deformation behaviour is reported of specimens of Zircaloy PWR fuel cladding when directly heated in flowing steam. The range of internal pressures studied was 0.69-2.07 MPa; this extended earlier studies using higher pressures. The specimens were ramped and then held at a steady test temperature until rupture or until 600 seconds had elapsed. Under these conditions it was found that extended deformation occurred with pressures down to 1 MPa at temperatures up to 900 deg C. At lower pressures and higher temperatures there was no large extended deformation; this is believed to result from the effects of oxidation

  18. Ultrafast Fabry-Perot fiber-optic pressure sensors for multimedia blast event measurements.

    Science.gov (United States)

    Zou, Xiaotian; Wu, Nan; Tian, Ye; Zhang, Yang; Fitek, John; Maffeo, Michael; Niezrecki, Christopher; Chen, Julie; Wang, Xingwei

    2013-02-20

    A shock wave (SW) is characterized as a large pressure fluctuation that typically lasts only a few milliseconds. On the battlefield, SWs pose a serious threat to soldiers who are exposed to explosions, which may lead to blast-induced traumatic brain injuries. SWs can also be used beneficially and have been applied to a variety of medical treatments due to their unique interaction with tissues and cells. Consequently, it is important to have sensors that can quantify SW dynamics in order to better understand the physical interaction between body tissue and the incident acoustic wave. In this paper, the ultrafast fiber-optic sensor based on the Fabry-Perot interferometric principle was designed and four such sensors were fabricated to quantify a blast event within different media, simultaneously. The compact design of the fiber-optic sensor allows for a high degree of spatial resolution when capturing the wavefront of the traveling SW. Several blast event experiments were conducted within different media (e.g., air, rubber membrane, and water) to evaluate the sensor's performance. This research revealed valuable knowledge for further study of SW behavior and SW-related applications.

  19. The Concept of Steam Pressure Control by Changing the Feedwater Flow during Heatup Operation for an Integral Reactor with a Once-Through Steam Generator

    International Nuclear Information System (INIS)

    Seo, Jae Kwang; Choi, Ki Yong; Kang, Han Ok; Kim, Young In; Yoon, Ju Hyeon; Zee, Sung Qunn

    2005-01-01

    The design features of a once-through steam generator (OTSG) for an integral reactor are significantly different from the commercial U-tube type steam generator from several aspects such as the general arrangement, size, operation conditions, and so on. Therefore a sufficient understanding of the thermal-hydraulic characteristics of the OTSG is essential for the design of the nuclear steam supply system (NSSS) and the power conversion system (PCS). It is also necessary to develop operation procedures complying to the unique design features of the OTSG of interest. The OTSG is sized to produce a sufficiently superheated steam during a normal power operation and therefore the secondary system can be simple relative to that of the other types of steam generators. For the plant adopting the OTSG, the steam pressure in the secondary circuit (tube side of the OTSG) is controlled to be constant during a normal power operation. Constant steam pressure is realized by regulating the control valve on the main steam line dedicated for this purpose. However during a heatup operation, at which the fluid state at the exit of the OTSG is a single phase hot water or two phases, it is not proper to use the control valve on the main steam line due to a control problem at low and multi-phase flow conditions and possibly an erosion problem. For these reasons, another dedicated line called a startup cooling line is used during a heatup condition. There may be several operational conditions for the secondary fluid required to pass through during heatup operation, depending on the design of the PCS. In general, there are two conditions: One is a condition for a vacuum operation for the condenser and another is an entry condition for a steam pressure control operation for an auxiliary power system. In this study, the concept of using a simple startup cooling line with a fixed flow resistance and changing the feedwater flow for the pressure control of the PCS during a heatup period are

  20. Literature investigation of air/steam ingress through small cracks in concrete wall under pressure differences

    International Nuclear Information System (INIS)

    Jiang, J.T.

    2008-01-01

    Traditionally within CANDU safety analysis, a loss coefficient of ∼2.8 is used to characterize turbulent flow leakage through narrow, sharp-edged cracks into, and out of Steam Protected Rooms (SPRs). In the event of main steam line break (MSLB), the pressure differences observed between SPRs and the surrounding area of the powerhouse range from 0.01kPa to 0.1 kPa. The relatively low pressure differences, coupled with narrow crack sizes, for instance, below 1 mm, may result in laminar flow leakage pathways as opposed to the turbulent variety assumed in analysis. The main purpose of this paper is thus (a) to calculate the loss coefficient for laminar flow through small cracks; and (b) to assess the effect of steam ingress to SPRs when the flow through some or all of the room leakage area is assumed to be laminar. Based on the literature review, the loss coefficient for laminar flow, through 1 mm crack size at 0.1 kPa pressure difference, ranges from 10 to about 65. This value represents an increase in loss coefficient of 3 ∼ 22 times the loss coefficient used for SPR safety analysis. The actual volumetric leakage rate is therefore 3 ∼ 8 times smaller than the amount previously applied. This paper demonstrates how the traditional loss coefficient used in safety analysis is extremely conservative in the analysis of the SPRs steam ingress phenomenon. (author)

  1. Literature investigation of air/steam ingress through small cracks in concrete wall under pressure differences

    Energy Technology Data Exchange (ETDEWEB)

    Jiang, J.T. [McMaster Univ., Engineering Physics Dept., Hamilton, Ontario (Canada)], E-mail: jiangj3@mcmaster.ca

    2008-07-01

    Traditionally within CANDU safety analysis, a loss coefficient of {approx}2.8 is used to characterize turbulent flow leakage through narrow, sharp-edged cracks into, and out of Steam Protected Rooms (SPRs). In the event of main steam line break (MSLB), the pressure differences observed between SPRs and the surrounding area of the powerhouse range from 0.01kPa to 0.1 kPa. The relatively low pressure differences, coupled with narrow crack sizes, for instance, below 1 mm, may result in laminar flow leakage pathways as opposed to the turbulent variety assumed in analysis. The main purpose of this paper is thus (a) to calculate the loss coefficient for laminar flow through small cracks; and (b) to assess the effect of steam ingress to SPRs when the flow through some or all of the room leakage area is assumed to be laminar. Based on the literature review, the loss coefficient for laminar flow, through 1 mm crack size at 0.1 kPa pressure difference, ranges from 10 to about 65. This value represents an increase in loss coefficient of 3 {approx} 22 times the loss coefficient used for SPR safety analysis. The actual volumetric leakage rate is therefore 3 {approx} 8 times smaller than the amount previously applied. This paper demonstrates how the traditional loss coefficient used in safety analysis is extremely conservative in the analysis of the SPRs steam ingress phenomenon. (author)

  2. Pressure distribution due to steam bubble collapse in a BWR suppression chamber

    International Nuclear Information System (INIS)

    Giencke, E.

    1979-01-01

    For the pressure time history at the walls of a suppression chamber due to a steam bubble collaps at the condenser pipes interests, expecially the influence of the wall elasticity and the position of the condenser pipes. Two problems are to solve: the pressure time history in the steam bubble and at the walls during the collaps and the pressure distribution at the walls. Both problems are coupled with each other, but the influence of the wall elasticity on the pressure time history in the steam bubble is usually small. Thus the two problems may be solved one after each other. For simplifying the analysis the steam bubble surface may be idealized as a sphere during the whole collaps time. Then the resulting pressure time history is be put on the fluid-structure-system. To show the influence of the containment-elasticity it is favourable to investigate both the rigid and the elastic containment. Because the condenser pipes are arranged in a regular scheme, two limit loading cases are to distinguish. Collapses occur simultaneously with the same intensity at all condenser pipes and a strong collaps occurs only at one condenser pipe or a small group of pipes. When including wall elasticity first the modes of the fluid-structure-system are to analyse and then the dynamical responses of the modes. The coupling effects between the pressure time history in the bubble and at the walls are discussed and then how the membrane and bending stiffness of the walls and the buttomstructure influence the pressure distribution, both for steel and concrete structure. Finally simple models for the analysis are derived and the analytical results are compared with experiments. (orig.)

  3. Accelerated growth of oxide film on aluminium alloys under steam: Part I: Effects of alloy chemistry and steam vapour pressure on microstructure

    DEFF Research Database (Denmark)

    Din, Rameez Ud; Gudla, Visweswara C.; Jellesen, Morten S.

    2015-01-01

    of the oxide layeras well as the compactness increased with steam vapour pressure. The increase in vapour pressure also resulted in a better coverage over the intermetallic particles. Oxide layer showed a layered structure with more compact layer at the Al interface and a nano-scale needle like structure...

  4. Dual Pressure versus Hybrid Recuperation in an Integrated Solid Oxide Fuel Cell Cycle – Steam Cycle

    DEFF Research Database (Denmark)

    Rokni, Masoud

    2014-01-01

    A SOFC (solid oxide fuel cell) cycle running on natural gas was integrated with a ST (steam turbine) cycle. The fuel is desulfurized and pre-reformed before entering the SOFC. A burner was used to combust the remaining fuel after the SOFC stacks. The off-gases from the burner were used to produce...... pressure configuration steam cycle combined with SOFC cycle (SOFC-ST) was new and has not been studied previously. In each of the configuration, a hybrid recuperator was used to recovery the remaining energy of the off-gases after the HRSG. Thus, four different plants system setups were compared to each...... other to reveal the most superior concept with respect to plant efficiency and power. It was found that in order to increase the plant efficiency considerably, it was enough to use a single pressure with a hybrid recuperator instead of a dual pressure Rankine cycle....

  5. Analysis of the Instability Phenomena Caused by Steam in High-Pressure Turbines

    Directory of Open Access Journals (Sweden)

    Paolo Pennacchi

    2011-01-01

    Full Text Available Instability phenomena in steam turbines may happen as a consequence of certain characteristics of the steam flow as well as of the mechanical and geometrical properties of the seals. This phenomenon can be modeled and the raise of the steam flow and pressure causes the increase of the cross coupled coefficients used to model the seal stiffness. As a consequence, the eigenvalues and eigenmodes of the mathematical model of the machine change. The real part of the eigenvalue associated with the first flexural normal mode of the turbine shaft may become positive causing the conditions for unstable vibrations. The original contribution of the paper is the application of a model-based analysis of the dynamic behavior of a large power unit, affected by steam-whirl instability phenomena. The model proposed by the authors allows studying successfully the experimental case. The threshold level of the steam flow that causes instability conditions is analyzed and used to define the stability margin of the power unit.

  6. Steam-explosion pretreatment of wood: effect of chip size, acid, moisture content and pressure drop

    Energy Technology Data Exchange (ETDEWEB)

    Brownell, H.H.; Yu, E.K.C.; Saddler, J.N.

    1986-06-01

    Material balances for pentosan, lignin, and hexosan, during steam-explosion pretreatment of aspenwood, showed almost quantitative recovery of cellulose in the water-insoluble fraction. Dilute acid impregnation resulted in more selective hydrolysis of pentosan relative to undesirable pyrolysis, and gave a more accessible substrate for enzymatic hydrolysis. Thermocouple probes, located inside simulated aspenwood chips heated in 240 degrees C-saturated steam, showed rapid heating of air-dry wood, whereas green or impregnated wood heated slowly. Small chips, 3.2 mm in the fiber direction, whether green or air dry gave approximately equal rates of pentosan destruction and solubilization, and similar yields of glucose and of total reducing sugars on enzmatic hydrolysis with Trichoderma harzianum. Partial pyrolysis, destroying one-third of the pentosan of aspenwood at atmospheric pressure by dry steam at 276 degrees C, gave little increase in yield of reducing sugars on enzymatic hydrolysis. Treatment with saturated steam at 240 degrees C gave essentially the same yields of butanediol and ethanol on fermentation with Klebsiella pneumoniae, whether or not 80% of the steam was bled off before explosion and even if the chips remained intact, showing that explosion was unnecessary. 17 references.

  7. Device for the condensation of pressurized steam and its application to the cooling of a nuclear reactor after an incident

    International Nuclear Information System (INIS)

    Dagard, P.; Couturier, M.

    1989-01-01

    This document describes an invention which relates to a device for condensation of pressurized water which is at a pressure considerably above atmospheric pressure, such as the steam produced by the steam generator of a pressurized-water nuclear reactor during the cooling of the reactor after an incident. The purpose of the invention is therefore to propose a device for the condensation of steam which is under a pressure which is considerably higher than atmospheric pressure by cooling this circulating steam as a result of contact with a heat-exchange wall which is cooled by water; such a device should be easy to install in a nuclear power plant to ensure passive cooling of the reactor, it should have a very good efficiency because of efficient heat exchangers, and it should require only a limited amount of cooling water in the equipment itself

  8. Use of superheated steam to anneal the reactor pressure vessel

    International Nuclear Information System (INIS)

    Porowski, J.S.

    1994-01-01

    Thermal annealing of an embrittled Reactor Pressure Shell is the only recognized means for recovering material properties lost due to long-term exposure of the reactor walls to radiation. Reduced toughness of the material during operation is a major concern in evaluations of structural integrity of older reactors. Extensive studies performed within programs related to life extension of nuclear plants have confirmed that the thermal treatment of 850 degrees F for 168 hours on irradiated material essentially recovers material properties lost due to neutron exposure. Dry and wet annealing methods have been considered. Wet annealing involves operating the reactor at near design temperatures and pressures. Since the temperature of wet annealing must be limited to vessel design temperature of 650 degrees F, only partial recovery of the lost properties is achieved. Thus dry annealing was selected as an alternative for future development and industrial implementation to extend the safe life of reactors

  9. Modelling of blowdown of steam in the pressurized PPOOLEX facility

    Energy Technology Data Exchange (ETDEWEB)

    Paettikangas, T.; Niemi, J.; Timperi, A. (VTT Technical Research Centre of Finland (Finland))

    2009-12-15

    PPOOLEX experiment WLL-04-02 on condensation of vapour is studied with CFD simulations. Wall condensation model has been adapted to an Euler-Euler multiphase model of the Fluent CFD code for this purpose. In addition, a simple direct-contact condensation model has also been included in the code. The main focus of the CFD modelling work was on modelling condensation in the drywell. The amount of condensation found in the CFD calculation was in fair agreement with the experiment. The present simulation was so short that the gas flowing into the wetwell contained significant amount of air. The mole fraction of vapour at the outlet of the vent pipe had the maximum value of about 0.3. Therefore, the noncondensable gas strongly affected the direct-contact condensation in the water pool. Much longer simulations are needed in order to study jugging and condensation oscillations. FSI calculations of the experiments were performed by using the Star-CD, ABAQUS and MpCCI codes. An approximate method that makes possible numerically stable FSI calculations for the experimental facilities was used. The method is based on linear perturbation method which necessitates small structural deformations. The calculations showed that FSI has to be taken into account for the POOLEX facility which has relatively light structures. A way for determining the pressure source for the acoustic model from pressure measured at the pool bottom was also examined. Separation of the pressure component due to wall motion from the blowdown load was attempted by conducting a Fourier analysis on the measured displacement signal. The study showed that in practise sufficiently accurate acceleration signal cannot be obtained this way because the transformed signal gets easily out of phase. A measurement system was proposed which could be used for determining the pressure fluctuations. (author)

  10. Modelling of blowdown of steam in the pressurized PPOOLEX facility

    International Nuclear Information System (INIS)

    Paettikangas, T.; Niemi, J.; Timperi, A.

    2009-12-01

    PPOOLEX experiment WLL-04-02 on condensation of vapour is studied with CFD simulations. Wall condensation model has been adapted to an Euler-Euler multiphase model of the Fluent CFD code for this purpose. In addition, a simple direct-contact condensation model has also been included in the code. The main focus of the CFD modelling work was on modelling condensation in the drywell. The amount of condensation found in the CFD calculation was in fair agreement with the experiment. The present simulation was so short that the gas flowing into the wetwell contained significant amount of air. The mole fraction of vapour at the outlet of the vent pipe had the maximum value of about 0.3. Therefore, the noncondensable gas strongly affected the direct-contact condensation in the water pool. Much longer simulations are needed in order to study jugging and condensation oscillations. FSI calculations of the experiments were performed by using the Star-CD, ABAQUS and MpCCI codes. An approximate method that makes possible numerically stable FSI calculations for the experimental facilities was used. The method is based on linear perturbation method which necessitates small structural deformations. The calculations showed that FSI has to be taken into account for the POOLEX facility which has relatively light structures. A way for determining the pressure source for the acoustic model from pressure measured at the pool bottom was also examined. Separation of the pressure component due to wall motion from the blowdown load was attempted by conducting a Fourier analysis on the measured displacement signal. The study showed that in practise sufficiently accurate acceleration signal cannot be obtained this way because the transformed signal gets easily out of phase. A measurement system was proposed which could be used for determining the pressure fluctuations. (author)

  11. Steam turbine cycle

    International Nuclear Information System (INIS)

    Okuzumi, Naoaki.

    1994-01-01

    In a steam turbine cycle, steams exhausted from the turbine are extracted, and they are connected to a steam sucking pipe of a steam injector, and a discharge pipe of the steam injector is connected to an inlet of a water turbine. High pressure discharge water is obtained from low pressure steams by utilizing a pressurizing performance of the steam injector and the water turbine is rotated by the high pressure water to generate electric power. This recover and reutilize discharged heat of the steam turbine effectively, thereby enabling to improve heat efficiency of the steam turbine cycle. (T.M.)

  12. Remote-controlled television for locating leaking tubes in pressurized-water reactor steam generators

    International Nuclear Information System (INIS)

    Cormault, P.; Denis, J.

    1978-01-01

    The Scarabee system is designed for observation of the tubes in water boxes of pressurized-water reactor nuclear-power-station steam generators. It consists essentially of a camera and a projector used as a marker, both of which swivel freely. The whole unit is housed in a water-tight container which can easily be decontaminated. Remote control of camera and marker movement is carried out from a console. (author)

  13. Investigation of Bioethanol Productivity from Sargassum sp. (Brown Seaweed) by Pressure Cooker and Steam Cooker Pretreatments

    International Nuclear Information System (INIS)

    Yu Yu Wah; Kyaw Nyein Aye; Tint Tint Kyaw; Moe Moe Kyaw

    2011-12-01

    Production of biothanol from Sargassum sp. (Brown seaweed) is more suitable than using any other raw materials because it can easily collect on Chaung Tha Beach in Myanmar without any environmental damages. In this regard an attempt for bioethanol production from sargassum sp. by separation hydrolysis and fermentation (SHF) with saccharomyces cerevisiae was made. Sargassum sp. was pretreated with steam cooker at 120 C and 1 bar for 30 min and pressure cooker at 65 C for 2 hour. The pretreated sargassum sp. was liquefied with the crude enzyme from Trichoderma sp. at the temperature of 50 C and pH of 4 for the first liquefaction step and 95 C, pH of 5 and enzyme of SPEZYME FERD were employed for the second liquefaction step. OPTIDEX L-400 was used as saccharified enzyme with the temperature of 65 C and pH of 4.5 at saccharification step. The process of fermentation was followed by distillation at 78 C for alcohol extraction. Concentrations of crude ethanol were about 1.8% by using steam cooker and 2% for pressure cooker treatment with enzyme mediated saccharification followed by yeast fermentation. Yields of bioethanol were 23% for pressure cooker treatment and 21% for steam cooker treatment at SHF process.

  14. Study of steam condensation at sub-atmospheric pressure: setting a basic research using MELCOR code

    Science.gov (United States)

    Manfredini, A.; Mazzini, M.

    2017-11-01

    One of the most serious accidents that can occur in the experimental nuclear fusion reactor ITER is the break of one of the headers of the refrigeration system of the first wall of the Tokamak. This results in water-steam mixture discharge in vacuum vessel (VV), with consequent pressurization of this container. To prevent the pressure in the VV exceeds 150 KPa absolute, a system discharges the steam inside a suppression pool, at an absolute pressure of 4.2 kPa. The computer codes used to analyze such incident (eg. RELAP 5 or MELCOR) are not validated experimentally for such conditions. Therefore, we planned a basic research, in order to have experimental data useful to validate the heat transfer correlations used in these codes. After a thorough literature search on this topic, ACTA, in collaboration with the staff of ITER, defined the experimental matrix and performed the design of the experimental apparatus. For the thermal-hydraulic design of the experiments, we executed a series of calculations by MELCOR. This code, however, was used in an unconventional mode, with the development of models suited respectively to low and high steam flow-rate tests. The article concludes with a discussion of the placement of experimental data within the map featuring the phenomenon characteristics, showing the importance of the new knowledge acquired, particularly in the case of chugging.

  15. Themoeconomic optimization of triple pressure heat recovery steam generator operating parameters for combined cycle plants

    Directory of Open Access Journals (Sweden)

    Mohammd Mohammed S.

    2015-01-01

    Full Text Available The aim of this work is to develop a method for optimization of operating parameters of a triple pressure heat recovery steam generator. Two types of optimization: (a thermodynamic and (b thermoeconomic were preformed. The purpose of the thermodynamic optimization is to maximize the efficiency of the plant. The selected objective for this purpose is minimization of the exergy destruction in the heat recovery steam generator (HRSG. The purpose of the thermoeconomic optimization is to decrease the production cost of electricity. Here, the total annual cost of HRSG, defined as a sum of annual values of the capital costs and the cost of the exergy destruction, is selected as the objective function. The optimal values of the most influencing variables are obtained by minimizing the objective function while satisfying a group of constraints. The optimization algorithm is developed and tested on a case of CCGT plant with complex configuration. Six operating parameters were subject of optimization: pressures and pinch point temperatures of every three (high, intermediate and low pressure steam stream in the HRSG. The influence of these variables on the objective function and production cost are investigated in detail. The differences between results of thermodynamic and the thermoeconomic optimization are discussed.

  16. High Materials Performance in Supercritical CO2 in Comparison with Atmospheric Pressure CO2 and Supercritical Steam

    Energy Technology Data Exchange (ETDEWEB)

    Holcomb, Gordon [National Energy Technology Lab. (NETL), Pittsburgh, PA, (United States); Tylczak, Joseph [National Energy Technology Lab. (NETL), Pittsburgh, PA, (United States); Carney, Casey [National Energy Technology Lab. (NETL), Pittsburgh, PA, (United States); Dogan, Omer N. [National Energy Technology Lab. (NETL), Pittsburgh, PA, (United States)

    2017-02-26

    This presentation covers environments (including advanced ultra-supercritical (A-USC) steam boiler/turbine and sCO2 indirect power cycle), effects of pressure, exposure tests, oxidation results, and mechanical behavior after exposure.

  17. Effects of the Pressurized Water Reactor Main Steam Line Break Location on the Blowdown Loading

    International Nuclear Information System (INIS)

    Jo, Jong Chull; Kang, Soon Ho; Chan, Won Joon

    2016-01-01

    The thermal hydraulic analysis has been performed generally using a simple lumped model or one dimensional numerical model. However, those models have limitations in predicting the transient variations of the steam velocity, pressure and hydrodynamic load at a local point and the most conservative condition. Furthermore, it cannot be confirmed if the blowdown loads predicted by either of the models are conservative to evaluate every part of the SG internal structures. In this study, the transient hydraulic response of the SG secondary side to the MSLB case where the pipe break is assumed to occur at the SG outlet nozzle end, another weld point on the MSL, was numerically simulated using a CFD code. The present CFD calculation results was compared with those in ref. to investigate the effect of break location (friction loss) on the blowdown load in the SG secondary side. The result shows that the friction loss along the steam line span between the SG nozzle end and the MSIV would cause reduction in steam velocity disturbance or dynamic pressure. It implies that the consequence of the MSLB at the SG nozzle end would be much severer that those of other MSLB cases where the break locations are far from the SG. Therefore, to assure a conservative safety evaluation of the SG structural integrity, the blowdown loading on the SG internal structures including tubes during a MSLB accident in terms of the transient steam velocity, dynamic pressure and decompression wave fluctuations should be assessed for the MSLB case where the break is assumed to occur at the SG nozzle end.

  18. Effects of the Pressurized Water Reactor Main Steam Line Break Location on the Blowdown Loading

    Energy Technology Data Exchange (ETDEWEB)

    Jo, Jong Chull; Kang, Soon Ho; Chan, Won Joon [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    The thermal hydraulic analysis has been performed generally using a simple lumped model or one dimensional numerical model. However, those models have limitations in predicting the transient variations of the steam velocity, pressure and hydrodynamic load at a local point and the most conservative condition. Furthermore, it cannot be confirmed if the blowdown loads predicted by either of the models are conservative to evaluate every part of the SG internal structures. In this study, the transient hydraulic response of the SG secondary side to the MSLB case where the pipe break is assumed to occur at the SG outlet nozzle end, another weld point on the MSL, was numerically simulated using a CFD code. The present CFD calculation results was compared with those in ref. to investigate the effect of break location (friction loss) on the blowdown load in the SG secondary side. The result shows that the friction loss along the steam line span between the SG nozzle end and the MSIV would cause reduction in steam velocity disturbance or dynamic pressure. It implies that the consequence of the MSLB at the SG nozzle end would be much severer that those of other MSLB cases where the break locations are far from the SG. Therefore, to assure a conservative safety evaluation of the SG structural integrity, the blowdown loading on the SG internal structures including tubes during a MSLB accident in terms of the transient steam velocity, dynamic pressure and decompression wave fluctuations should be assessed for the MSLB case where the break is assumed to occur at the SG nozzle end.

  19. Design of air blast pressure sensors based on miniature silicon membrane and piezoresistive gauges

    Science.gov (United States)

    Riondet, J.; Coustou, A.; Aubert, H.; Pons, P.; Lavayssière, M.; Luc, J.; Lefrançois, A.

    2017-11-01

    Available commercial piezoelectric pressure sensors are not able to accurately reproduce the ultra-fast transient pressure occurring during an air blast experiment. In this communication a new pressure sensor prototype based on a miniature silicon membrane and piezoresistive gauges is reported for significantly improving the performances in terms of time response. Simulation results demonstrate the feasibility of a pressure transducer having a fundamental resonant frequency almost ten times greater than the commercial piezoelectric sensors one. The sensor uses a 5μm-thick SOI membrane and four P-type silicon gauges (doping level ≅ 1019 at/cm3) in Wheatstone bridge configuration. To obtain a good trade-off between the fundamental mechanical resonant frequency and pressure sensitivity values, the typical dimension of the rectangular membrane is fixed to 30μm x 90μm with gauge dimension of 1μm x 5μm. The achieved simulated mechanical resonant frequency of these configuration is greater than 40MHz with a sensitivity of 0.04% per bar.

  20. Steam drum level dynamics in a multiple loop natural circulation system of a pressure-tube type boiling water reactor

    International Nuclear Information System (INIS)

    Jain, Vikas; Kulkarni, P.P.; Nayak, A.K.; Vijayan, P.K.; Saha, D.; Sinha, R.K.

    2011-01-01

    Highlights: → We have highlighted the problem of drum level dynamics in a multiple loop type NC system using RELAP5 code. → The need of interconnections in steam and liquid spaces close to drum is established. → The steam space interconnections equalize pressure and liquid space interconnections equalize level. → With this scheme, the system can withstand anomalous conditions. → However, the controller is found to be inevitable for inventory balance. - Abstract: Advanced Heavy Water Reactor (AHWR) is a pressure tube type boiling water reactor employing natural circulation as the mode of heat removal under all the operating conditions. Main heat transport system (MHTS) of AHWR is essentially a multi-loop natural circulation system with all the loops connected to each other. Each loop of MHTS has a steam drum that provides for gravity based steam-water separation. Steam drum level is a very critical parameter especially in multi-loop natural circulation systems as large departures from the set point may lead to ineffective separation of steam-water or may affect the driving head. However, such a system is susceptible to steam drum level anomalies under postulated asymmetrical operating conditions among the different quadrants of the core like feedwater flow distribution anomaly among the steam drums or power anomaly among the core quadrants. Analyses were carried out to probe such scenarios and unravel the underlying dynamics of steam drum level using system code RELAP5/Mod3.2. In addition, a scheme to obviate such problem in a passive manner without dependence on level controller was examined. It was concluded that steam drums need to be connected in the liquid as well as steam space to make the system tolerant to asymmetrical operating conditions.

  1. Steam-generator-tube-rupture transients for pressurized-water reactors

    International Nuclear Information System (INIS)

    Dobranich, D.; Henninger, R.J.; DeMuth, N.S.

    1982-01-01

    Steam generator tube ruptures with and without concurrent main-steam-line break are investigated for pressurized water reactors supplied by the major US vendors. The goal of these analyses is to provide thermodynamic and flow conditions for the determination of iodine transport to the environment and to provide an evaluation of the adequacy of the plant safety systems and operating procedures for controlling these transients. The automatic safety systems of the plant were found to be adequate for the mitigation of these transients. Emergency injection system flows equilibrated with the leakage flows and prevented core uncovery. Sufficient time was afforded by the plant safety systems for the operators to identify the problem and to take appropriate measures

  2. Process for resuperheating steam coming from the high-pressure stage of a turbine and device to bring into use this process

    International Nuclear Information System (INIS)

    Pacault, P.H.

    1977-01-01

    A process is described for resuperheating steam coming from the high pressure stage of a turbine fed by a steam generator, itself heated from a base thermal source. The resuperheating is done by desuperheating at least a part of the steam coming from the generator, taken from the inflow of the turbine high pressure stage, the desuperheated steam being condensed, partially at least, in a condensation exchanger forming a preliminary resuperheater [fr

  3. Device for achieving pressure balance in the steam generator of a power plant in case of a main-steam pipe or a feedwater pipe break

    International Nuclear Information System (INIS)

    Wietelmann, F.

    1978-01-01

    In order to increase the safety in the steam generator of a power plant in case of a pipe break, the possibility of a pressure balance between the feedwater inlet and the initial steam outlet chambers is allowed for. According to the invention, the partition wall separating these two chambers will exhibit several overflow openings, each of which will be provided with a closure and half of which may be opened to one side only, care having been taken that in case of an accident on occurrence of a certain differential pressure they will always be opened to the low-pressure side. As closures caps, which may be swing out of the way, or rupture diaphragms are mentioned. (UWI) 891 HP [de

  4. Two-Phase Instability Characteristics of Printed Circuit Steam Generator for the Low Pressure Condition

    International Nuclear Information System (INIS)

    Kang, Han-Ok; Han, Hun Sik; Kim, Young-In; Kim, Keung Koo

    2015-01-01

    Reduction of installation space for steam generators can lead to much smaller reactor vessel with resultant decrease of overall manufacturing cost for the components. A PCHE(Printed Circuit Heat Exchanger) is one of the compact types of heat exchangers available as an alternative to conventional shell and tube heat exchangers. Its name is derived from the procedure used to manufacture the flat metal plates that form the core of the heat exchanger, which is done by chemical milling. These plates are then stacked and diffusion bonded, converting the plates into a solid metal block containing precisely engineered fluid flow passages. PCSG(Printed Circuit Steam Generator) is a potential candidate to be applied to the integral reactor with its compactness and mechanical robustness. For the introduction of new steam generator, design requirement for the two-phase flow instability should be considered. This paper describes two-phase flow instability characteristics of PCSG for the low pressure condition. PCSG is a potential candidate to be applied to the integral reactor with its compactness and mechanical robustness. Interconnecting flow path was developed to mitigate the two-phase flow instability in the cold side. The flow characteristics of two-phase flow instability at the PCSG is examined experimentally in this study

  5. High pressure hydriding of sponge-Zr in steam-hydrogen mixtures

    International Nuclear Information System (INIS)

    Kim, Y.S.

    1997-01-01

    Hydriding kinetics of thin sponge-Zr layers metallurgically bonded to a Zircaloy disk has been studied by thermogravimetry in the temperature range 350-400 C in 7 MPa hydrogen-steam mixtures. Some specimens were prefilmed with a thin oxide layer prior to exposure to the reactant gas; all were coated with a thin layer of gold to avoid premature reaction at edges. Two types of hydriding were observed in prefilmed specimens, viz., a slow hydrogen absorption process that precedes an accelerated (massive) hydriding. At 7 MPa total pressure, the critical ratio of H 2 /H 2 O above which massive hydriding occurs at 400 C is ∝200. The critical H 2 /H 2 O ratio is shifted to ∝2.5 x 10 3 at 350 C. The slow hydriding process occurs only when conditions for hydriding and oxidation are approximately equally favorable. Based on maximum weight gain, the specimen is completely converted to δ-ZrH 2 by massive hydriding in ∝5 h at a hydriding rate of ∝10 -6 mol H/cm 2 s. Incubation times of 10-20 h prior to the onset of massive hydriding increases with prefilm oxide thickness in the range of 0-10 μm. By changing to a steam-enriched gas, massive hydriding that initially started in a steam-starved condition was arrested by re-formation of a protective oxide scale. (orig.)

  6. Pressure transients resulting from sodium-water reaction following a large leak in LMFBR steam generator

    International Nuclear Information System (INIS)

    Rajput, A.K.

    1984-01-01

    The study of sodium water reaction, following a large leak, concerns primarily with the estimation of pressure/flow transients that are developed in the steam generator and the associated secondary circuit. This paper describes the mathematical formulations used in SWRT (Sodium Water Reaction Transients) code developed to estimate such pressure transients for FBTR plant. The results, obtained using SWRT have been presented for a leak in economiser (20m from bottom water header) and for a leak in super heater portions. A time lag of 50 m sec was considered for rupture disc takes to burst once the pressure experienced by it exceeds the set value. Also described in annexure to this paper is the mathematical formulation for two phase transient flow for the better estimation of leak rate from the ruptured end of the damaged heat transfer tube. This leak model considers slip but assumes thermal equilibrium between the liquid and vapour phases

  7. Selective component degradation of oil palm empty fruit bunches (OPEFB) using high-pressure steam

    International Nuclear Information System (INIS)

    Baharuddin, Azhari Samsu; Sulaiman, Alawi; Kim, Dong Hee; Mokhtar, Mohd Noriznan; Hassan, Mohd Ali; Wakisaka, Minato; Shirai, Yoshihito; Nishida, Haruo

    2013-01-01

    In order to accelerate the bioconversion process of press-shredded empty fruit bunches (EFB), the effect of high-pressure steam pre-treatment (HPST) in degrading the lignocellulosic structure was investigated. HPST was carried out under various sets of temperature/pressure conditions such as 170/0.82, 190/1.32, 210/2.03, and 230 °C/3.00 MPa. It was noted that after HPST, the surface texture, color, and mechanical properties of the treated EFB had obviously altered. Scanning electron micrographs of the treated EFB exhibited effective surface erosion that had occurred along the structure. Moreover, the Fourier transform infrared and thermogravimetric analyses showed the removal of silica bodies and hemicellulose ingredients. X-ray diffraction profiles of the treated EFB indicated significant increases in crystallinity. These results reveal that HPST is an effective pre-treatment method for altering the physicochemical properties of the EFB and enhancing its biodegradability characteristics for the bioconversion process. -- Highlights: ► Bioconversion of empty fruit bunches (EFB) was accelerated by high-pressure steam pre-treatment. ► Scanning electron micrographs exhibited surface erosion as well as composting over 20 days. ► FT-IR and TG data showed the selective removal of silica bodies and hemicellulose ingredient. ► X-ray diffraction profiles of the treated EFB indicated significant increases in crystallinity

  8. Automatic systems for opening and closing reactor vessels, steam generators, and pressurizers

    International Nuclear Information System (INIS)

    Samblat, C.

    1990-01-01

    The need for shorter working assignments, reduced dose rates and less time consumption have caused Electricite de France and Framatome to automate the entire procedure of opening and closing the main components in the primary system, such as the reactor vessel, steam generator, and pressurizer. The experience accumulated by the two companies in more than 300 annual revisions of nuclear generating units worldwide has been used as a basis for automating all bolt opening and closing steps as well as cleaning processes. The machines and automatic systems currently in operation are the result of extensive studies and practical tests. (orig.) [de

  9. Apparatus for inspecting and repairing a pressurized-water reactor's steam generator heat exchanger tubes

    International Nuclear Information System (INIS)

    Mueller, O.; Roettger, H.; Kasti, H.; Hagen, H.G.

    1976-01-01

    Described is an apparatus provided for use with a pressurized-water reactor' steam generator having a manifold chamber enclosing the bottom side of a horizontal tube sheet having holes therethrough in which are mounted the tubes of a heat exchanger tube bundle. The manifold chamber has a manhole giving access to the tube's bottom side to permit internal inspection or repair of the tubes by registration of an end of a flexible guide conduit with the tube sheet holes and through which a flexible carrier can be guided for insertion via these holes in the tube sheet and through the tubes extending from the tube sheet's other side

  10. Steam condensation behavior of high pressure water's blow down directly into water in containment under LOCA

    International Nuclear Information System (INIS)

    Kusunoki, Tsuyoshi; Ishida, Toshihisa; Yoritsune, Tsutomu; Kasahara, Y.

    1995-01-01

    JAERI has been conducting a design study of an advanced type Marine Reactor X (MRX) for merchant ships. By employing 'Integral type PWR', In-vessel type control rod drive systems', 'Water filled containment system' and 'Decay heat removal system by natural convection', MRX achieved a compact, light weight and highly safe plant. Experiments on steam condensation behavior of high pressure water's blow down into water have been conducted in order to investigate a major safety issue related to the design decision of 'Water filled containment system'. (author)

  11. Expandable antivibration bar for heat transfer tubes of a pressurized water reactor steam generator

    International Nuclear Information System (INIS)

    Appleman, R.H.

    1985-01-01

    An expandable antivibration bar for use in stabilizing the U-bend portion of heat transfer tubes in a pressurized water reactor steam generator comprises two adjustable rods connected together by an arcuate connector. The two adjustable rods preferably comprise two mating rod sections having complementary angular sliding surfaces thereon, with means provided to move the rod sections relative to each other along the sliding surfaces so as to expand the rods from a first mated cross-sectional width to a second larger cross-sectional width. The ends of the rod sections have means for aligning the two rod sections and maintaining them in alignment during expansion. (author)

  12. R.B. pressure and temperature transient following main steam line break

    International Nuclear Information System (INIS)

    Das, M.; Bhawal, R.N.; Prakash, P.

    1989-01-01

    The R.B. containment plays an important role in mitigating the consequences of any accident core. The analysis of Main Steam Line Break (MSLB), though not of relevance from activity release considerations, is essentially from structural integrity point of view. In this paper the outline of the likely scenario is drawn and the approach for thermal hydraulic simulation of the system for carrying out transient blowdown analysis is discussed. The results of the containment pressure and temperature transient analysis are also presented. (author). 4 refs., 7 figs

  13. Effect of non-condensation gas on pressure oscillation of submerged steam jet condensation

    International Nuclear Information System (INIS)

    Zhao, Quanbin; Cong, Yuelei; Wang, Yingchun; Chen, Weixiong; Chong, Daotong; Yan, Junjie

    2016-01-01

    Highlights: • Oscillation intensity of steam–air jet increases with rise of water temperature. • Oscillation intensity reduces obviously when air is mixed. • Both first and second dominant frequencies decrease with rise of air mass fraction. • Air has little effect on power of 1st & 2nd frequency bands under low temperature. • The maximum oscillation power occurs under case of A = 1% and T ⩾ 50 °C. - Abstract: The effect of air with low mass fraction on the oscillation intensity and oscillation frequency of a submerged steam jet condensation is investigated under stable condensation region. With air mixing in steam, an obvious dynamic pressure peak appears along the jet direction. The intensity peak increases monotonously with the rise of steam mass flux and water temperature. Peak position moves downstream with the rise of air mass fraction. Moreover, when compared with that of pure steam jet, the oscillation intensity clearly decreases as air is mixed. However, when water temperature is lower than approximately 45 °C, oscillation intensity increases slightly with the rise of air mass fraction, and when water temperature is higher than 55 °C, the oscillation intensity decreases greatly with the rise of air mass fraction. Both the first and second dominant frequencies decrease with rise of air mass fraction. Finally, effect of air mass fractions on the oscillation power of the first and second dominant frequency bands shows similar trends. Under low water temperature, the mixed air has little effect on the oscillation power of both first and second frequency bands. However, when water temperature is high, the oscillation power of both first and second frequency bands appears an obvious peak when air mass fraction is about 1%. With further rise of air mass fraction, the oscillation power decreases gradually.

  14. Pressurization rate effect on ligament rupture and burst pressures of cracked steam generator tubes

    International Nuclear Information System (INIS)

    Majumdar, S.; Kasza, K.

    2009-01-01

    The question of whether ligament rupture pressure or unstable burst pressure may vary significantly with pressurization rate at room temperature arose from the results of pressure tests by industry on tubes with machined part-throughwall notches. Slow (quasi-static) and fast 14 MPa/s (2000 psi/s) pressurization rate tests on specimens with nominally the same notch geometry appeared to show a significant effect of the rate of pressurization on the unstable burst pressure. Unfortunately, the slow and fast loading rate tests were conducted following two different test procedures, which could confound the results. The current series of tests were conducted on a variety of specimen geometries using a consistent test procedure to better establish the effect of pressurization rate on ligament rupture and burst pressures. (author)

  15. Pressurization rate effect on ligament rupture and burst pressures of cracked steam generator tubes

    Energy Technology Data Exchange (ETDEWEB)

    Majumdar, S.; Kasza, K. [Argonne National Laboratory, Nuclear Energy Division, Lemont, Illinois (United States)

    2009-07-01

    The question of whether ligament rupture pressure or unstable burst pressure may vary significantly with pressurization rate at room temperature arose from the results of pressure tests by industry on tubes with machined part-throughwall notches. Slow (quasi-static) and fast 14 MPa/s (2000 psi/s) pressurization rate tests on specimens with nominally the same notch geometry appeared to show a significant effect of the rate of pressurization on the unstable burst pressure. Unfortunately, the slow and fast loading rate tests were conducted following two different test procedures, which could confound the results. The current series of tests were conducted on a variety of specimen geometries using a consistent test procedure to better establish the effect of pressurization rate on ligament rupture and burst pressures. (author)

  16. Investigation of thermodynamic cycle for generic 1200 MW{sub el} pressure channel reactor with nuclear steam superheat

    Energy Technology Data Exchange (ETDEWEB)

    Vincze, A.; Sidawi, K.; Abdullah, R.; Baldock, M.; Saltanov, E.; Pioro, I., E-mail: andrei.vincze@uoit.net, E-mail: khalil.sidawi@uoit.net, E-mail: rand.abdullah@uoit.net, E-mail: matthew.baldock@uoit.net, E-mail: eugene.saltanov@uoit.ca, E-mail: igor.pioro@uoit.ca [Univ. of Ontario Inst. of Tech., Oshawa, ON (Canada)

    2014-07-01

    Current Nuclear Power Plants (NPPs) play a significant role in energy production around the world. All NPPs operating today employ a Rankine steam cycle for the conversion of thermal power to electricity. This paper will examine the steam cycle arrangement an experimental pressure channel reactor using Nuclear Steam Superheat (NSS) and compare it to two advanced reactor designs, the Advanced CANDU Reactor 1000 (ACR-1000) and the Advanced Boiling Water Reactor (ABWR) designs. The thermodynamic cycle layout and thermal efficiencies of the three reactor types will be discussed. (author)

  17. Commercial-Industrial Cleaning, by Pressure-Washing, Hydro-Blasting and UHP-Jetting The Business Operating Model and How-To Manual for 450 Specific Applications

    CERN Document Server

    Maasberg, Wolfgang

    2012-01-01

    Commercial-Industrial Cleaning, by Pressure-Washing, Hydro-Blasting and UHP-Jetting is the first proprietary manual for cleaning and rehabilitation through pressure-washing, hydro-blasting and ultra high pressure water jetting (UHP).   It examines the cleaning, restoration and rehabilitation of statuary and historical structures; manufacturing hardware; and application technologies for residential, commercial and industrial areas, structures and buildings. Commercial-Industrial Cleaning, by Pressure-Washing, Hydro-Blasting and UHP-Jetting contains over 450 applications from agricultural, marine, municipal, food processing, paper-pulp, pharmaceutical and cosmetic, industrial and power generating maintenance areas. It includes gear lists to help readers easily identify the appropriate tooling and equipment for each specific application and industry.   Commercial-Industrial Cleaning, by Pressure-Washing, Hydro-Blasting and UHP-Jetting supplies readers with the tools to create a successful business model for re...

  18. Influence of the loop design of the feedwater- and steam quality in a power plant with pressurized water reactor

    International Nuclear Information System (INIS)

    Bennert, J.; Becher, L.

    1977-01-01

    At nuclear power plants with pressurized water reactors, condensate occurs on the high pressure part of the water-steam circuit, caused by the operation with low steam parameters. The behaviour of the electrolytes which entered into the circuit (solubility, distribution in water and/or steam) shows that these electrolytes (salts) are to be found mainly in the condensate. The insinuated electrolytes are reconcentrated during the common arrangements with 'Small Circuit' - consisting of steam generator, high pressure turbine, water separator, feedwater vessel, and have a negative influence on the feedwater - boiler water - and the steam quality. Remedy is possible by modified arrangements, during which these electrolyte-containing condensates will be treated and traced back into the main circuit. Nevertheless that the efficiency decrease is insignificant and additional efforts are necessary, a change over to these arrangements is recommendable, due to the fact that the feedwater quality, the boiler water quality, the steam quality in front of the turbine, and finally also the operational safety, as well as the availability will be improved. (orig.) [de

  19. Blast from pressurized carbon dioxide released into a vented atmospheric chamber

    Science.gov (United States)

    Hansen, P. M.; Gaathaug, A. V.; Bjerketvedt, D.; Vaagsaether, K.

    2018-03-01

    This study describes the blast from pressurized carbon dioxide (CO2) released from a high-pressure reservoir into an openly vented atmospheric chamber. Small-scale experiments with pure vapor and liquid/vapor mixtures were conducted and compared with simulations. A motivation was to investigate the effects of vent size and liquid content on the peak overpressure and impulse response in the atmospheric chamber. The comparison of vapor-phase CO2 test results with simulations showed good agreement. This numerical code described single-phase gas dynamics inside a closed chamber, but did not model any phase transitions. Hence, the simulations described a vapor-only test into an unvented chamber. Nevertheless, the simulations reproduced the incident shock wave, the shock reflections, and the jet release inside the atmospheric chamber. The rapid phase transition did not contribute to the initial shock strength in the current test geometry. The evaporation rate was too low to contribute to the measured peak overpressure that was in the range of 15-20 kPa. The simulation results produced a calculated peak overpressure of 12 kPa. The liquid tests showed a significantly higher impulse compared to tests with pure vapor. Reducing the vent opening from 0.1 to 0.01 m2 resulted in a slightly higher impulse calculated at 100 ms. The influence of the vent area on the calculated impulse was significant in the vapor-phase tests, but not so clear in the liquid/vapor mixture tests.

  20. Lower blasthole pressures: a means of reducing costs when blasting rocks of low to moderate strength

    Energy Technology Data Exchange (ETDEWEB)

    Hagan, T.N.; Gibson, I.M.

    1988-03-01

    From a purely mechanical viewpoint, each explosive charge should produce a peak blasthole pressure (P/sub b/) that just fails to crush (i.e. pulverise or plastically deform) the rock which surrounds it. Where P/sub b/ exceeds a critical value, some explosion energy is wasted in crushing an annular section of rock immediately around each charge. As a rock's dynamic compressive breaking strain decreases, so should P/sub b/ (Hagan, 1977b). This paper reviews information on, and anticipates the blasting performance of, bulk charges having effective densities which are as low as about 40% of that for ammonium nitrate fuel oil (ANFO). It also outlines the potential advantages of extending the reaction periods of charges, even to the extent that explosive reactions continue after the blasthole wall and stemming have started to move. The paper then proceeds to define situations in which the use of such lower-pressure charges is likely to result in greatest reductions in mining costs. Some methods of applying bulk charges having effective densities in the 0.3-0.8 g cm/sup -3/ range and/or lower reaction rates are suggested. 15 refs., 3 figs.

  1. Investigation of high pressure steaming (HPS) as a thermal treatment for lipid extraction from Chlorella vulgaris.

    Science.gov (United States)

    Aguirre, Ana-Maria; Bassi, Amarjeet

    2014-07-01

    Biofuels from algae are considered a technically viable energy source that overcomes several of the problems present in previous generations of biofuels. In this research high pressure steaming (HPS) was studied as a hydrothermal pre-treatment for extraction of lipids from Chlorella vulgaris, and analysis by response surface methodology allowed finding operational points in terms of target temperature and algae concentration for high lipid and glucose yields. Within the range covered by these experiments the best conditions for high bio-crude yield are temperatures higher than 174°C and low biomass concentrations (<5 g/L). For high glucose yield there are two suitable operational ranges, either low temperatures (<105°C) and low biomass concentrations (<4 g/L); or low temperatures (<105°C) and high biomass concentrations (<110 g/L). High pressure steaming is a good hydrothermal treatment for lipid recovery and does not significantly change the fatty acids profile for the range of temperatures studied. Copyright © 2014 Elsevier Ltd. All rights reserved.

  2. Materials, manufacture and testing of pressurized components of high-power steam power plants

    International Nuclear Information System (INIS)

    Blind, D.; Foehl, J.; Issler, L.; Schellhammer, W.; Sturm, D.; Kussmaul, K.; Heinrich, D.; Meyer, H.J.; Prestel, W.

    1981-01-01

    This is the first German review of materials, production and testing of pressure components of high-capacity steam power plants. The authors have been working in this field for years; their special subject has been the availability and reliability of pressure vessels, in particular in nuclear engineering. Fundamentals are presented as well as the findings obtained at the state Materials Testing Institute in Stuttgart. The material is presented in a well-structured classification; the most recent international findings, especially of the USA, are presented. This is possible due to the close cooperation between the Stuttgart institute and a number of US research institutes. The new subject of fracture mechanics is treated in some detail; its fundamentals are discussed from the American point of view while German considerations - in particular of the Reactor Safety Commission - are taken into account in the field of applications. (orig.) [de

  3. Effect of Ovality on Maximum External Pressure of Helically Coiled Steam Generator Tubes with a Rectangular Wear

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Dong In; Lim, Eun Mo; Huh, Nam Su [Seoul National Univ. of Science and Technology, Seoul (Korea, Republic of); Choi, Shin Beom; Yu, Je Yong; Kim, Ji Ho; Choi, Suhn [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    A structural integrity of steam generator tubes of nuclear power plants is one of crucial parameters for safe operation of nuclear power plants. Thus, many studies have been made to provide engineering methods to assess integrity of defective tubes of commercial nuclear power plants considering its operating environments and defect characteristics. As described above, the geometric and operating conditions of steam generator tubes in integral reactor are significantly different from those of commercial reactor. Therefore, the structural integrity assessment of defective tubes of integral reactor taking into account its own operating conditions and geometric characteristics, i. e., external pressure and helically coiled shape, should be made to demonstrate compliance with the current design criteria. Also, ovality is very specific characteristics of the helically coiled tube because it is occurred during the coiling processes. The wear, occurring from FIV (Flow Induced Vibration) and so on, is main degradation of steam generator tube. In the present study, maximum external pressure of helically coiled steam generator tube with wear is predicted based on the detailed 3-dimensional finite element analysis. As for shape of wear defect, the rectangular shape is considered. In particular, the effect of ovality on the maximum external pressure of helically coiled tubes with rectangular shaped wear is investigated. In the present work, the maximum external pressure of helically coiled steam generator tube with rectangular shaped wear is investigated via detailed 3-D FE analyses. In order to cover a practical range of geometries for defective tube, the variables affecting the maximum external pressure were systematically varied. In particular, the effect of tube ovality on the maximum external pressure is evaluated. It is expected that the present results can be used as a technical backgrounds for establishing a practical structural integrity assessment guideline of

  4. An investigation of nucleating flows of steam in a cascade of turbine blading: Effect of overall pressure ratios

    International Nuclear Information System (INIS)

    Bakhtar, F.; Savage, R.A.

    1993-01-01

    In the course of expansion of steam in turbines the state path crosses the saturation line and the fluid becomes a two-phase mixture. To reproduce turbine nucleating and wet conditions realistically requires a supply of supercooled steam which can be obtained under blow down conditions. An experimental short duration cascade tunnel working on this principle has been constructed. The blade profile studied is that of a typical nozzle The paper is one of a set and describes the surface pressure measurements carried out to investigate the effect of the overall pressure ratio on the performance of the blade

  5. Effect of steam condensation on pressure and temperature under hydrogen jet fire in a vented enclosure

    International Nuclear Information System (INIS)

    Kuznetsov, Mike; Xiao, Jianjun; Travis, Jack

    2017-01-01

    Hydrogen release through leaks due to the LOCA and MCCI accidents and its immediate ignition leads to formation of hydrogen jet fire in a containment of reactor building. An experimental study of hydrogen jet fire in a chamber of 1x1x1 m 3 volume with different vent position, vent areas from 1 to 90 cm 2 and hydrogen mass flow rates from 0.027 to 1.087 g/s were performed in current work. Depending on hydrogen mass flow rate and vent area a well-ventilated or under-ventilated jet fire regime may occur. In the case of relatively small hydrogen release rate and large vent area, relatively stable jet fire behaviour for well-ventilated jet fire leading to over-pressure not more than 0.8 mbar was found. Three different scenarios of under-ventilated jet fire behaviour with self-extinction, re-ignition and external flame leading to relatively high overpressure of 10-100 mbar were found experimentally and numerically. Numerical simulations with GASFLOW-MPI code were performed with/without modelling heat conduction in solid walls, steam condensation, convective heat transfer and thermal radiation. With heat transfer modelling, both initial pressure peak and pressure decay were very well predicted compared to the experimental data. Numerical simulations were then compared with experimental Background Oriented Schlieren (BOS) images obtained to visualize the hydrogen combustion process. Self-extinction and re-ignition events were captured in the numerical simulation as well. An adiabatic case indicates that heat transfer and steam condensation must be included into the combustion model to accurately predict the physical phenomena of turbulent hydrogen jet flames in a vented enclosure. (author)

  6. Steam generator

    International Nuclear Information System (INIS)

    Fenet, J.-C.

    1980-01-01

    Steam generator particularly intended for use in the coolant system of a pressurized water reactor for vaporizing a secondary liquid, generally water, by the primary cooling liquid of the reactor and comprising special arrangements for drying the steam before it leaves the generator [fr

  7. The deterministic prediction of failure of low pressure steam turbine disks

    International Nuclear Information System (INIS)

    Liu, Chun; Macdonald, D.D.

    1993-01-01

    Localized corrosion phenomena, including pitting corrosion, stress corrosion cracking, and corrosion fatigue, are the principal causes of corrosion-induced damage in electric power generating facilities and typically result in more than 50% of the unscheduled outages. Prediction of damage, so that repairs and inspections can be made during scheduled outages, could have an enormous impact on the economics of electric power generation. To date, prediction of corrosion damage has been made on the basis of empirical/statistical methods that have proven to be insufficiently robust and accurate to form the basis for the desired inspection/repair protocol. In this paper, we describe a deterministic method for predicting localized corrosion damage. We have used the method to illustrate how pitting corrosion initiates stress corrosion cracking (SCC) for low pressure steam turbine disks downstream of the Wilson line, where a thin condensed liquid layer exists on the steel disk surfaces. Our calculations show that the SCC initiation and propagation are sensitive to the oxygen content of the steam, the environment in the thin liquid condensed layer, and the stresses that the disk experiences in service

  8. Stress analysis of pressurized water reactor steam generator tube denting phenomena. Interim report

    International Nuclear Information System (INIS)

    Thomas, J.M.; Cipolla, R.C.; Ranjan, G.V.; Derbalian, G.

    1978-07-01

    In some Pressurized Water Reactor (PWR) steam generators, a corrosion product has formed on the carbon steel support plate in the crevice between the tube and support plate. The corrosion product occupies more volume than the original metal; the tube-to-support plate crevice volume is thus consumed with corrosion product, and further corrosive action results in a radially inward force on the tube and a radially outward force on the corroding support plate. This has resulted in indentation (''denting'') of the tube, accompanied by occasional cracking. Large in-plane deformation and cracking of support plates has also been observed in the most severely affected plants along with some serious side effects, such as deformation and cracking of inner row tube U-bends caused by support plate movement. Mechanical aspects of the tube denting phenomena have been studied using analytical models. The models used ranged from closed form analytical solutions to state-of-the-art numerical elastic-plastic computer program for moderate strains. It was found that tube dents, such as those observed in operating steam generators, are associated with yielding of both the tubes and support plates. Also studied were the stresses in tube U-bends caused by support plate flow slot deformation

  9. Influence of upstream stator on rotor flutter stability in a low pressure steam turbine stage

    Energy Technology Data Exchange (ETDEWEB)

    Huang, X.; He, L. [University of Durham (United Kingdom). School of Engineering; Bell, D. [ALSTOM Power Ltd., Rugby (United Kingdom)

    2006-07-01

    Conventional blade flutter prediction is normally based on an isolated blade row model, however, little is known about the influence of adjacent blade rows. In this article, an investigation is presented into the influence of the upstream stator row on the aero-elastic stability of rotor blades in the last stage of a low pressure (LP) steam turbine. The influence of the upstream blade row is computed directly by a time-marching, unsteady, Navier-Stokes flow solver in a stator-rotor coupled computational domain. The three-dimensional flutter solution is obtained, with adequate mesh resolution, in a single passage domain through application of the Fourier-Transform based Shape-Correction method. The capability of this single-passage method is examined through comparison with predictions obtained from a complete annulus model, and the results demonstrate a good level of accuracy, while achieving a speed up factor of 25. The present work shows that the upstream stator blade row can significantly change the aero-elastic behaviour of an LP steam turbine rotor. Caution is, therefore, advised when using an isolated blade row model for blade flutter prediction. The results presented also indicated that the intra-row interaction is of a strong three-dimensional nature. (author)

  10. Steam drums

    International Nuclear Information System (INIS)

    Crowder, R.

    1978-01-01

    Steam drums are described that are suitable for use in steam generating heavy water reactor power stations. They receive a steam/water mixture via riser headers from the reactor core and provide by means of separators and driers steam with typically 0.5% moisture content for driving turbines. The drums are constructed as prestressed concrete pressure vessels in which the failure of one or a few of the prestressing elements does not significantly affect the overall strength of the structure. The concrete also acts as a radiation shield. (U.K.)

  11. Thermodynamic performance analysis and algorithm model of multi-pressure heat recovery steam generators (HRSG) based on heat exchangers layout

    International Nuclear Information System (INIS)

    Feng, Hongcui; Zhong, Wei; Wu, Yanling; Tong, Shuiguang

    2014-01-01

    Highlights: • A general model of multi-pressure HRSG based on heat exchangers layout is built. • The minimum temperature difference is introduced to replace pinch point analysis. • Effects of layout on dual pressure HRSG thermodynamic performances are analyzed. - Abstract: Changes of heat exchangers layout in heat recovery steam generator (HRSG) will modify the amount of waste heat recovered from flue gas; this brings forward a desire for the optimization of the design of HRSG. In this paper the model of multi-pressure HRSG is built, and an instance of a dual pressure HRSG under three different layouts of Taihu Boiler Co., Ltd. is discussed, with specified values of inlet temperature, mass flow rate, composition of flue gas and water/steam parameters as temperature, pressure etc., steam mass flow rate and heat efficiency of different heat exchangers layout of HRSG are analyzed. This analysis is based on the laws of thermodynamics and incorporated into the energy balance equations for the heat exchangers. In the conclusion, the results of the steam mass flow rate, heat efficiency obtained for three heat exchangers layout of HRSGs are compared. The results show that the optimization of heat exchangers layout of HRSGs has a great significance for waste heat recovery and energy conservation

  12. Fuzzy logic control of steam generator water level in pressurized water reactors

    International Nuclear Information System (INIS)

    Kuan, C.C.; Lin, C.; Hsu, C.C.

    1992-01-01

    In this paper a fuzzy logic controller is applied to control the steam generator water level in a pressurized water reactor. The method does not require a detailed mathematical mode of the object to be controlled. The design is based on a set of linguistic rules that were adopted from the human operator's experience. After off-line fuzzy computation, the controller is a lookup table, and thus, real-time control is achieved. Shrink-and-swell phenomena are considered in the linguistic rules, and the simulation results show that their effect is dramatically reduced. The performance of the control system can also be improved by changing the input and output scaling factors, which is convenient for on-line tuning

  13. Simulating the steam generator and the pressurizer of a PWR nuclear power plant

    International Nuclear Information System (INIS)

    De Greef, J.F.

    1985-01-01

    In a PWR nuclear power plant, considered as a power generating device, the steam generator as a subset plays an important role in the generation process, whereas the pressurizer rather acts as a control device for security purposes. Nevertheless, from a thermodynamical point of view, the two subsets behave basically in the same way, so that a common set of basic equations may be suggested to develop for each the proper mathematical simulation model. In this paper the generation of this common set of basic equations is described, from which a specific model for each device is derived. A numerical illustration of the behaviour of the two devices for typical inputs to the derived simulation model is pictured. (author)

  14. Evaluation of cracking in steam generator feedwater piping in pressurized water reactor plants

    International Nuclear Information System (INIS)

    Goldberg, A.; Streit, R.D.

    1981-05-01

    Cracking in feedwater piping was detected near the inlet to steam generators in 15 pressurized water reactor plants. Sections with cracks from nine plants are examined with the objective of identifying the cracking mechanism and assessing various factors that might contribute to this cracking. Using transmission electron microscopy, fatigue striations are observed on replicas of cleaned crack surfaces. Calculations based on the observed striation spacings gave a cyclic stress value of 150 MPa (22 ksi) for one of the major cracks. The direction of crack propagation was invariably related to the piping surface and not to the piping axis. These two factors are consistent with the proposed concept of thermally induced, cyclic, tensile surface stresses and it is concluded that the overriding factor in the cracking problem was the presence of such undocumented cyclic loads

  15. On synthesis and optimization of steam system networks. 3. Pressure drop consideration

    CSIR Research Space (South Africa)

    Price, T

    2010-08-01

    Full Text Available Heat exchanger networks in steam systems are traditionally designed to operate in parallel. Coetzee and Majozi (Ind. Eng. Chem. Res. 2008, 47, 4405-4413) found that by reusing steam condensate within the network the steam flow rate could be reduced...

  16. Pressure loss characteristics of LSTF steam generator heat-transfer tubes. Pressure loss increase due to tube internal instruments

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro

    1994-11-01

    The steam generator of the Large-Scale Test Facility (LSTF) includes 141 heat-transfer U-tubes with different lengths. Six U-tubes among them are furnished with 15 or 17 probe-type instruments (conduction probe with a thermocouple; CPT) protuberant into the primary side of the U-tubes. Other 135 U-tubes are not instrumented. This results in different hydraulic conditions between the instrumented and non-instrumented U-tubes with the same length. A series of pressure loss characteristics tests was conducted at a test apparatus simulating both types of U-tube. The following pressure loss coefficient (K CPT ) was reduced as a function of Reynolds number (Re) from these tests under single-phase water flow conditions. K CPT =0.16 5600≤Re≤52820, K CPT =60.66xRe -0.688 2420≤Re≤5600, K CPT =2.664x10 6 Re -2.06 1371≤Re≤2420. The maximum uncertainty is 22%. By using these results, the total pressure loss coefficients of full length U-tubes were estimated. It is clarified that the total pressure loss of the shortest instrumented U-tube is equivalent to that of the middle-length non-instrumented U-tube and also that a middle-length instrumented U-tube is equivalent to the longest non-instrumented U-tube. Concludingly. it is important to take account of the CPT pressure loss mentioned above in estimation of fluid behavior at the non-instrumented U-tubes either by using the LSTF experiment data from the CPT-installed U-tubes or by using any analytical codes. (author)

  17. Microstructural studies on steam oxidised Zr-2.5%Nb pressure tube under simulated LOCA condition

    International Nuclear Information System (INIS)

    Banerjee, Suparna; Sawarn, Tapan K.; Pandit, K.M.; Anantharaman, S.; Srivastava, D.; Sah, D.N.

    2013-03-01

    Study of the microstructural evolution of Zr-2.5%Nb pressure tube material of Indian Pressurized Heavy Water Reactors (PHWRs) due to steam oxidation at high temperature (in the range 500-1050°C) was carried out on pressure tube coupons. Hydrogen pick up was less than 55 ppm in the samples oxidized at temperatures up to 850°C but high (250-400 ppm) in the samples oxidized in the β phase region (900°C and above). The microstructure of the samples oxidized above the α-Zr/β-Zr transition temperature showed from the surface inwards sequentially the presence of an oxide layer, an underlying oxygen stabilized α-Zr layer and a prior β-Zr phase containing hydride precipitates. An increase in the hardness was observed near the oxide-metal interface in the coupons oxidized above 900°C, due to formation of oxygen stabilized α-Zr layer. Higher hardness was also observed in the base metal in the samples oxidized at 1000 and 1050°C (author)

  18. The deformation of Zircaloy PWR cladding with low internal pressures, under mainly convective cooling by steam

    International Nuclear Information System (INIS)

    Hindle, E.D.; Mann, C.A.; Reynolds, A.E.

    1981-08-01

    Simulated PWR fuel rods clad with Zircaloy-4 were tested under convective steam cooling conditions, by pressurising to 0.69-2.07MPa (100-300lb/in 2 ), then ramping at 10 0 C/s to various temperatures in the region 800-955 0 C and holding until either 600 s elapsed or rupture occurred. The length of cladding strained 33% or more was greatest (about 20 times the original diameter) when the initial internal pressure was 1.38+-0.17 PMa (200+-25lb/in 2 ), and the temperature 885 0 C. It is thought that this results from oxidation strengthening of the surface layers acting as an additional mechanism for stabilising the deformation and/or partial superplastic deformation. To avoid adjacent rods in a fuel assembly touching at any temperature, the pressure would have to be less than about 1MPa (145 1b/in 2 ). If the pressure was 1.38MPa (200lb/in 2 ) then the rods would not swell sufficiently to touch if the temperature did not exceed about 840 0 C. (author)

  19. Flow instability research on steam generator with straight double-walled heat transfer tube for FBR. Pressure drop under high pressure condition

    International Nuclear Information System (INIS)

    Liu, Wei; Tamai, Hidesada; Yoshida, Hiroyuki; Takase, Kazuyuki; Hayafune, Hiroki; Futagami, Satoshi; Kisohara, Naoyuki

    2008-01-01

    For the Steam Generator (SG) with straight double-walled heat transfer tube that used in sodium cooled Faster Breeder Reactor, flow instability is one of the most important items need researching. As the first step of the research, thermal hydraulics experiments were performed under high pressure condition in JAEA with using a straight tube. Pressure drop, heat transfer coefficients and void fraction data were derived. This paper evaluates the pressure drop data with TRAC-BF1 code. The Pffan's correlation for single phase flow and the Martinelli-Nelson's two-phase flow multiplier are found can be well predicted the present pressure drop data under high pressure condition. (author)

  20. Performance of a pilot-scale, steam-blown, pressurized fluidized bed biomass gasifier

    Science.gov (United States)

    Sweeney, Daniel Joseph

    With the discovery of vast fossil resources, and the subsequent development of the fossil fuel and petrochemical industry, the role of biomass-based products has declined. However, concerns about the finite and decreasing amount of fossil and mineral resources, in addition to health and climate impacts of fossil resource use, have elevated interest in innovative methods for converting renewable biomass resources into products that fit our modern lifestyle. Thermal conversion through gasification is an appealing method for utilizing biomass due to its operability using a wide variety of feedstocks at a wide range of scales, the product has a variety of uses (e.g., transportation fuel production, electricity production, chemicals synthesis), and in many cases, results in significantly lower greenhouse gas emissions. In spite of the advantages of gasification, several technical hurdles have hindered its commercial development. A number of studies have focused on laboratory-scale and atmospheric biomass gasification. However, few studies have reported on pilot-scale, woody biomass gasification under pressurized conditions. The purpose of this research is an assessment of the performance of a pilot-scale, steam-blown, pressurized fluidized bed biomass gasifier. The 200 kWth fluidized bed gasifier is capable of operation using solid feedstocks at feedrates up to 65 lb/hr, bed temperatures up to 1600°F, and pressures up to 8 atm. Gasifier performance was assessed under various temperatures, pressure, and feedstock (untreated woody biomass, dark and medium torrefied biomass) conditions by measuring product gas yield and composition, residue (e.g., tar and char) production, and mass and energy conversion efficiencies. Elevated temperature and pressure, and feedstock pretreatment were shown to have a significant influence on gasifier operability, tar production, carbon conversion, and process efficiency. High-pressure and temperature gasification of dark torrefied biomass

  1. Steam condenser

    International Nuclear Information System (INIS)

    Masuda, Fujio

    1980-01-01

    Purpose: To enable safe steam condensation by providing steam condensation blades at the end of a pipe. Constitution: When high temperature high pressure steam flows into a vent pipe having an opening under water in a pool or an exhaust pipe or the like for a main steam eacape safety valve, non-condensable gas filled beforehand in the steam exhaust pipe is compressed, and discharged into the water in the pool. The non-condensable gas thus discharged from the steam exhaust pipe is introduced into the interior of the hollow steam condensing blades, is then suitably expanded, and thereafter exhausted from a number of exhaust holes into the water in the pool. In this manner, the non-condensable gas thus discharged is not directly introduced into the water in the pool, but is suitable expanded in the space of the steam condensing blades to suppress extreme over-compression and over-expansion of the gas so as to prevent unstable pressure vibration. (Yoshihara, H.)

  2. AMPTRACT: an algebraic model for computing pressure tube circumferential and steam temperature transients under stratified channel coolant conditions

    International Nuclear Information System (INIS)

    Gulshani, P.; So, C.B.

    1986-10-01

    In a number of postulated accident scenarios in a CANDU reactor, some of the horizontal fuel channels are predicted to experience periods of stratified channel coolant condition which can lead to a circumferential temperature gradient around the pressure tube. To study pressure tube strain and integrity under stratified flow channel conditions, it is, necessary to determine the pressure tube circumferential temperature distribution. This paper presents an algebraic model, called AMPTRACT (Algebraic Model for Pressure Tube TRAnsient Circumferential Temperature), developed to give the transient temperature distribution in a closed form. AMPTRACT models the following modes of heat transfer: radiation from the outermost elements to the pressure tube and from the pressure to calandria tube, convection between the fuel elements and the pressure tube and superheated steam, and circumferential conduction from the exposed to submerged part of the pressure tube. An iterative procedure is used to solve the mass and energy equations in closed form for axial steam and fuel-sheath transient temperature distributions. The one-dimensional conduction equation is then solved to obtain the pressure tube circumferential transient temperature distribution in a cosine series expansion. In the limit of large times and in the absence of convection and radiation to the calandria tube, the predicted pressure tube temperature distribution reduces identically to a parabolic profile. In this limit, however, radiation cannot be ignored because the temperatures are generally high. Convection and radiation tend to flatten the parabolic distribution

  3. Main Steam Line Break Mass/Energy and Pressure/Temperature Analysis for the Environmental Qualification

    International Nuclear Information System (INIS)

    Park, Yong-Chan; Song, Dong-Soo; Jun, Hwang-Yong

    2006-01-01

    The Main steam line break(MSLB) occurring inside a reactor containment structure may result in significant releases of high energy fluid to the containment, possibly result in high containment pressure and temperature. The MSLB accident, along with the Loss Of Coolant Accident, is a design basis accident for determining the peak containment pressure and temperature. The analysis for a MSLB for inside containment should be performed to justify the structural integrity and equipment qualification in accordance with revision 1 of Reg. Guide 1.89. Rev1(1984), which is also required as part of obtaining the extended operating license for WestingHouse(WH) 3-Loops Nuclear Power Plant(NPP). Now, the WH NPP has been performed power uprating. Therefore, all initial conditions, setpoints and uncertainties were considered with MSLB analysis for environment qualification(EQ). The transient was analyzed to determine the worst set of mass and energy releases that impact the EQ aspects of safety related equipment inside containment. The most limiting single failure in this event was determined by a sensitivity study. The MSLB event was analyzed for a full set of power conditions and break sizes

  4. Control of the thermostressed state of low-pressure cylinder rotors for power steam turbines

    International Nuclear Information System (INIS)

    Lejzerovich, A.Sh.

    1980-01-01

    The principle arrangement of an analog device for operation control of the low pressure cylinder (LPC) heating at large steam turbine start-up has been developed. Different forms of representation of the thermal conductivity equation used for realization by means of analog models are analized. Presented are the results of calculating the heating indices for the welded rotor of LPC during the turbine start-up from a cold state and the curves of temperature distribution in the disc of the first sections of welded LPC rotor at start-up from a cold state and in a steady-state regime. The results obtained show that in the process of start-up the error of the temperature difference DELTAt determination according to the suggested scheme does not exceed 10 deg C. After achieving the maximum of DELTAt in the process of the rotor temperature field flattening, this error increases and constitutes 32 deg C in steady-state regime, mainly, due to the error of temperature determination on the rotation axis in controlled cross section. As far as the control for the LPC rotor heating is necessary only during start-up and the requirements for its accuracy are not equivalent, therefore, for all regimes, representativity and accuracy of control provided by the accepted calculation scheme is quite satisfactory

  5. Denting of Inconel 600 steam generator tubes in pressurized water reactors

    International Nuclear Information System (INIS)

    Van Rooyen, D.; Weeks, J.R.

    1976-10-01

    Rapid, localized corrosion of carbon steel tube support plates (TSP) has led to cases of denting of steam generator tubes, due to the pressure of corrosion products formed in crevices between the tubes and TSP holes. The corrosion product is mainly magnetite (Fe 3 O 4 ), formed in ''run-away'' fashion as a result of local chemistry changes when an extended operation with phosphate (PO 4 ) treatment of the secondary coolant is followed by an all volatile treatment (AVT). The rate of the ''run-away'' magnetite formation, and therefore, the extent of damage will probably vary with the amounts of the harmful chemicals present and with temperature. Leaky condensers are felt to be responsible for the presence of Cl - ions, and for the observation that denting is more extensive in plants with salt water cooled condensers. It is possible that thermal cycles assist the denting process, both by mechanical and chemical ratchetting mechanisms. Recommendations are presented concerning the continued operation of plants with observed denting

  6. Expandable antivibration bar for heat transfer tubes of a pressurized water reactor steam generator

    International Nuclear Information System (INIS)

    Appleman, R.H.

    1987-01-01

    This patent describes a pressurized water reactor steam generator having spaced rows of heat transfer tubes through which primary coolant from the reactor flows, the tubes being of a U-shaped design, with the U-bend portions of the U-shaped tubes stabilized by antivibration bars. The improvement described here comprises expandable antivibration bars for stabilizing the U-bend portions of the U-shaped tubes, the expandable bars having a pair of adjustable rods, formed from a pair of rod sections affixed to a connector, one rod section of each of the pair of rod sections having a plurality of protrusions. Each of the protrusions has slidable surfaces thereon. The other rod section of each of the pair of rod sections has indentations, each of the indentations having slidable surfaces thereon complementary to the sliding surfaces of the protrusions, such that the rods are expandable from a first cross-sectional width less than the spacing between two adjacent rows of the tubes, to a second cross-sectional width greater than the first cross-sectional width. The expanded rods are adapted to contact tubes of the two adjacent rows of the tubes

  7. Experimental study of heat transfer and pressure drop characteristics of air/water and air-steam/water heat exchange in a polymer compact heat exchanger

    NARCIS (Netherlands)

    Cheng, L.; Geld, van der C.W.M.

    2005-01-01

    Experiments of heat transfer and pressure drop in a polymer compact heat exchanger made of PolyVinyliDene-Fluoride were conducted under various conditions for air/water heat exchange and air-steam/water heat exchange, respectively. The overall heat transfer coefficients of air-steam/water heat

  8. Improved algorithm based on equivalent enthalpy drop method of pressurized water reactor nuclear steam turbine

    International Nuclear Information System (INIS)

    Wang Hu; Qi Guangcai; Li Shaohua; Li Changjian

    2011-01-01

    Because it is difficulty to accurately determine the extraction steam turbine enthalpy and the exhaust enthalpy, the calculated result from the conventional equivalent enthalpy drop method of PWR nuclear steam turbine is not accurate. This paper presents the improved algorithm on the equivalent enthalpy drop method of PWR nuclear steam turbine to solve this problem and takes the secondary circuit thermal system calculation of 1000 MW PWR as an example. The results show that, comparing with the design value, the error of actual thermal efficiency of the steam turbine cycle obtained by the improved algorithm is within the allowable range. Since the improved method is based on the isentropic expansion process, the extraction steam turbine enthalpy and the exhaust enthalpy can be determined accurately, which is more reasonable and accurate compared to the traditional equivalent enthalpy drop method. (authors)

  9. Experimental modal analysis of the steam inlet pipe to the Chooz B1 high pressure turbine

    International Nuclear Information System (INIS)

    Guihot, O.; Anne, J.P.; Chartain, G.; Le Pironnec, D.

    1993-05-01

    This report presents the results of the modal analysis carried out on one of the steam inlet pipe of the high pressure turbine of the Chooz B1 power plant. This experimental analysis is made within the frame of the research and development project ''dynamical, acoustical and aerodynamical behaviour of the turbogenerator N4''. This research program provides amongst others, numerical studies with the software CIRCUS and ASTER, in order to verify the dynamical behaviour of the designed inlet pipe. The numerical models will be updated from results of the experimental modal analysis to improve the numerical representation of this pipe. All the identified modes in the frequency band [5.2000] Hz are presented in the report. The modal characteristics of the main modes are detailed. Further analysis have been made, in order ease the updating of the numerical models. They consisted in an analysis of the evolution of the dynamical behaviour due to a change of the boundary conditions of the inlet valve frame on one hand and resulting from the presence of an additional mass on the pipe, at the level of the middle flange, on the other hand. The analysis made in low frequency range shows that the pipe is thoroughly embedded in the frame of the high pressure turbine. On the other hand, the boundary conditions on the inlet valve frame are more difficult to determine, because the dynamical behaviour of the valve frame and the upper pipe can not be uncoupled from the considered pipe. The main shell modes of ranks 2, 3 and 4 have been very accurately identified. The most relevant modes to update the numerical models are given. (authors). 48 figs., 18 tabs., 4 refs

  10. Application of Confined Blasting in Water-Filled Deep Holes to Control Strong Rock Pressure in Hard Rock Mines

    Directory of Open Access Journals (Sweden)

    Jingxuan Yang

    2017-11-01

    Full Text Available In extra-thick coal seams, mining operations can lead to large-scale disturbances, complex overburden structures, and frequent and strong strata behavior in the stope, which are serious threats to mine safety. This study analyzed the overburden structure and strata behavior and proposed the technique of confined blasting in water-filled deep holes as a measure to prevent strong rock pressure. It found that there are two primary reasons for the high effectiveness of the proposed technique in presplitting hard coal and rock. First, the fracture water enables much more efficient transfer of dynamic load due to its incompressibility. Second, the subsequent expansion of water can further split the rock by compression. A mechanical model was used to reveal how the process of confined blasting in water-filled deep holes presplit roof. Moreover, practical implementation of this technique was found to improve the structure of hard, thick roof and prevent strong rock pressure, demonstrating its effectiveness in roof control.

  11. The development of a neutralizing amines based reagent for maintaining the water chemistry for medium and high pressures steam boilers

    Science.gov (United States)

    Butakova, M. V.; Orlov, K. A.; Guseva, O. V.

    2017-11-01

    An overview of the development for neutralizing amine based reagent for water chemistry of steam boilers for medium and high pressures was given. Total values of the neutralization constants and the distribution coefficients of the compositions selected as a main criteria for reagent composition. Experimental results of using this new reagent for water chemistry in HRSG of power plant in oil-production company are discussed.

  12. Pressure waves transient occurred in the steam generators feedwater lines of the Atucha-1 Nuclear Power Plant

    International Nuclear Information System (INIS)

    Balino, J.L.; Carrica, P.M.; Larreteguy, A.E.

    1993-01-01

    The pressure transient occurred at Atucha I Nuclear Power Plant in March 1990 is simulated. The transient was due to the fast closure of a flow control valve at the steam generators feedwater lines. The system was modelled, including the actuation of the relief valves. The minimum closure time for no actuation of the relief valves and the evolution of the velocity and piezo metric head for different cases were calculated. (author)

  13. High-temperature deformation and rupture behavior of internally-pressurized Zircaloy-4 cladding in vacuum and steam enivronments

    International Nuclear Information System (INIS)

    Chung, H.M.; Garde, A.M.; Kassner, T.F.

    1977-01-01

    The high-temperature diametral expansion and rupture behavior of Zircaloy-4 fuel-cladding tubes have been investigated in vacuum and steam environments under transient-heating conditions that are of interest in hypothetical loss-of-coolant accident situations in light-water reactors. The effects of internal pressure, heating rate, axial constraint, and localized temperature nonuniformities in the cladding on the maximum circumferential strain have been determined for burst temperatures between approximately 650 and 1350 0 C

  14. An Isothermal Steam Expander for an Industrial Steam Supplying System

    Directory of Open Access Journals (Sweden)

    Chen-Kuang Lin

    2015-01-01

    Full Text Available Steam is an essential medium used in the industrial process. To ensure steam quality, small and middle scale boilers are often adopted. However, because a higher steam pressure (compared to the necessary steam pressure is generated, the boiler’s steam pressure will be reduced via a pressure regulator before the steam is directed through the process. Unfortunately, pressure is somewhat wasted during the reducing process. Therefore, in order to promote energy efficiency, a pressure regulator is replaced by a steam expander. With this steam expander, the pressure will be transformed into mechanical energy and extracted during the expansion process. A new type of isothermal steam expander for an industrial steam supplying system will be presented in the paper. The isothermal steam expander will improve the energy efficiency of a traditional steam expander by replacing the isentropic process with an isothermal expansion process. With this, steam condensation will decrease, energy will increase, and steam quality will be improved. Moreover, the mathematical model of the isothermal steam expander will be established by using the Schmidt theory, the same principle used to analyze Stirling engines. Consequently, by verifying the correctness of the theoretical model for the isothermal steam expander using experimental data, a prototype of 100 c.c. isothermal steam expander is constructed.

  15. Welding repair of the high-intermediate pressure steam casings made of Cr-Mo and Cr-Mo-V steel

    International Nuclear Information System (INIS)

    Mazur, Z.; Cristalinas, V.; Kubiak, J.

    1996-01-01

    An analysis of typical failure causes and their location at high-intermediate pressure steam turbine casing, and weldability analysis of the Cr-Mo and Cr-Mo-V steels, is carried out. basing on the steam turbine of 158 MW capacity, the internal high pressure casing failures and development of in situ repair welding technology is described. After repair, the casing was put back into service

  16. Steam turbine installations

    International Nuclear Information System (INIS)

    Bainbridge, A.

    1976-01-01

    The object of the arrangement described is to enable raising steam for driving steam turbines in a way suited to operating with liquid metals, such as Na, as heat transfer medium. A preheated water feed, in heat transfer relationship with the liquid metals, is passed through evaporator and superheater stages, and the superheated steam is supplied to the highest pressure stage of the steam turbine arrangement. Steam extracted intermediate the evaporator and superheater stages is employed to provide reheat for the lower pressure stage of the steam turbine. Only a major portion of the preheated water feed may be evaporated and this portion separated and supplied to the superheater stage. The feature of 'steam to steam' reheat avoids a second liquid metal heat transfer and hence represents a simplification. It also reduces the hazard associated with possible steam-liquid metal contact. (U.K.)

  17. Field test of two high-pressure, direct-contact downhole steam generators. Volume I. Air/diesel system

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, B.W.

    1983-05-01

    As a part of the Project DEEP STEAM to develop technology to more efficiently utilize steam for the recovery of heavy oil from deep reservoirs, a field test of a downhole steam generator (DSG) was performed. The DSG burned No. 2 diesel fuel in air and was a direct-contact, high pressure device which mixed the steam with the combustion products and injected the resulting mixture directly into the oil reservoir. The objectives of the test program included demonstration of long-term operation of a DSG, development of operational methods, assessment of the effects of the steam/combustion gases on the reservoir and comparison of this air/diesel DSG with an adjacent oxygen/diesel direct contact generator. Downhole operation of the air/diesel DSG was started in June 1981 and was terminated in late February 1982. During this period two units were placed downhole with the first operating for about 20 days. It was removed, the support systems were slightly modified, and the second one was operated for 106 days. During this latter interval the generator operated for 70% of the time with surface air compressor problems the primary source of the down time. Thermal contact, as evidenced by a temperature increase in the production well casing gases, and an oil production increase were measured in one of the four wells in the air/diesel pattern. Reservoir scrubbing of carbon monoxide was observed, but no conclusive data on scrubbing of SO/sub x/ and NO/sub x/ were obtained. Corrosion of the DSG combustor walls and some other parts of the downhole package were noted. Metallurgical studies have been completed and recommendations made for other materials that are expected to better withstand the downhole combustion environment. 39 figures, 8 tables.

  18. Field test of two high-pressure direct-contact downhole steam generators. Volume II. Oxygen/diesel system

    Energy Technology Data Exchange (ETDEWEB)

    Moreno, J.B.

    1983-07-01

    A field test of an oxygen/diesel fuel, direct contact steam generator has been completed. The field test, which was a part of Project DEEP STEAM and was sponsored by the US Department of Energy, involved the thermal stimulation of a well pattern in the Tar Zone of the Wilmington Oil Field. The activity was carried out in cooperation with the City of Long Beach and the Long Beach Oil Development Company. The steam generator was operated at ground level, with the steam and combustion products delivered to the reservoir through 2022 feet of calcium-silicate insulated tubing. The objectives of the test included demonstrations of safety, operational ease, reliability and lifetime; investigations of reservoir response, environmental impact, and economics; and comparison of those points with a second generator that used air rather than oxygen. The test was extensively instrumented to provide the required data. Excluding interruptions not attributable to the oxygen/diesel system, steam was injected 78% of the time. System lifetime was limited by the combustor, which required some parts replacement every 2 to 3 weeks. For the conditions of this particular test, the use of trucked-in LOX resulted in liess expense than did the production of the equivalent amount of high pressure air using on site compressors. No statistically significant production change in the eight-acre oxygen system well pattern occurred during the test, nor were any adverse effects on the reservoir character detected. Gas analyses during the field test showed very low levels of SOX (less than or equal to 1 ppM) in the generator gaseous effluent. The SOX and NOX data did not permit any conclusion to be drawn regarding reservoir scrubbing. Appreciable levels of CO (less than or equal to 5%) were measured at the generator, and in this case produced-gas analyses showed evidence of significant gas scrubbing. 64 figures, 10 tables.

  19. In-service diagnostic systems of steam generators, pressurizers and other components of WWER type nuclear power plants

    International Nuclear Information System (INIS)

    Matal, O.

    1988-01-01

    A detailed description is presented of the systems of vibration inspections and systems of determining residual service life, implemented as in-service diagnostic systems for steam generators and pressurizers at the Dukovany nuclear power plant. Low temperature accelerometers of the KD or KS type and high temperature accelerometers CA 91 are used as vibration sensors. In the system of vibration inspection a total of 64 vibration measuring chains of Czechoslovak make and design are installed in the power plant. Systems are being built for determining residual service life which consist of 75 special chains for heat monitoring with thermocouples installed on selected assemblies of the steam generators and the pressurizers serving to monitor and evaluate heat stress. Also included in the system for determining residual service life are 16 routes for water withdrawal from steam generators. Their purpose is to make in-service determinations of places of biggest concentrations of impurities in secondary water, to determine the biggest local chemical exposure of primary collector and heat exchange tube materials and to optimize the size and place of leachate withdrawal. (Z.M.). 2 figs., 2 tabs., 15 refs

  20. Thermal-hydraulics of wave propagation and pressure distribution under hypothetical steam explosion conditions in the ANS reactor

    Energy Technology Data Exchange (ETDEWEB)

    Taleyarkhan, R.P.; Georgevich, V.; N-Valenit, S.; Kim, S.H. [Oak Ridge National Lab., TN (United States)

    1995-09-01

    This paper describes salient aspects of the modeling and analysis framework for evaluation of dynamic loads, wave propagation, and pressure distributions (under hypothetical steam explosion conditions) around key structural boundaries of the Advanced Neutron Source (ANS) reactor core region. A staged approach was followed, using simple thermodynamic models for bounding loads and the CTH code for evaluating realistic estimates in a staged multidimensional framework. Effects of nodalization, melt dispersal into coolant during explosion, single versus multidirectional dissipation, energy level of melt, and rate of energy deposition into coolant were studied. The importance of capturing multidimensional effects that simultaneously account for fluid-structural interactions was demonstrated. As opposed to using bounding loads from thermodynamic evaluations, it was revealed that the ANS reactor system will not be vulnerable to vertically generated missiles that threaten containment if realistic estimates of energetics are used (from CTH calculations for thermally generated steam explosions without significant aluminum ignition).

  1. Modelling and simulation of the steam line, the high and low pressure turbines and the pressure regulator for the SUN-RAH nucleo electric university simulator

    International Nuclear Information System (INIS)

    Lopez R, A.

    2003-01-01

    In the following article the development of a simulator that allows to represent the dynamics of the following systems: steam line, nozzle, vapor separator, reheater, high pressure turbine, low pressure turbine, power generator and the pressure regulator of a nucleo electric power station. We start from the supposition that this plant will be modeled from a nuclear reactor type BWR (Boiling Water Reactor), using models of reduced order that represent the more important dynamic variables of the physical processes that happen along the steam line until the one generator. To be able to carry out the simulation in real time the Mat lab mathematical modeling software is used, as well as the specific simulation tool Simulink. It is necessary to point out that the platform on which the one is executed the simulator is the Windows operating system, to allow the intuitive use that only this operating system offers. The above-mentioned obeys to that the objective of the simulator it is to help the user to understand some of the dynamic phenomena that are present in the systems of a nuclear plant, and to provide a tool of analysis and measurement of variables to predict the desirable behavior of the same ones. The model of a pressure controller for the steam lines, the high pressure turbine and the low pressure turbine is also presented that it will be the one in charge of regulating the demand of the system according to the characteristics and critic restrictions of safety and control, assigned according to those wanted parameters of performance of this system inside the nucleo electric plant. This simulator is totally well defined and it is part of the University student nucleo electric simulator with Boiling Water Reactor (SUN-RAH), an integral project and of greater capacity. (Author)

  2. A study on emergency response guideline during the loss of steam generator secondary heat sink in pressurizer water reactor

    International Nuclear Information System (INIS)

    Yoon, D. J.; Lee, J. Y.; Song, D. S.

    1999-01-01

    A loss of secondary heat sink can occur as a result of several different initiating events, which are a loss of main feedwater during power operation, a loss of off-site power, or any other scenario for which main feedwater is isolated or lost. At this point the opening and closing of the PORV or safety valves will result in a loss of RCS inventory similar in nature to a small break loss of coolant accident. If operator action is not taken, the pressurizer PORV or safety valves will continue to cycle open and closed at the valve setpoint pressure removing RCS inventory and a limited amount of core decay heat until eventually enough inventory will be lost to result in core uncovery. We conclude that a requirement to successfully initiate bleed and feed on steam generator dryout, without any significant core uncovery expected to occur, is that the PORV flow to power ratio must exceed 140 (lbm/hr)/Mwt. For all plants whose PORV capacity is less than 140 (lbm/hr)/Mwt, since symptoms of SG dryout cannot be used to initiate bleed and feed, increasing RCS pressure and temperature or pressure greater than 2335 psig cannot be used. The only alternative symptom available is SG narrow range level. Since Kori 1,2,3 and 4' PORV capacity is more than the criteria, the bleed and feed operation can be initiated at steam generator dryout

  3. Steam generator with U-tube bank arranged within an oblong pressure vessel

    International Nuclear Information System (INIS)

    Beckmann, G.; Fritz, K.

    1976-01-01

    This steam generator equipped with a U-tube bundle differs substantially from standard types because of its operational condition. The boiler is at a tilt of 45 0 , the piping base, the inlet, and the outlet for the primary medium are arranged at the top. This improves the heat flow of the secondary medium within the boiler. The steam room placed near the piping base is enlarged on the hot side of the U-tube bundle due to the tilt of the water level, allowing drying and overheating of the steam without additional mounting of water separators and special overheaters. The additional space obtained by this construction is estimated at 6%. (FW) [de

  4. 900 MW CP1 nuclear steam turbine retrofit thermal effects on low pressure diaphragms

    International Nuclear Information System (INIS)

    Buguin, A.; Gruau, P.; Lamarque, F.; Huggett, J.

    2015-01-01

    The steam turbines of the Koeberg units 1 and 2 operated by ESKOM in South Africa have been retrofitted in order to mitigate the generic problems of stress corrosion cracking of the original shrunk-on disk rotor design. As already done in Belgium and France, the implementation of welded rotors improves the turbine reliability and availability. Moreover, the new technology implemented associated with a new steam path allows a significant performance improvement. With a wealth of experience in CP1 retrofit, ALSTOM has put in place new technical features in the steam path in order to further improve the heat rate. Among them, steam balance holes drilled in the rotor disks have exacerbated the thermal sensitivity of the LP diaphragms. During the commissioning of the Unit 1 LP turbines following the retrofit, the load increase led to unacceptable vibrations. An investigation program was launched to determine the root causes of the problem. This paper presents the findings following the turbine inspection, as well as the recommendations and modifications to allow a smooth return to service of the unit. In addition, the results of the root cause analysis of the vibration incident are explained. Based on finite element calculations and site measurements, ALSTOM has established that the diaphragm thermal behavior, intensified by the steam balance holes, has led to radial rubbing. It was also established that the phenomena had no effect on the diaphragms mechanical integrity. Design changes have been proposed to ensure a safe and reliable long term operation of the units. These modifications have been successfully implemented onto the Koeberg Unit 2 Nuclear Steam Turbine commissioned in November 2012. (authors)

  5. Initial pressure spike and its propagation phenomena in sodium-water reaction tests for MONJU steam generators

    International Nuclear Information System (INIS)

    Sato, M.; Hiroi, H.; Tanaka, N.; Hori, M.

    1977-01-01

    With the objective of demonstrating the safe design of steam generators for prototype LMFBR MONJU against the postulated large-leak accident, a number of large-leak sodium-water reaction tests have been conducted using the SWAT-1 and SWAT-3 rigs. Investigation of the potential effects of pressure load on the system is one of the major concerns in these tests. This paper reports the behavior of initial pressure spike in the reaction vessel, its propagation phenomena to the simulated secondary cooling system, and the comparisons with the computer code for one-dimensional pressure wave propagation problems. Both rigs used are the scaled-down models of the helically coiled steam generators of MONJU. The SWAT-1 rig is a simplified model and consists of a reaction vessel (1/8 scale of MONJU evaporator with 0.4 m dia. and 2.5 m height) and a pressure relief system i.e., a pressure relief line and a reaction products tank. On the other hand, the SWAT-3 rig is a 1/2.5 scale of MONJU SG system and consists of an evaporator (reaction vessel with 1.3 m dia. and 6.35 m height), a superheater, an intermediate heat exchanger (IHX), a piping system simulating the secondary cooling circuit and a pressure relief system. The both water injection systems consist of a water injection line with a rupture disk installed in front of injection hole and an electrically heated water tank. Choice of water injection rates in the scaled-down models is made based on the method of iso-velocity modeling. Test results indicated that the characteristics of the initial pressure spike are dominated by those of initial water injection which are controlled by the conditions of water heater and the size of water injection hole, etc

  6. Basic survey project for Joint Implementation, etc. Blast furnace top pressure recovery turbine (TRT) project (Panzhihua Iron and Steel (Group) Company, People's Republic of China)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-03-01

    For the purpose of reducing greenhouse effect gas emissions, a feasibility study was conducted on the energy conservation at Panzhihua Iron and Steel Company in Sichuan Province, China. In this project, the pressure energy of blast furnace is to be recovered in the form of electric power by installing the blast furnace top pressure recovery turbine (TRT). In the project, the pressure-reducing valve was removed, and the dry dust collector/dry TRT were installed to make the scale of electric power production largest. A model of TRT was installed at No. 4 blast furnace and is now in operation. In this project, TRTs are to be installed at Nos. 1, 2 and 3 blast furnaces. As a result of the study, the investment totaled 5.46 billion yen. The capacity of power generation by TRT is 16,890 kW, and the generated output is 137,822 MWh/y. Moreover, the amount of energy conservation is 36,467 toe/y, and the reduction amount of greenhouse effect gas emissions is 112,830 CO2-ton/y. The term of investment recovery is 8.3 years. The effect of reduction in greenhouse effect gas emissions is 20.66 CO2-ton/y/million yen. (NEDO)

  7. Thermo-hydraulic behavior of saturated steam-water mixture in pressure vessel during injection of cold water

    International Nuclear Information System (INIS)

    Aya, Izuo; Kobayashi, Michiyuki; Inasaka, Fujio; Nariai, Hideki.

    1983-01-01

    The thermo-hydraulic behavior of saturated steam water mixture in a pressure vessel during injection of cold water was experimentally investigated with the Facility for Mixing Effect of Emergency Core Cooling Water. The dimensions of the pressure vessel used in the experiments were 284mm ID and 1,971mm height. 11 experiments were conducted without blowdown in order to comprehend the basic process excluding the effect of blowdown at injection of cold water. The initial pressure and water level, the injection flow rate and the size of injection nozzle were chosen as experimental parameters. Temperatures and void fractions at 6 elevations as well as pressure in the pressure vessel were measured, and new data especially on the pressure undershoot just after the initation of water injection and the vertical distribution of temperature and void fraction were gotten. The transients of pressure, average temperature and void fraction were caluculated using single-volume analysis code BLODAC-1V which is based on thermal equilibrium and so-called bubble gradient model. Some input parameters included in the analysis code were evaluated through the comparison of analysis with experimental data. Moreover, the observed pressure undershoot which is evaluated to be induced by a time lag of vapourization in water due to thermal nonequilibrium, was also discussed with the aid of another simple analysis model. (author)

  8. Effects of radio frequency and high pressure steam sterilisation on the colour and flavour of prepared Nostoc sphaeroides.

    Science.gov (United States)

    Xu, Jicheng; Zhang, Min; An, Yanjun; Roknul, Azam Sm; Adhikari, Benu

    2018-03-01

    Nostoc sphaeroides has been used as a highly effective herbal medicine and dietary supplement for thousands of years. The desired dark green colour of fresh N. sphaeroides is converted into an undesirable dark brown during conventional high pressure (HP) steam sterilisation. Radio frequency (RF) sterilisation technology was used in this study to determine its effectiveness in sterilising N. sphaeroides and to achieve better preservation of natural colour and desirable flavour. Sterilisation was carried out using a 6 kW, 27 MHz RF instrument for 10, 20 and 30 min. The degree of microbial kill and the effects of RF sterilisation on colour and flavour were determined and compared with those obtained from HP steam (121 °C, 30 min) sterilisation. The effects of RF sterilisation on colour and flavour (measured using electronic nose) parameters were significantly lower than that in HP steam sterilisation. The RF sterilisation carried out for 20 min achieved logarithmic reduction of bacterial population and met China's national standard while preserving the colour and flavour better. Results of the present study indicated that application of RF sterilisation would improve the quality of sterilised N. sphaeroides and broaden its application in the food and health food industries. © 2017 Society of Chemical Industry. © 2017 Society of Chemical Industry.

  9. Burst protection device for largely cylindrical steam raising units, preferably of pressurized water nuclear power stations

    International Nuclear Information System (INIS)

    Mutzl, J.

    1978-01-01

    This burst protection device controls forces to be expected in an accident by resolving them into axial (vertical) and radial (horizontal) components, which are taken by a large number of elements stressed in tension. The steam raising unit is surrounded by a containment, but remains easily accessible. The containment consists of a steel jacket, lid and floor. Several cylindrical sections above one another form the steel jacket, which surrounds the steam raising unit with an intermediate insulating layer of concrete. The insulating concrete cylinder is of several times the thickness of the steel jacket, and also consists of cylindrical sections. An outer supporting ring for the lid and floor of the containment have outside diameters which project beyond the jacket. Prestressed circumferential vertical tension ropes between the supporting ring and floor take any additional tensional forces. The lid is domed with downward curvature towards the upper boiler dome. Internal bursting forces produce compressive stresses in the lid, which thus pass along its outside diameter into the surrounding ring. The lid, which is devided along one diameter, makes dismantling and access to the boiler easy even with a central steam pipe going upwards. The floor of the burst protection is also the floor of the steam raising unit. It is of several times the thickness of the tube floor, which, with its spacing above the floor forms the usual inlet and outlet space for the reactor cooling water. The main coolant pump installed there is driven by an external motor through a floor penetration. (HP) [de

  10. Continuous ultrasonic waves to detect steam bubbles in water under high pressure

    Energy Technology Data Exchange (ETDEWEB)

    Hulshof, H J.M.; Schurink, F

    1985-01-01

    Steam in the recirculation circuit of boilers may lead to unacceptable high thermal loads on the evaporator tubes. The ability to detect steam in the recirculation circuit during process transients is therefore important. A simple detector using continuous ultrasonic waves and able to detect bubbles in water contained in steel tubes is described in this paper. The variation of the transmitted wave caused by the bubbles was determined by demodulation. The results have met the objectives set for cold water with air bubbles. A clear indication of the presence of steam bubbles was found in fast-flowing hot water in a steel tube with a diameter of 60 mm. A change in the low-frequency region of the modulation was the only indication of the presence of steam bubbles in the large-diameter downcomer of the water-separator drum of a boiler in an electrical power plant. Possible causes of the differences in the results obtained are discussed on the basis of differences in bubble sizes and in focusing and reflection of the ultrasonic waves. (orig.). 11 refs.; 10 figs.

  11. Continuous ultrasonic waves to detect steam bubbles in water under high pressure

    International Nuclear Information System (INIS)

    Hulshof, H.J.M.; Schurink, F.

    1985-01-01

    Steam in the recirculation circuit of boilers may lead to unacceptable high thermal loads on the evaporator tubes. The ability to detect steam in the recirculation circuit during process transients is therefore important. A simple detector using continuous ultrasonic waves and able to detect bubbles in water contained in steel tubes is described in this paper. The variation of the transmitted wave caused by the bubbles was determined by demodulation. The results have met the objectives set for cold water with air bubbles. A clear indication of the presence of steam bubbles was found in fast-flowing hot water in a steel tube with a diameter of 60 mm. A change in the low-frequency region of the modulation was the only indication of the presence of steam bubbles in the large-diameter downcomer of the water-separator drum of a boiler in an electrical power plant. Possible causes of the differences in the results obtained are discussed on the basis of differences in bubble sizes and in focusing and reflection of the ultrasonic waves. (orig.)

  12. Evaluation of maintenance strategies for steam generator tubes in pressurized waster reactors. 2. Cost and profitability analyses

    International Nuclear Information System (INIS)

    Isobe, Y.; Sagisaka, M.; Yoshimura, S.; Yagawa, G.

    2000-01-01

    As an application of probabilistic fracture mechanics (PFM), risk-benefit analysis was carried out to evaluate maintenance activities of steam generator (SG) tubes used in pressurized water reactors (PWRs). The analysis was conducted for SG tubes made of Inconel 600, and Inconel 690 as well assuming its crack initiation and crack propagation law based on Inconel 600 data. The following results were drawn from the analysis. Improvement of inspection accuracy reduces the maintenance costs significantly and is preferable from the viewpoint of profitability due to reduction of SG tube leakage and rupture. There is a certain region of SCC properties of SG tubes where sampling inspection is effective. (author)

  13. Modeling of electrochemistry and steam-methane reforming performance for simulating pressurized solid oxide fuel cell stacks

    Energy Technology Data Exchange (ETDEWEB)

    Recknagle, Kurtis P.; Ryan, Emily M.; Koeppel, Brian J.; Mahoney, Lenna A.; Khaleel, Moe A. [Pacific Northwest National Laboratory, Richland, WA 99352 (United States)

    2010-10-01

    This paper examines the electrochemical and direct internal steam-methane reforming performance of the solid oxide fuel cell when subjected to pressurization. Pressurized operation boosts the Nernst potential and decreases the activation polarization, both of which serve to increase cell voltage and power while lowering the heat load and operating temperature. A model considering the activation polarization in both the fuel and the air electrodes was adopted to address this effect on the electrochemical performance. The pressurized methane conversion kinetics and the increase in equilibrium methane concentration are considered in a new rate expression. The models were then applied in simulations to predict how the distributions of direct internal reforming rate, temperature, and current density are effected within stacks operating at elevated pressure. A generic 10 cm counter-flow stack model was created and used for the simulations of pressurized operation. The predictions showed improved thermal and electrical performance with increased operating pressure. The average and maximum cell temperatures decreased by 3% (20 C) while the cell voltage increased by 9% as the operating pressure was increased from 1 to 10 atm. (author)

  14. Experimental study of the zirconium alloy oxidation under high pressure of steam and modelling of the mechanisms

    International Nuclear Information System (INIS)

    Dali, Yacoub

    2007-01-01

    The corrosion of the cladding materials used for the fuel rods is one of the limiting factor of their lifetime in light water reactors. In this field, the aim of the nuclear industry is today to increase the time and the number of cycles and to submit the claddings in zirconium alloys to higher corrosive conditions. In this way, new alloys devoted to replace the standard Zircaloy-4, for instance Nb containing alloys, have been recently developed and licensed and show better corrosion resistance. A better understanding of the corrosion mechanisms of the zirconium alloys is necessary to predict the corrosion behaviour of these materials. In this work, the oxidation rate of model alloys of two metallurgic families has been studied in steam in a pressure range between 100 milli-bars and 100 bars. The Zircaloy type alloys contain as alloying elements oxygen and/or tin and/or iron and chromium. For the Zr-Nb family, three niobium contents have been studied, respectively 0.2, 0.4 and 1 weight percent of niobium. Our objectives were to understand the variations of the reactivity between the low pressure and the high pressure range, in quantifying the dependency of the corrosion rate with the steam pressure and the alloying element concentrations. The segregation process of the niobium at the surface has also been studied on the Zr-Nb alloys. During this work, a magnetic suspension thermo-balance has been developed and used to follow in-situ the corrosion rate at high pressure of water vapour. The oxide layers have been characterized by many techniques, macro and micro-photo-electrochemistry, XRD, FEG-SEM, XPS, HR-TEM and SIMS. For the Zircaloy type alloys, we have confirmed the major role of the intermetallic precipitates Zr(Fe,Cr) 2 on the corrosion resistance. Unlike the standard Zircaloy-4, for which the oxidation rate does not depend on the pressure of the water vapour and is thus limited by the vacancy diffusion in the oxide layer, we have shown that the rate of the

  15. Effect of aviation fuel type and fuel injection conditions on the spray characteristics of pressure swirl and hybrid air blast fuel injectors

    Science.gov (United States)

    Feddema, Rick

    Feddema, Rick T. M.S.M.E., Purdue University, December 2013. Effect of Aviation Fuel Type and Fuel Injection Conditions on the Spray Characteristics of Pressure Swirl and Hybrid Air Blast Fuel Injectors. Major Professor: Dr. Paul E. Sojka, School of Mechanical Engineering Spray performance of pressure swirl and hybrid air blast fuel injectors are central to combustion stability, combustor heat management, and pollutant formation in aviation gas turbine engines. Next generation aviation gas turbine engines will optimize spray atomization characteristics of the fuel injector in order to achieve engine efficiency and emissions requirements. Fuel injector spray atomization performance is affected by the type of fuel injector, fuel liquid properties, fuel injection pressure, fuel injection temperature, and ambient pressure. Performance of pressure swirl atomizer and hybrid air blast nozzle type fuel injectors are compared in this study. Aviation jet fuels, JP-8, Jet A, JP-5, and JP-10 and their effect on fuel injector performance is investigated. Fuel injector set conditions involving fuel injector pressure, fuel temperature and ambient pressure are varied in order to compare each fuel type. One objective of this thesis is to contribute spray patternation measurements to the body of existing drop size data in the literature. Fuel droplet size tends to increase with decreasing fuel injection pressure, decreasing fuel injection temperature and increasing ambient injection pressure. The differences between fuel types at particular set conditions occur due to differences in liquid properties between fuels. Liquid viscosity and surface tension are identified to be fuel-specific properties that affect the drop size of the fuel. An open aspect of current research that this paper addresses is how much the type of aviation jet fuel affects spray atomization characteristics. Conventional aviation fuel specifications are becoming more important with new interest in alternative

  16. On the Accurate Determination of Shock Wave Time-Pressure Profile in the Experimental Models of Blast-Induced Neurotrauma

    Directory of Open Access Journals (Sweden)

    Maciej Skotak

    2018-02-01

    Full Text Available Measurement issues leading to the acquisition of artifact-free shock wave pressure-time profiles are discussed. We address the importance of in-house sensor calibration and data acquisition sampling rate. Sensor calibration takes into account possible differences between calibration methodology in a manufacturing facility, and those used in the specific laboratory. We found in-house calibration factors of brand new sensors differ by less than 10% from their manufacturer supplied data. Larger differences were noticeable for sensors that have been used for hundreds of experiments and were as high as 30% for sensors close to the end of their useful lifetime. These observations were despite the fact that typical overpressures in our experiments do not exceed 50 psi for sensors that are rated at 1,000 psi maximum pressure. We demonstrate that sampling rate of 1,000 kHz is necessary to capture the correct rise time values, but there were no statistically significant differences between peak overpressure and impulse values for low-intensity shock waves (Mach number <2 at lower rates. We discuss two sources of experimental errors originating from mechanical vibration and electromagnetic interference on the quality of a waveform recorded using state-of-the-art high-frequency pressure sensors. The implementation of preventive measures, pressure acquisition artifacts, and data interpretation with examples, are provided in this paper that will help the community at large to avoid these mistakes. In order to facilitate inter-laboratory data comparison, common reporting standards should be developed by the blast TBI research community. We noticed the majority of published literature on the subject limits reporting to peak overpressure; with much less attention directed toward other important parameters, i.e., duration, impulse, and dynamic pressure. These parameters should be included as a mandatory requirement in publications so the results can be properly

  17. Optimum design of dual pressure heat recovery steam generator using non-dimensional parameters based on thermodynamic and thermoeconomic approaches

    International Nuclear Information System (INIS)

    Naemi, Sanaz; Saffar-Avval, Majid; Behboodi Kalhori, Sahand; Mansoori, Zohreh

    2013-01-01

    The thermodynamic and thermoeconomic analyses are investigated to achieve the optimum operating parameters of a dual pressure heat recovery steam generator (HRSG), coupled with a heavy duty gas turbine. In this regard, the thermodynamic objective function including the exergy waste and the exergy destruction, is defined in such a way to find the optimum pinch point, and consequently to minimize the objective function by using non-dimensional operating parameters. The results indicated that, the optimum pinch point from thermodynamic viewpoint is 2.5 °C and 2.1 °C for HRSGs with live steam at 75 bar and 90 bar respectively. Since thermodynamic analysis is not able to consider economic factors, another objective function including annualized installation cost and annual cost of irreversibilities is proposed. To find the irreversibility cost, electricity price and also fuel price are considered independently. The optimum pinch point from thermoeconomic viewpoint on basis of electricity price is 20.6 °C (75 bar) and 19.2 °C (90 bar), whereas according to the fuel price it is 25.4 °C and 23.7 °C. Finally, an extensive sensitivity analysis is performed to compare optimum pinch point for different electricity and fuel prices. -- Highlights: ► Presenting thermodynamic and thermoeconomic optimization of a heat recovery steam generator. ► Defining an objective function consists of exergy waste and exergy destruction. ► Defining an objective function including capital cost and cost of irreversibilities. ► Obtaining the optimized operating parameters of a dual pressure heat recovery boiler. ► Computing the optimum pinch point using non-dimensional operating parameters

  18. Bursting-protection configuration for cylindrical steam generators of pressurized water reactors

    International Nuclear Information System (INIS)

    Mutzl, J.

    1979-01-01

    The bursting-protection jacket consists of cylinder courses, being joined together in axial direction, and of a bottom and a cover, being connected by means of axial prestressing tendons. For absorption and transmission of the steam generator weight and the bursting forces the bottom consists of a conical shell, tapered towards the side of the steam generator, and a support ring supporting the bottom circle of the cone. This support ring is built in sandwich construction and is connected with the axial tendons. The conical shell may be reinforced by radial ribs. If a primary coolant pump is built in there is provided for a rocking bearing between its pump casing flange and the bottom. (DG) [de

  19. Investigation of temperature fluctuation phenomena in a stratified steam-water two-phase flow in a simulating pressurizer spray pipe of a pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Miyoshi, Koji, E-mail: miyoshi.koj@inss.co.jp; Takenaka, Nobuyuki; Ishida, Taisuke; Sugimoto, Katsumi

    2017-05-15

    Highlights: • Thermal hydraulics phenomena were discussed in a spray pipe of pressurizer. • Temperature fluctuation was investigated in a stratified steam-water two-phase. • Remarkable liquid temperature fluctuations were observed in the liquid layer. • The observed temperature fluctuations were caused by the internal gravity wave. • The temperature fluctuations decreased with increasing dissolved oxygen. - Abstract: Temperature fluctuation phenomena in a stratified steam-water two-phase flow in a horizontal rectangular duct, which simulate a pressurizer spray pipe of a pressurized water reactor, were studied experimentally. Vertical distributions of the temperature and the liquid velocity were measured with water of various dissolved oxygen concentrations. Large liquid temperature fluctuations were observed when the water was deaerated well and dissolved oxygen concentration was around 10 ppb. The large temperature fluctuations were not observed when the oxygen concentration was higher. It was shown that the observed temperature fluctuations were caused by the internal gravity wave since the Richardson numbers were larger than 0.25 and the temperature fluctuation frequencies were around the Brunt-Väisälä frequencies in the present experimental conditions. The temperature fluctuations decreased by the non-condensable gas since the non-condensable gas suppressed the condensation and the temperature difference in the liquid layer was small.

  20. Steam Generator Group Project. Task 6. Channel head decontamination

    International Nuclear Information System (INIS)

    Allen, R.P.; Clark, R.L.; Reece, W.D.

    1984-08-01

    The Steam Generator Group Project utilizes a retired-from-service pressurized-water-reactor steam generator as a test bed and source of specimens for research. An important preparatory step to primary side research activities was reduction of the radiation field in the steam generator channel head. This task report describes the channel head decontamination activities. Though not a programmatic research objective it was judged beneficial to explore the use of dilute reagent chemical decontamination techniques. These techniques presented potential for reduced personnel exposure and reduced secondary radwaste generation, over currently used abrasive blasting techniques. Two techniques with extensive laboratory research and vendors prepared to offer commercial application were tested, one on either side of the channel head. As indicated in the report, both techniques accomplished similar decontamination objectives. Neither technique damaged the generator channel head or tubing materials, as applied. This report provides details of the decontamination operations. Application system and operating conditions are described

  1. Study on frictional pressure drop of steam-water two phase flow in optimized four-head internal-ribbed tube

    International Nuclear Information System (INIS)

    Wang Weishu; Zhu Xiaojing; Bi Qincheng; Wu Gang; Yu Shuiqing

    2012-01-01

    The optimized internal-ribbed tube is different from the normal internal-ribbed tube on the frictional pressure drop characteristics. The frictional pressure drop characteristics of steam-water two phase flow in horizontal four-head optimized internal-ribbed were studied under adiabatic condition. According to the experimental and calculation results, the two-phase multiplier is greatly affected by the steam quality and pressure. The two-phase multiplier increases with increasing quality, and decreases with increasing pressure. In the near-critical pressure region, the two-phase multiplier is close to 1. The frictional pressure drop of two phase flow in optimized tube is less than that in the normal tube under the same work condition. The good hydrodynamic condition could be achieved when the optimized internal-ribbed tube is used in the heat transfer equipment because the self-compensating characteristics exist due to the reduction of frictional pressure drop. (authors)

  2. Steam chugging in a boiling water reactor pressure-suppression system

    International Nuclear Information System (INIS)

    Pitts, J.H.

    1980-01-01

    Results of a transient analysis predicting the general characteristics of steam chugging compare well with the results of two large scale experiments: GKM II, test 21 and GKSS, test 16. Predicted fundamental periods of chugging are within 5 and 16 per cent of the respective experimental values. The results of the analysis include effects of air in the drywell, momentum loss and heat transfer in the condensation pipe, direct contact condensation heat transfer at the gas-water interface and momentum and heat transfer in the wetwell water pool. Bubble shape is calculated in two-dimensional cylindrical coordinates. Required inputs to the analysis include the geometry, initial conditions and constants to determine both the steam inlet mass flowrate to the drywell as a function of time and conduction heat transfer through the wall of the condensation pipe. There are no arbitrary free parameters which must be specified to predict specific experiments. Rather, the analysis is based on fundamental physical phenomena, experimental coefficients documented for general heat transfer and fluid mechanics characteristics and standard analytical techniques. The random nature of steam chugging observed in some experiments is partially explained by predicted regimes of chugging and changes in the maximum extent of a bubble below the condensation pipe exit during each regime. (orig.)

  3. Experimental investigation of pressure and blockage effects on combustion limits in H2-air-steam mixtures

    International Nuclear Information System (INIS)

    Sherman, M.P.; Berman, M.; Beyer, R.F.

    1993-06-01

    Experiments with hydrogen-air-steam mixtures, such as those found within a containment system following a reactor accident, were conducted in the Heated Detonation Tube (43 cm diameter and 12 m long) to determine the region of benign combustion; i.e., the region between the flammability limits and the deflagration-to-detonation transition limits. Obstacles were used to accelerate the flame; these include 30% blockage ratio annular rings, and alternate rings and disks of 60% blockage ratio. The initial conditions were 110 degree C and one or three atmospheres pressure. A benign burning region exists for rich mixtures, but is generally smaller than for lean mixtures. Effects of the different obstacles and of the different pressures are discussed

  4. Runout distance and dynamic pressure of pyroclastic density currents: Evidence from 18 May 1980 blast surge of Mount St. Helens

    Science.gov (United States)

    Gardner, J. E.; Andrews, B. J.

    2016-12-01

    Pyroclastic density currents (flows and surges) are one of the most deadly hazards associated with volcanic eruptions. Understanding what controls how far such currents will travel, and how their dynamic pressure evolves, could help mitigate their hazards. The distance a ground hugging, pyroclastic density current travels is partly limited by when it reverses buoyancy and lifts off into the atmosphere. The 1980 blast surge of Mount St. Helens offers an example of a current seen to lift off. Before lofting, it had traveled up to 20 km and leveled more than 600 km3 of thick forest (the blowdown zone). The outer edge of the devastated area - where burned trees that were left standing (the singe zone) - is where the surge is thought to have lifted off. We recently examined deposits in the outer parts of the blowdown and in the singe zone at 32 sites. The important finding is that the laterally moving surge travelled into the singe zone, and hence the change in tree damage does not mark the run out distance of the ground hugging surge. Eyewitness accounts and impacts on trees and vehicles reveal that the surge consisted of a fast, dilute "overcurrent" and a slower "undercurrent", where most of the mass (and heat) was retained. Reasonable estimates for flow density and velocity show that dynamic pressure of the surge (i.e., its ability to topple trees) peaked near the base of the overcurrent. We propose that when the overcurrent began to lift off, the height of peak dynamic pressure rose above the trees and stopped toppling them. The slower undercurrent continued forward, burning trees but it lacked the dynamic pressure needed to topple them. Grain-size variations argue that it slowed from 30 m/s when it entered the singe zone to 3 m/s at the far end. Buoyancy reversal and liftoff are thus not preserved in the deposits where the surge lofted upwards.

  5. Dynamic loads caused by pressure blasts, steam explosions, and earth quakes; Dynamische Belastungen durch Druckstoesse, Dampfexplosionen und Erdbeben

    Energy Technology Data Exchange (ETDEWEB)

    Hofmann, H H [SDK Ingenieurunternehmen GmbH, Basel (Switzerland)

    1998-11-01

    The paper deals with description of structures and the relevant dynamic loads. As to the structures, gas, fluid, or solid structures are to be considered. They determine the characteristic vibrational behaviour of the structures in the interconnected system. The excitation type determines the component that will be induced to change characteristic vibrational behaviour of the structure, depending on the load increasing time and the period of excitation. Three examples are given to illustrate the processes. (Water tank subject to quasi-seismic conditions; pipeline affected by blow-down; shut-off valve for a pipe). (orig./CB) [Deutsch] In diesem Beitrag soll auf die Erfassung der Strukturen und die Erfassung der dynamischen Belastungen eingegangen werden. Zur Erfassung der Strukturen sind `Gas-, Fluid- und Festkoerper-Strukturen` zu beachten. Sie bestimmen das Eigenschwingverhalten im Verbund. Die Erregung bestimmt nun, welcher Bereich aus dem Eigen-Schwingverhalten der Struktur ueber die Lastanstiegs-Zeit und die Zeitdauer der Erregung anregbar ist. Drei Beispiele sollen die Aufgabenstellung erlaeutern (Wasserbehaelter unter erdbebenaehnlichen Bedingungen; Rohrleitung unter `Blow-down-Belastung`; Absperrklappe fuer eine Rohrleitung). (orig./MM)

  6. Energy efficiency of a direct-injection internal combustion engine with high-pressure methanol steam reforming

    International Nuclear Information System (INIS)

    Poran, Arnon; Tartakovsky, Leonid

    2015-01-01

    This article discusses the concept of a direct-injection ICE (internal combustion engine) with thermo-chemical recuperation realized through SRM (steam reforming of methanol). It is shown that the energy required to compress the reformate gas prior to its injection into the cylinder is substantial and has to be accounted for. Results of the analysis prove that the method of reformate direct-injection is unviable when the reforming is carried-out under atmospheric pressure. To reduce the energy penalty resulted from the gas compression, it is suggested to implement a high-pressure reforming process. Effects of the injection timing and the injector's flow area on the ICE-SRM system's fuel conversion efficiency are studied. The significance of cooling the reforming products prior to their injection into the engine-cylinder is demonstrated. We show that a direct-injection ICE with high-pressure SRM is feasible and provides a potential for significant efficiency improvement. Development of injectors with greater flow area shall contribute to further efficiency improvements. - Highlights: • Energy needed to compress the reformate is substantial and has to be accounted for. • Reformate direct-injection is unviable if reforming is done at atmospheric pressure. • Direct-injection engine with high-pressure methanol reforming is feasible. • Efficiency improvement by 12–14% compared with a gasoline-fed engine was shown

  7. A theoretical analysis of flow through the nucleating stage in a low pressure steam turbine

    International Nuclear Information System (INIS)

    Skillings, S.A.; Walters, P.T.; Jackson, R.

    1989-01-01

    In order to improve steam turbine efficiency and reliability, the phenomena associated with the formation and growth of water droplets must be understood. This report describes a theoretical investigation into flow behaviour in the nucleating stage, where the predictions of a one-dimensional theory are compared with measured turbine data. Results indicate that droplet sizes predicted by homogeneous condensation theory cannot be reconciled with measurements unless fluctuating shock waves arise. Heterogeneous effects and flow turbulence are also discussed along with their implications for the condensation process. (author)

  8. Denting of Inconel steam generator tubes in pressurized water reactors. Third informal report

    International Nuclear Information System (INIS)

    van Rooyen, D.; Weeks, J.R.

    1977-08-01

    The recent plant operating experience and laboratory test results on the phenomenon of denting in recirculating PWR steam generators is reviewed. Although denting was first reported only in plants that were converted from phosphate to AVT, it has now also been observed in plants still on phosphate, as well as in some that started on AVT. In some units, slightly abnormal eddy current signals have been observed at the top of the tube sheets. The degree of denting in operating steam generators may be related to the levels and duration of chloride inleakage. Chloride, however, is not the only active ingredient, and does not seem to give denting until local acid conditions arise; consequently, it may be necessary for soluble copper and/or nickel ions to be present to promote the denting reaction. Chloride concentrations in actively corroding crevices can increase by several orders of magnitude over the bulk coolant. It is thus difficult to develop a basis for Cl - specifications for secondary water. Maintaining Cl - low enough to prevent denting may be unmanageable without full flow condensate demineralization in coastal plants with copper alloy condensors and feedwater lines. Cathodic depolarization by oxidizing species are thought to promote the formation of acid chlorides in crevices and trigger the denting reactions; some ions may also catalyze the rapid formation of magnetite. These, and other mechanistic aspects of denting are discussed. The implications of the Inconel 600 tube defects at Ginna in non-dented areas, originating from the primary side, are also discussed

  9. Imaging optical probe for pressurized 6200K steam-water environment

    International Nuclear Information System (INIS)

    Donaldson, M.R.; Pulfrey, R.E.; Merrill, S.K.

    1979-01-01

    An air-cooled imaging optical probe, 0.3 m long with a 25.4-mm outside diameter, has been built to provide high resolution viewing of flow regimes in a steam-water environment at 620 0 K and 15.5 MPa. The probe consists of a 3.5-mm-diameter rod lens borescope, surrounded by two coaxial coolant flow channels and two coaxial insulating dead air spaces. With air flowing through the probe at 5.7 g/s, thermal analysis shows that no part of the optical borescope will exceed 366 0 K when the probe is immersed in a 620 0 K environment. The objective lens is protected by a sapphire window which tests have shown can survive over 200 hours in 620 0 K water or steam with negligible loss of resolution and contrast. Condensation on the protective window is boiled off by electrically heating the window. Computer stress analysis, plus actual tests, shows that the probe can operate successfully with conservative safety factors

  10. Applicability of Alignment and Combination Rules to Burst Pressure Prediction of Multiple-flawed Steam Generator Tube

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Myeong Woo; Kim, Ji Seok; Kim, Yun Jae [Korea University, Seoul (Korea, Republic of); Jeon, Jun Young [Doosan Heavy Industries and Consruction, Seoul (Korea, Republic of); Lee, Dong Min [Korea Plant Service and Engineering, Technical Research and Development Institute, Naju (Korea, Republic of)

    2016-05-15

    Alignment and combination rules are provided by various codes and standards. These rules are used to determine whether multiple flaws should be treated as non-aligned or as coplanar, and independent or combined flaws. Experimental results on steam generator (SG) tube specimens containing multiple axial part-through-wall (PTW) flaws at room temperature (RT) are compared with assessment results based on the alignment and combination rules of the codes and standards. In case of axial collinear flaws, ASME, JSME, and BS7910 treated multiple flaws as independent flaws and API 579, A16, and FKM treated multiple flaws as combined single flaw. Assessment results of combined flaws were conservative. In case of axial non-aligned flaws, almost flaws were aligned and assessment results well correlate with experimental data. In case of axial parallel flaws, both effective flaw lengths of aligned flaws and separated flaws was are same because of each flaw length were same. This study investigates the applicability of alignment and combination rules for multiple flaws on the failure behavior of Alloy 690TT steam generator (SG) tubes that widely used in the nuclear power plan. Experimental data of burst tests on Alloy 690TT tubes with single and multiple flaws that conducted at room temperature (RT) by Kim el al. compared with the alignment rules of these codes and standards. Burst pressure of SG tubes with flaws are predicted using limit load solutions that provide by EPRI Handbook.

  11. Application of acoustic emission monitoring to pressure tests of a steam receiver vessel with flawed nozzle welds

    International Nuclear Information System (INIS)

    Woodward, B.; McDonald, N.R.; Hincksman, M.J.

    1976-01-01

    As part of the first stage of an Australian Welding Research Association co-operative research project, acoustic emission monitoring has been applied to a steam receiver vessel withdrawn from service owing to severe weld cracking. This technique is used to check acceptance standards for defects in nozzle welds and to apply modern methods of assessing the integrity of pressurised plant. Acoustic emission monitoring has been used, together with strain gauge measurements and ultrasonic scanning, to detect the occurrence of any significant defect growth during cyclic pressurisation of the vessel. During this first stage, no significant defect growth has been produced by 1000 cycles of pressure up to 24.1 MPa (3500 psi), subsequent pressurisation up to 35.8 MPa (5200 psi), or 97 per cent of the expected yield stress of the vessel shell. The small amount of acoustic emission detected was consistent with this result. (author)

  12. A fuzzy decision tree method for fault classification in the steam generator of a pressurized water reactor

    International Nuclear Information System (INIS)

    Zio, Enrico; Baraldi, Piero; Popescu, Irina Crenguta

    2009-01-01

    This paper extends a method previously introduced by the authors for building a transparent fault classification algorithm by combining the fuzzy clustering, fuzzy logic and decision trees techniques. The baseline method transforms an opaque, fuzzy clustering-based classification model into a fuzzy logic inference model based on linguistic rules which can be represented by a decision tree formalism. The classification model thereby obtained is transparent in that it allows direct interpretation and inspection of the model. An extension in the procedure for the development of the fuzzy logic inference model is introduced to allow the treatment of more complicated cases, e.g. splitted and overlapping clusters. The corresponding computational tool developed relies on a number of parameters which can be tuned by the user to optimally compromise the level of transparency of the classification process and its efficiency. A numerical application is presented with regards to the fault classification in the Steam Generator of a Pressurized Water Reactor.

  13. GERDA test facility for pressurized water reactors with straight tube steam generators

    International Nuclear Information System (INIS)

    Ahrens, G.; Haury, G.; Lahner, K.; Schatz, A.

    1983-01-01

    A number of large-scale experimental facilities have been constructed and operate in order to experiment on the thermodynamic and thermohydraulic behaviour of nuclear facilities in case of LOCA. Most of them were designed for ''large leak'' accidents, but as ''small leak'' accidents became the focus of interest, such experiments were also carried out. Experiments carried out with this arrangement for PWR-type reactors with straight-tube steam generators are only partially evaluable. BBR and B and W therefore cooperated in the construction of the test facility GERDA, designed for testing reactors of BBR design. It supplied relevant experimental results for the nuclear power plant at Muelheim-Kaerlich. (orig.) [de

  14. Saturated steams pressure of HfCl/sub 4/-KCl molten mixtures

    Energy Technology Data Exchange (ETDEWEB)

    Salyulev, A B; Smirnov, M V; Kudyakov, V Ya [AN SSSR, Sverdlovsk. Inst. Ehlektrokhimii

    1980-02-01

    A bellows null pressure gauge and the dynamic method were used to measure the total and partial pressures of saturated vapors of individual components of molten HfCl/sub 4/-KCl mixtures, as a function of temperature (260 to 1000 deg C) and composition (1.9 to 64.3 mol.% HfCl/sub 4/). Empirical equations expressing the relationship between pressure and temperature are presented. It is shown that in molten mixtures of hafnium tetrachloride with chlorides of alkaline metals its partial pressure dramatically increases when potassium chloride substitutes for cesium chloride.

  15. Saturated steams pressure of HfCl4-KCl molten mixtures

    International Nuclear Information System (INIS)

    Salyulev, A.B.; Smirnov, M.V.; Kudyakov, V.Ya.

    1980-01-01

    A bellows null pressure gauge and the dynamic method were used to measure the total and partial pressures of saturated vapors of individual components of molten HfCl 4 -KCl mixtures, as a function of temperature (260 to 1000 deg C) and composition (1.9 to 64.3 mol.% HfCl 4 ). Empirical equations expressing the relationship between pressure and temperature are presented. It is shown that in molten mixtures of hafnium tetrachloride with chlorides of alkaline metals its partial pressure dramatically increases when potassium chloride substitutes for cesium chloride

  16. Steam generator sludge removal apparatus

    International Nuclear Information System (INIS)

    Schafer, B.W.; Werner, C.E.; Klahn, F.C.

    1992-01-01

    The present invention relates to equipment for cleaning steam generators and in particular to a high pressure fluid lance for cleaning sludge off the steam generator tubes away from an open tube lane. 6 figs

  17. Reduction of Erosion Wear of Mean Pressure Cylinder of Steam Turbines Operating Beyond Critical Parameters

    Directory of Open Access Journals (Sweden)

    V. P. Kascheev

    2009-01-01

    Full Text Available The paper considers problems leading to erosion wear of flowing part of a mean pressure turbine cylinder operating beyond critical parameters. Explanation of erosion wear of flowing part of a mean pressure turbine cylinder which is proved in practice and recommendations for wear reduction are given in the paper

  18. Optimization of High Temperature and Pressurized Steam Modified Wood Fibers for High-Density Polyethylene Matrix Composites Using the Orthogonal Design Method

    Directory of Open Access Journals (Sweden)

    Xun Gao

    2016-10-01

    Full Text Available The orthogonal design method was used to determine the optimum conditions for modifying poplar fibers through a high temperature and pressurized steam treatment for the subsequent preparation of wood fiber/high-density polyethylene (HDPE composites. The extreme difference, variance, and significance analyses were performed to reveal the effect of the modification parameters on the mechanical properties of the prepared composites, and they yielded consistent results. The main findings indicated that the modification temperature most strongly affected the mechanical properties of the prepared composites, followed by the steam pressure. A temperature of 170 °C, a steam pressure of 0.8 MPa, and a processing time of 20 min were determined as the optimum parameters for fiber modification. Compared to the composites prepared from untreated fibers, the tensile, flexural, and impact strength of the composites prepared from modified fibers increased by 20.17%, 18.5%, and 19.3%, respectively. The effect on the properties of the composites was also investigated by scanning electron microscopy and dynamic mechanical analysis. When the temperature, steam pressure, and processing time reached the highest values, the composites exhibited the best mechanical properties, which were also well in agreement with the results of the extreme difference, variance, and significance analyses. Moreover, the crystallinity and thermal stability of the fibers and the storage modulus of the prepared composites improved; however, the hollocellulose content and the pH of the wood fibers decreased.

  19. Niobium alloys production with elements of high steam pressure and high ductilidate Nb46,5%Ti, Nb 1%Zr, Nb 1%Ti and Nb20% Ta

    International Nuclear Information System (INIS)

    Pinatti, D.G.; Baldan, C.A.; Dainesi, C.R.; Sandim, H.R.Z.

    1988-01-01

    The melting technology of niobium alloys with high ductilidade and high steam pressure, having the Ti, Zr and Ta as alloying elements is described. The electron beam technique for production of Nb 46,5%Ti, Nb 1%Zr and Nb 20%Ta alloys is analysed, aiming a product with high grade and low cost. (C.G.C.) [pt

  20. Frictional pressure drop of steam-water two-phase flow in helical coils with small helix diameter of HTR-10

    International Nuclear Information System (INIS)

    Bi Qincheng; Chen Tingkuan; Luo Yushan; Zheng Jianxue

    1996-01-01

    Experiments of steam-water two-phase flow frictional pressure drop through five vertically and horizontally positioned helical coils were carried out in the high pressure steam water test loop of Xi'an Jiaotong University. Two kinds of tube with inner diameters of 10 mm and 12 mm were used to form the coils. The helix diameter was 115 mm with coil pitch 22.5 mm. The experimental conditions were: pressure p = 4-14 MPa, mass velocity G = 400-2000 kg/(m 2 ·s), and inner wall heat flux q = 0-750 kW/m 2 . Theoretical analysis with a semi-empirical correlation was made to predict the two-phase flow fictional pressure drop through these kinds of helical coils

  1. Mathematical modelling of stress-deformation state of the steam generator collector (WWER-type) under pressure loading during fracture mechanics calculations

    International Nuclear Information System (INIS)

    Zaitsev, M.; Lyssakov, V.

    1993-01-01

    This paper describes a steam generator collector (WWER-type) designed as part of a Russian reactor power station. The collector is a thick cylindrical shell with a constant inner diameter of 850 mm and a height of 4,970 mm. The wall thickness varies from 78 to 163 mm. In the thicker section, a series of holes allows connection of steam generator heat exchanging tubes. Because of design considerations, the tubes are not symmetrically located about the circumference of the collector. This paper presents a model of the stress concentrations resulting from this design feature for a device operating at a nominal pressure of 16 MPa. 4 refs., 8 figs

  2. Transfer laws between water and freon 113 for average volumetric steam quality, pressure drop, and critical heat flux

    International Nuclear Information System (INIS)

    Nabizadeh, H.

    1977-01-01

    Simulation of the thermohydraulic processes of the steady-state reactor operation with boiling water and typical fuel element geometries leads to considerable increase of the heat rates to be tranferred and thus to an increase of the experimental cost which can hardly be justified. By proper choice of a model fluid with low heat of evaporation the system parameters like pressure, temperature, and heat rate, while retaining the original geometry, may be reduced to a fraction of those of the original fluid water. This permits not only a decrease in experimental cost but also a modification of the existing calculation data under more favorable experimental conditions. Starting from these considerations the cooling medium R113 was used as model fluid in carrying out the experiments. The necessary knowledge of the thermodynamical laws of simularity, however, have to be determined first of all in simple geometries and the scaling factors are then derived from them. In this connection the following experimental studies have been carried out with R113: a) average volumetric steam quality; b) two-phase pressure drop; c) critical heat flux. (orig.) [de

  3. Regulations for pressurized equipment in the European Single Market - construction of steam boilers, containers and pipelines

    International Nuclear Information System (INIS)

    Grassmuck, J.

    1992-01-01

    The impulses produced by the data of the standardized EC Single Market have now reached pressurized equipment in the field of EC Guidelines and European standardisation. This must be regarded as a great challenge to the interested and concerned parties. All efforts to represent the interested parties in European Committees must be made. In order to reach the goal quickly and successfully, a considerable readiness to compromise is, however, necessary. At the end of the development process, a comprehensible, standardized set of regulations will be available for pressurized equipment throughout Europe. The regulations will consist of national ones converted into European Guidelines and Standards. (orig.) [de

  4. Physical model of lean suppression pressure oscillation phenomena: steam condensation in the light water reactor pressure suppression system (PSS)

    International Nuclear Information System (INIS)

    McCauley, E.W.; Holman, G.S.; Aust, E.; Schwan, H.; Vollbrandt, J.

    1980-01-01

    Using the results of large scale multivent tests conducted by GKSS, a physical model of chugging is developed. The unique combination of accurate digital data and cinematic data has provided the derivation of a detailed, quantified correlation between the dynamic physical variables and the associated two-phase thermo-hydraulic phenomena occurring during lean suppression (chugging) phases of the loss-of-coolant accident in a boiling water reactor pressure suppression system

  5. Operational reliability of high pressure steam lines of pearlitic steels after 150-200 thousand h service

    International Nuclear Information System (INIS)

    Veksler, E.Ya.; Chajkovskij, V.M.; Osasyuk, V.V.

    1980-01-01

    Usage of both calculational and physical methods is recommended to estimate a service operating life of long-term working steam line materials. Application of these methods is demonstrated when studying steam line bends made of 12MKh and 12Kh1MF pearlitic steels. Good coincidence of results for the determination of residual durability of steam lines is obtained using these two methods [ru

  6. Pressure drop in the flow of gas/steam liquid mixtures in pipes

    International Nuclear Information System (INIS)

    Friedel, L.

    1978-01-01

    Pressure drop in two phase flow is considered to be made up of terms for geodetical elevation or depression, acceleration, and friction. The geodetical and momentum pressure drop are discussed and reasonable correlations are presented, along with their limitations and range of application. Various relationships are available for calculating the technically important friction component. If purely empirical correlations are neclected, all the remaining predictive schemes can be related to three basic physical models. These models as well as the commonly accepted and most reliable relationships are discussed and classified according to type. Furthermore, their scope is defined and the accuracy of prediction systematicaπly compared with the aid of a newly set-up data bank. The extensive literature data consulted refer to single component two phase flow of water and various refrigerants and several two-component systems in horizontal and vertical unheated straight pipes under industrially relevant flow conditions. Finally, the accuracy of the prediction of some generally accepted void correlations is dealt with: here too, numerous published void fraction data have been gathered and checked. (orig./HP) [de

  7. The Schoonebeek Oilfield: the Rw-2e High Pressure Steam Injection Project Gisement de Schoonebeek : le projet RW-2E d'injection de vapeur à haute pression

    Directory of Open Access Journals (Sweden)

    Holtam V. R.

    2006-11-01

    Full Text Available The daily oil production from the Schoonebeek Oilfield amounts to some 1400 m3 /d, of which ca. 65% is produced from a high pressure (85 bar steam injection project. This project was started in 1981 and originally consisted of 7 structurally downdip/middip steam injectors. However, following the initially somewhat disappointing project performance, steam injection was moved to 4 middip/ updip injectors in 1984. This change in the location of the steam injectors, together with an increase in the level of surveillance and a more pragmatic reservoir management policy, has resulted in improved project performance. The ultimate extra oil/steam ratio for the total project is now expected to be 0. 7 m3 oil/ton of steam injected. La production de pétrole du gisement de Schoonebeek est d'environ 1400 m3/jour, dont près de 65% sont obtenus par injection de vapeur à haute pression (85 bar. Ce projet lancé en 1981 comportait initialement 7 injecteurs de vapeur orientés vers l'aval-pendage. En raison de performances décevantes, l'injection de vapeur a été transférée en 1984 sur 4 injecteurs travaillant vers l'amont-pendage. Ce changement de position des injecteurs, accompagné d'une surveillance renforcée et d'une politique de gestion du gisement plus pragmatique, a donné des résultats favorables. On pense que le rapport pétrole/vapeur pour l'ensemble du projet devrait être en dernière analyse de 0,7 m3 de pétrole par tonne de vapeur injectée.

  8. Active latent heat storage with a screw heat exchanger - experimental results for heat transfer and concept for high pressure steam

    Science.gov (United States)

    Zipf, Verena; Willert, Daniel; Neuhäuser, Anton

    2016-05-01

    An innovative active latent heat storage concept was invented and developed at Fraunhofer ISE. It uses a screw heat exchanger (SHE) for the phase change during the transport of a phase change material (PCM) from a cold to a hot tank or vice versa. This separates heat transfer and storage tank in comparison to existing concepts. A test rig has been built in order to investigate the heat transfer coefficients of the SHE during melting and crystallization of the PCM. The knowledge of these characteristics is crucial in order to assess the performance of the latent heat storage in a thermal system. The test rig contains a double shafted SHE, which is heated or cooled with thermal oil. The overall heat transfer coefficient U and the convective heat transfer coefficient on the PCM side hPCM both for charging and discharging have been calculated based on the measured data. For charging, the overall heat transfer coefficient in the tested SHE was Uch = 308 W/m2K and for discharging Udis = 210 W/m2K. Based on the values for hPCM the overall heat transfer coefficients for a larger SHE with steam as heat transfer fluid and an optimized geometry were calculated with Uch = 320 W/m2K for charging and Udis = 243 W/m2K for discharging. For pressures as high as p = 100 bar, an SHE concept has been developed, which uses an organic fluid inside the flight of the SHE as working media. With this concept, the SHE can also be deployed for very high pressure, e.g. as storage in solar thermal power plants.

  9. Cracking of low-pressure steam turbine rotor discs in nuclear power plants

    International Nuclear Information System (INIS)

    McMinn, A.; Burghard, H.C. Jr.; Lyle, F.F. Jr.; Leverant, G.R.

    1984-01-01

    This paper describes the results of several metallurgical analyses of retired low pressure (LP) turbine discs that had suffered in-service cracking. Cracks were found in four locations; keyways, bores, web faces and rim attachment areas. In every case, the metallurgical analyses identified intergranular stress corrosion cracking (IGSCC) as the operative mechanism. The cracks normally have been filled with iron oxides; but chlorides, sulphates, carbonates, copper and copper oxide have been found in, or near cracks. In some cases deposits have been strongly alkaline. However, no specific corrodent has been identified as being uniquely responsible for the cracking in any of the discs. In every case, the disc materials met all mechanical-properties and chemical-composition requirements, and had normal microstructures

  10. Creep behavior under internal pressure of zirconium alloy cladding oxidized in steam at high temperature

    International Nuclear Information System (INIS)

    Chosson, Raphael

    2014-01-01

    During hypothetical Loss-Of-Coolant-Accident (LOCA) scenarios, zirconium alloy fuel cladding tubes creep under internal pressure and are oxidized on their outer surface at high temperature (HT). Claddings become stratified materials: zirconia and oxygen-stabilized α phase, called α(O), are formed on the outer surface of the cladding whereas the inner part remains in the β domain. The strengthening effect of oxidation on the cladding creep behavior under internal pressure has been highlighted at HT. In order to model this effect, the creep behavior of each layer had to be determined. This study focused on the characterization of the creep behavior of the α(O) phase at HT, through axial creep tests performed under vacuum on model materials, containing from 2 to 7 wt.% of oxygen and representative of the α(O) phase. For the first time, two creep flow regimes have been observed in this phase. Underlying physical mechanisms and relevant microstructural parameters have been discussed for each regime. The strengthening effect due to oxygen on the α(O) phase creep behavior at HT has been quantified and creep flow equations have been identified. A ductile to brittle transition criterion has been also suggested as a function of temperature and oxygen content. Relevance of the creep flow equations for each layer, identified in this study or from the literature, has been discussed. Then, a finite element model, describing the oxidized cladding as a stratified material, has been built. Based on this model, a fraction of the experimental strengthening during creep is predicted. (author) [fr

  11. The effect of water chemistry on a change in the composition of gas phase in the steam-water path of a supercritical-pressure boiler

    Science.gov (United States)

    Belyakov, I. I.; Belokonova, A. F.

    2010-07-01

    We present the results from an experimental research work on studying the behavior of the gas phase in the path of a supercritical-pressure boiler during its operation with different water chemistries, including all-volatile (hydrazine-ammonia), complexone, neutral oxygenated, and combined oxygenated-ammonia chemistries. It is shown that the minimal content of hydrogen in steam is achieved if feedwater is treated with oxygen.

  12. Inverse method for stress monitoring in pressure components of steam generators

    International Nuclear Information System (INIS)

    Duda, P.

    2003-01-01

    The purpose of this work is to formulate a space marching method, which can be used to solve inverse multidimensional heat conduction problems. The method is designed to reconstruct the transient temperature distribution in a whole construction element based on measured temperatures taken at selected points inside or on the outer surface of the construction element. Next, the Finite Element Method is used to calculate thermal stresses and stresses caused by other loads such as, for instance, internal pressure. The developed method for solving temperature and total stress distribution will be tested using the measured temperatures generated from a direct solution. Transient temperature and total stress distribution obtained from method presented below will be compared with the values obtained from the direct solution. Finally, the presented method will be applied in order to monitor temperature and stress distribution in an outlet header using the real measured temperature values at seven points on the header's outer surface during the power boiler's shut down operation. The presented method allows to optimize the power block's start-up and shut-down operations, contributes to the reduction of heat loss during these operations and to the extension of power block's life. The fatigue and creep usage factor can be computed in an on-line mode. The presented method herein can be applied to monitoring systems that work in conventional as well as in nuclear power plants. (author)

  13. Biomass oxygen/steam gasification in a pressurized bubbling fluidized bed: Agglomeration behavior

    International Nuclear Information System (INIS)

    Zhou, Chunguang; Rosén, Christer; Engvall, Klas

    2016-01-01

    Highlights: • Dolomite is a superior material in preventing bed agglomeration. • Small molten ash particles deposited on magnesite at bed temperatures above 1000 °C. • The performance, when using magnesite, is sensitive to temperature disturbances. • The anti-agglomeration mechanisms of Ca- and Mg-bearing materials were discussed. - Abstract: In this study, the anti-agglomeration abilities of Ca- and Mg-containing bed materials, including dolomite and magnesite, in a pressurized bubbling fluidized bed gasifier using pine pellets and birch chips as feedstock, is investigated. The most typical bed material—silica sand—was also included as a reference for comparison. The sustainability of the operation was evaluated via analyzing the temperatures at different levels along the bed height. During the performances, the aim was to keep the temperature at the bottom zone of the reactor at around 870 °C. However, the success highly depends on the bed materials used in the bed and the temperature can vary significantly in case of agglomeration or bad mixing of bed materials and char particles. Both Glanshammar and Sala dolomites performed well with no observed agglomeration tendencies. In case of magnesite, the bed exhibited a high agglomeration tendency. Silica sand displayed the most severe agglomeration among all bed materials, even when birch chips with a low silica content was fed at a relatively low temperature. The solid samples of all the bed materials were inspected by light microscopy and Scanning Electron Microscopy (SEM). The Energy Dispersive Spectroscopy (EDS) detector was used to detect the elemental distribution in the surface. The crystal chemical structure was analyzed using X-ray Diffraction (XRD). Magnesite agglomerates glued together by big molten ash particles. There was no coating layer detected on magnesite particles at bed temperatures – below 870 °C. But when the temperature was above 1000 °C, a significant amount of small molten

  14. Shear Forces during Blast, Not Abrupt Changes in Pressure Alone, Generate Calcium Activity in Human Brain Cells

    Science.gov (United States)

    2012-06-29

    8217[-·:’ 50 100 150 200 250 300 Time (sec) Mean Peak Shear Stress (Pa) 0.6 ɘ.0001 200 250 300 350 400 Fluid Volume (I- LL ) lack of correlation to...Blood Flow Metab 30: 255–266. 5. Nakagawa A, Manley GT, Gean AD, Ohtani K, Armonda R, et al. (2011) Mechanisms of primary blast-induced traumatic brain

  15. Cycle improvement for nuclear steam power plant

    International Nuclear Information System (INIS)

    Silvestri, G.J. Jr.

    1976-01-01

    A pressure-increasig ejector element is disposed in an extraction line intermediate to a high pressure turbine element and a feedwater heater. The ejector utilizes high pressure fluid from a reheater drain as the motive fluid to increase the pressure at which the extraction steam is introduced into the feedwater heater. The increase in pressure of the extraction steam entering the feedwater heater due to the steam passage through the ejector increases the heat exchange capability of the extraction steam thus increasing the overall steam power plant efficiency

  16. Digital implementation, simulation and tests in MATLAB of the models of Steam line, the turbines, the pressure regulator of a BWR type nucleo electric power plant

    International Nuclear Information System (INIS)

    Lopez R, A.

    2004-01-01

    In this phase of the project they were carried out exhaustive tests to the models of the steam lines, turbines and pressure regulator of a BWR type nucleo electric central for to verify that their tendencies and behaviors are it more real possible. For it, it was necessary to also analyze the transfer functions of the different components along the steam line until the power generator. Such models define alone the dominant poles of the system, what is not limitation to reproduce a wide range of anticipated transitoriness of a power station operation. In the same manner, it was integrated and proved the integrated model form with the models of feeding water of the SUN-RAH, simulating the nuclear reactor starting from predetermined entrances of the prospective values of the vessel. Also it was coupled with the graphic interface developed with the libraries DirectX implementing a specific monitoring panel for this system. (Author)

  17. Experimental investigation of condensation and mixing during venting of a steam / non-condensable gas mixture into a pressure suppression pool

    Energy Technology Data Exchange (ETDEWEB)

    De Walsche, C.; Cachard, F. de

    2000-07-01

    Experiments have been performed in the LINX facility to investigate condensation and mixing phenomena in pressure Suppression Pools (SPs), in the context of the European Simplified Boiling Water Reactor (ESBWR) study. As a contribution to the TEPSS project of the 4th European Framework Programme, eight medium-scale, separate-effect tests were carried out in which constant steam/air flow rates were injected below the surface of a two-metre diameter water pool, maintained at constant pressure, through a large downward vent. The vessel pressure was regulated, the pool temperature rising until equilibrium conditions with the incoming gas were reached. The SP temperature distribution was measured, as well as the inlet and outlet gas flow rates, and the overall condensation rate was estimated using mass and heat balances. The test matrix was based on steam mass floret and air mass fraction of the injected gas, the vent immersion depth, and the vessel pressure. Overall, the condensation was shown to be efficient for all tests performed, even for high non-condensable gas concentrations of the injected gas. Thermal stratification above the vent outlet was shown to be moderate. The tests performed allowed a better understanding to be gained of the mechanisms of condensation and mixing in the SP and Wetwell, and results were incorporated into an ORACLE database, to be used for further model development. (authors)

  18. Blast management

    OpenAIRE

    Shouraki, Mohammad Kargar; Naserkheil, Ali Asghar

    2011-01-01

    Blast Management (BM) is composed of the combination of human resources management (HRM) principles and concepts and various methods of quality management (QM) with a financial approach. BM is made up of three aspects: hard, soft and concept and BLAST means an explosive shift in organization's mindset and thought and rapid action against it. The first aspect, hard, includes a set of managerial toots and philosophies to improve the quality and productivity with a financial approach.

  19. Steaming ahead

    International Nuclear Information System (INIS)

    Anon.

    1997-01-01

    An example of the development of geothermal power in Indonesia is described. Wells are being drilled into the Salak volcano on Java, about 60km south of Jakarta. These let out high pressure hot water trapped 1 to 3km below the surface which can be flashed into steam for driving turbines. The hot water field has already produced 110MW of power since 1994 and is currently being expanded to 330MW. Some details of the drilling and civil engineering are given. Since Indonesia sits on the edge of giant tectonic boundary known as the ''Pacific ring of fire'', the potential for further development is enormous. Ultimately volcanic activity could release an estimated 27,000MW capacity. More realistically, 2,000MW of crustal power by 2020 is spoken of. (UK)

  20. Steam generator life management

    International Nuclear Information System (INIS)

    Tapping, R.L.; Nickerson, J.; Spekkens, P.; Maruska, C.

    1998-01-01

    Steam generators are a critical component of a nuclear power reactor, and can contribute significantly to station unavailability, as has been amply demonstrated in Pressurized Water Reactors (PWRs). CANDU steam generators are not immune to steam generator degradation, and the variety of CANDU steam generator designs and tube materials has led to some unexpected challenges. However, aggressive remedial actions, and careful proactive maintenance activities, have led to a decrease in steam generator-related station unavailability of Canadian CANDUs. AECL and the CANDU utilities have defined programs that will enable existing or new steam generators to operate effectively for 40 years. Research and development work covers corrosion and mechanical degradation of tube bundles and internals, chemistry, thermal hydraulics, fouling, inspection and cleaning, as well as provision for specially tool development for specific problem solving. A major driving force is development of CANDU-specific fitness-for-service guidelines, including appropriate inspection and monitoring technology to measure steam generator condition. Longer-range work focuses on development of intelligent on-line monitoring for the feedwater system and steam generator. New designs have reduced risk of corrosion and fouling, are more easily inspected and cleaned, and are less susceptible to mechanical damage. The Canadian CANDU utilities have developed programs for remedial actions to combat degradation of performance (Gentilly-2, Point Lepreau, Bruce A/B, Pickering A/B), and have developed strategic plans to ensure that good future operation is ensured. This report shows how recent advances in cleaning technology are integrated into a life management strategy, discusses downcomer flow measurement as a means of monitoring steam generator condition, and describes recent advances in hideout return as a life management tool. The research and development program, as well as operating experience, has identified

  1. Tachometric flowmeters for measuring circulation water parameters in steam generators of the NPPs running on pressurized water reactors

    International Nuclear Information System (INIS)

    Ageev, A.G.; Korolkov, B.M.; Nigmatulin, B.I.; Belov, V.I.; Vasileva, R.V.; Trubkin, N.I.

    1997-01-01

    Tachometric flowmeters used in steam generators for determining the velocity and direction of the flow have a limited service life owing to the use of corundum for the bearing assembly components. Various materials were investigated for the feasibility of using them as alternatives for replacing the corundum bearing and guide bushing under conditions close to the conditions in steam generators: 7 MPa, 260 degC. Good results were obtained with bearing assemblies fabricated from corrosion-resistant steel. Testing of the transducer design and optimization of the technique was accomplished in the course of testing steam generators of the WWER-1000 reactor at the Balakovskaya nuclear power plant. The velocity and direction of flow in the steam generator were measured within a wide range of unit power ratings up to the values corresponding to full power output. The service life of the transducers proved to be not less than 720 hours. The transducer parameters remained unchanged over the entire operation period. (M.D.)

  2. Thermodynamic performance simulation and concise formulas for triple-pressure reheat HRSG of gas–steam combined cycle under off-design condition

    International Nuclear Information System (INIS)

    Zhang, Guoqiang; Zheng, Jiongzhi; Yang, Yongping; Liu, Wenyi

    2016-01-01

    Highlights: • An off-design performance simulation of triple-pressure reheat HRSG is executed. • The bottoming cycle characteristics of energy transfer/conversion are analyzed. • Concise formulas for the off-design performance of bottoming cycle are proposed. • The accuracy of the formulas is verified under different load control strategies. • The errors of the formulas are generally within 1% at a load of 100–50%. - Abstract: Concise semi-theoretical, semi-empirical formulas are developed in this study to predict the off-design performance of the bottoming cycle of the gas–steam turbine combined cycle. The formulas merely refer to the key thermodynamic design parameters (full load parameters) of the bottoming cycle and off-design gas turbine exhaust temperature and flow, which are convenient in determining the overall performance of the bottoming cycle. First, a triple-pressure reheat heat recovery steam generator (HRSG) is modeled, and thermodynamic analysis is performed. Second, concise semi-theoretical, semi-empirical performance prediction formulas for the bottoming cycle are proposed through a comprehensive analysis of the heat transfer characteristics of the HRSG and the energy conversion characteristics of the steam turbine under the off-design condition. The concise formulas are found to be effective, i.e., fast, simple, and precise in obtaining the thermodynamic parameters for bottoming cycle efficiency, HRSG heat transfer capacity, HRSG efficiency, steam turbine power output, and steam turbine efficiency under the off-design condition. Accuracy is verified by comparing the concise formulas’ calculation results with the simulation results and practical operation data under different load control strategies. The calculation errors are within 1.5% (mainly less than 1% for both simulation and actual operation data) under combined cycle load (gas turbine load) ranging from 50% to 100%. However, accuracy declines sharply when the turbine

  3. Effects of Secondary Circuit Modeling on Results of Pressurized Water Reactor Main Steam Line Break Benchmark Calculations with New Coupled Code TRAB-3D/SMABRE

    International Nuclear Information System (INIS)

    Daavittila, Antti; Haemaelaeinen, Anitta; Kyrki-Rajamaeki, Riitta

    2003-01-01

    All of the three exercises of the Organization for Economic Cooperation and Development/Nuclear Regulatory Commission pressurized water reactor main steam line break (PWR MSLB) benchmark were calculated at VTT, the Technical Research Centre of Finland. For the first exercise, the plant simulation with point-kinetic neutronics, the thermal-hydraulics code SMABRE was used. The second exercise was calculated with the three-dimensional reactor dynamics code TRAB-3D, and the third exercise with the combination TRAB-3D/SMABRE. VTT has over ten years' experience of coupling neutronic and thermal-hydraulic codes, but this benchmark was the first time these two codes, both developed at VTT, were coupled together. The coupled code system is fast and efficient; the total computation time of the 100-s transient in the third exercise was 16 min on a modern UNIX workstation. The results of all the exercises are similar to those of the other participants. In order to demonstrate the effect of secondary circuit modeling on the results, three different cases were calculated. In case 1 there is no phase separation in the steam lines and no flow reversal in the aspirator. In case 2 the flow reversal in the aspirator is allowed, but there is no phase separation in the steam lines. Finally, in case 3 the drift-flux model is used for the phase separation in the steam lines, but the aspirator flow reversal is not allowed. With these two modeling variations, it is possible to cover a remarkably broad range of results. The maximum power level reached after the reactor trip varies from 534 to 904 MW, the range of the time of the power maximum being close to 30 s. Compared to the total calculated transient time of 100 s, the effect of the secondary side modeling is extremely important

  4. Steam purity in PWRs

    International Nuclear Information System (INIS)

    Hopkinson, J.

    1982-01-01

    Impurities enter the secondary loop of the PWR through both makeup water from lake or well and cooling-water leaks in the condenser. These impurities can be carried to the steam generator, where they cause corrosion deposits to form. Corrosion products in steam are swept further through the system and become concentrated at the point in the low-pressure turbine where steam begins to condense. Several plants have effectively reduced impurities, and therefore corrosion, by installing a demineralizer for the makeup water, a resin-bed system to clean condensed steam from the condenser, and a deaerator to remove oxygen from the water and so lower the risk of system metal oxidation. 5 references, 1 figure

  5. Fiscal 1999 technical survey report. Model project implementation feasibility study in India on effective utilization of blast furnace gas pressure energy; 1999 nendo Indo ni okeru koro gas atsuryoku energy yuko riyo model jigyo jisshi kanosei chosa hokokusho

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-03-01

    From the viewpoint of energy utilization stated above, blast furnace top pressure recovery turbine (TRT) unit installability was studied at Bhilai Steel Works of Steel Authority of India, Ltd., Bokaro Steel Works of Steel Authority of India, Ltd., and Visakhapatnum Steel Works of Rashtriya Ispat Nigem. The energy consumption rate at an Indian steelmaking plant is 8Gcal/t-steel, which is larger than 5-6Gcal/t-steel of Japan and therefore needs improvement. Out of the blast furnaces in India, 26 are larger than 1,000m{sup 3}, and two of them are provided with a TRT device of now-defunct Soviet Union manufacture. The blast furnaces were examined for pressure at the top, amount of gas at the top, amount of dust, and safeness in operation. The No. 2 blast furnace of the Borkaro plant was selected for the project, and studies were made for a wet type TRT device. Improvements to be achieved by TRT device installation were calculated to be a TRT output of 5,900kW, power output of 49,100MWh/year, saved crude oil amount of 12,990toe/year, and CO2 reduction of 40,200 tons-CO2/year. (NEDO)

  6. Steam generator tube failures

    International Nuclear Information System (INIS)

    MacDonald, P.E.; Shah, V.N.; Ward, L.W.; Ellison, P.G.

    1996-04-01

    A review and summary of the available information on steam generator tubing failures and the impact of these failures on plant safety is presented. The following topics are covered: pressurized water reactor (PWR), Canadian deuterium uranium (CANDU) reactor, and Russian water moderated, water cooled energy reactor (VVER) steam generator degradation, PWR steam generator tube ruptures, the thermal-hydraulic response of a PWR plant with a faulted steam generator, the risk significance of steam generator tube rupture accidents, tubing inspection requirements and fitness-for-service criteria in various countries, and defect detection reliability and sizing accuracy. A significant number of steam generator tubes are defective and are removed from service or repaired each year. This wide spread damage has been caused by many diverse degradation mechanisms, some of which are difficult to detect and predict. In addition, spontaneous tube ruptures have occurred at the rate of about one every 2 years over the last 20 years, and incipient tube ruptures (tube failures usually identified with leak detection monitors just before rupture) have been occurring at the rate of about one per year. These ruptures have caused complex plant transients which have not always been easy for the reactor operators to control. Our analysis shows that if more than 15 tubes rupture during a main steam line break, the system response could lead to core melting. Although spontaneous and induced steam generator tube ruptures are small contributors to the total core damage frequency calculated in probabilistic risk assessments, they are risk significant because the radionuclides are likely to bypass the reactor containment building. The frequency of steam generator tube ruptures can be significantly reduced through appropriate and timely inspections and repairs or removal from service

  7. Rupture of a high pressure gas or steam pipe in a tunnel: a preliminary investigation of the jet thrust exerted on a tunnel barrier

    International Nuclear Information System (INIS)

    Baum, M.R.

    1988-04-01

    On power plant, if a high pressure pipe containing high temperature gas or steam were to rupture, sensitive equipment necessary for safety shutdown of the plant could possibly be incapacitated if exposed to the subsequent high temperature environment. In many plant configurations the high pressure pipework is contained in tunnels where it is possible to construct barriers which isolate one section of the plant from another, thereby restricting the spread of the high temperature fluid/air mixture. This paper describes a preliminary experimental investigation of the magnitude of the thrust likely to be exerted on such barriers by a gas jet issuing from the failed pipe. Measurements of the thrust exerted on a flat plate by normal impingement of a highly underexpanded gas jet are in agreement with a semi-quantitative analysis assuming conservation of the axial momentum of the jet. (author)

  8. Device for extracting steam or gas from the primary coolant line leading from a reactor pressure vessel to a straight through boiler or from the top primary boiler chamber of a water-cooled nuclear reactor

    International Nuclear Information System (INIS)

    Schatz, K.

    1982-01-01

    In such a nuclear reactor, a steam or gas cushion can form when the primary system is refilled, which can cause blocking of the natural circulation or filling of the system in the area of the hot primary coolant pipe or in the top primary boiler chamber. In order to remove such a steam or gas cushion, a ventilation pipe starting from the bend of the primary coolant line is connected to the feed pipe for introducing water into the primary system. The feed pipe is designed on the principle of the vacuum pump in the area of the opening of the ventilation pipe. There is a sub-pressure in the ventilation pipe, which makes it possible to extract the steam or gas. After mixing in the area of the opening, the steam condenses or is distributed with the gas in the primary coolant. (orig.) [de

  9. Weld region corrosion during chemical cleaning of PWR [pressurized-water reactor] steam generators: Volume 2, Tests and analyses: Final report

    International Nuclear Information System (INIS)

    Barna, J.L.; Bozeka, S.A.; Jevec, J.M.

    1987-07-01

    The potential for preferential corrosion of steam generator weld regions during chemical cleaning using the generic SGOG solvents was investigated. The investigations included development and use of a corrosion assessment test facility which measured corrosion currents in a realistic model of the steam generator geometry in the vicinity of a specific weld during a simulated chemical dissolution of sludge consisting of essentially pure magnetite. A corrosion monitoring technique was developed and qualified. In this technique free corrosion rates measured by linear polarization techniques are added to corrosion rates calculated from galvanic current measured using a zero resistance ammeter to give an estimate of total corrosion rate for a galvanically corroding material. An analytic modeling technique was developed and proved useful in determining the size requirements for the weld region mockup used in the corrosion assessment test facility. The technique predicted galvanic corrosion rates consistent with that observed in a corrosion assessement test when polarization data used as model input were obtained on-line during the test. The test results obtained during this investigation indicated that chemical cleaning using the SGOG magnetite dissolution solvent can be performed with a small amount of corrosion of secondary side internals and pressure boundary welds. The maximum weld region corrosion measured during a typical chemical cleaning cycle to remove essentially pure magnetite sludge was about 8 mils. However, additional site specific weld region corrosion assessment testing and qualification will be required prior to chemical cleaning steam generators at a specific plant. Recommendations for site specific qualification of chemical cleaning processes and for use of process monitors and on-line corrosion instrumentation are included in this report

  10. Kinetics of reactions of oxidation of carbon by carbon dioxide and water steam at high temperatures and low pressures

    International Nuclear Information System (INIS)

    Boulangier, Francois

    1956-01-01

    The first objective of this research thesis was to obtain new and reliable experimental results about the reaction kinetics of the oxidation of carbon by carbon dioxide and water steam, and to avoid some disturbing phenomena, for example and more particularly the appearance of electric discharges beyond 1900 K initiated by the filament thermoelectric emission. The author tried to identify the mechanism of the accelerating effect. It appears that previous experiments had been performed only in these disturbed conditions. At the lowest temperatures, the author highlighted the existence of a surface contamination by the desorption products from the apparatus [fr

  11. Steam 80 steam generator instrumentation

    International Nuclear Information System (INIS)

    Carson, W.H.; Harris, H.H.

    1980-01-01

    This paper describes two special instrumentation packages in an integral economizer (preheater) steam generator of one of the first System 80 plants scheduled to go into commercial operation. The purpose of the instrumentation is to obtain accurate operating information from regions of the secondary side of the steam generator inaccessible to normal plant instrumentation. In addition to verification of the System 80 steam generator design predictions, the data obtained will assist in verification of steam generator thermal/hydraulic computer codes developed for generic use in the industry

  12. Selling steam

    International Nuclear Information System (INIS)

    Zimmer, M.J.; Goodwin, L.M.

    1991-01-01

    This article addresses the importance of steam sales contract is in financing cogeneration facilities. The topics of the article include the Public Utility Regulatory Policies Act provisions and how they affect the marketing of steam from qualifying facilities, the independent power producers market shift, and qualifying facility's benefits

  13. Nineteen-Foot Diameter Explosively Driven Blast Simulator; TOPICAL

    International Nuclear Information System (INIS)

    VIGIL, MANUEL G.

    2001-01-01

    This report describes the 19-foot diameter blast tunnel at Sandia National Laboratories. The blast tunnel configuration consists of a 6 foot diameter by 200 foot long shock tube, a 6 foot diameter to 19 foot diameter conical expansion section that is 40 feet long, and a 19 foot diameter test section that is 65 feet long. Therefore, the total blast tunnel length is 305 feet. The development of this 19-foot diameter blast tunnel is presented. The small scale research test results using 4 inch by 8 inch diameter and 2 foot by 6 foot diameter shock tube facilities are included. Analytically predicted parameters are compared to experimentally measured blast tunnel parameters in this report. The blast tunnel parameters include distance, time, static, overpressure, stagnation pressure, dynamic pressure, reflected pressure, shock Mach number, flow Mach number, shock velocity, flow velocity, impulse, flow duration, etc. Shadowgraphs of the shock wave are included for the three different size blast tunnels

  14. Porcine head response to blast.

    Science.gov (United States)

    Shridharani, Jay K; Wood, Garrett W; Panzer, Matthew B; Capehart, Bruce P; Nyein, Michelle K; Radovitzky, Raul A; Bass, Cameron R 'dale'

    2012-01-01

    Recent studies have shown an increase in the frequency of traumatic brain injuries related to blast exposure. However, the mechanisms that cause blast neurotrauma are unknown. Blast neurotrauma research using computational models has been one method to elucidate that response of the brain in blast, and to identify possible mechanical correlates of injury. However, model validation against experimental data is required to ensure that the model output is representative of in vivo biomechanical response. This study exposes porcine subjects to primary blast overpressures generated using a compressed-gas shock tube. Shock tube blasts were directed to the unprotected head of each animal while the lungs and thorax were protected using ballistic protective vests similar to those employed in theater. The test conditions ranged from 110 to 740 kPa peak incident overpressure with scaled durations from 1.3 to 6.9 ms and correspond approximately with a 50% injury risk for brain bleeding and apnea in a ferret model scaled to porcine exposure. Instrumentation was placed on the porcine head to measure bulk acceleration, pressure at the surface of the head, and pressure inside the cranial cavity. Immediately after the blast, 5 of the 20 animals tested were apneic. Three subjects recovered without intervention within 30 s and the remaining two recovered within 8 min following respiratory assistance and administration of the respiratory stimulant doxapram. Gross examination of the brain revealed no indication of bleeding. Intracranial pressures ranged from 80 to 390 kPa as a result of the blast and were notably lower than the shock tube reflected pressures of 300-2830 kPa, indicating pressure attenuation by the skull up to a factor of 8.4. Peak head accelerations were measured from 385 to 3845 G's and were well correlated with peak incident overpressure (R(2) = 0.90). One SD corridors for the surface pressure, intracranial pressure (ICP), and head acceleration are

  15. Brain injuries from blast.

    Science.gov (United States)

    Bass, Cameron R; Panzer, Matthew B; Rafaels, Karen A; Wood, Garrett; Shridharani, Jay; Capehart, Bruce

    2012-01-01

    Traumatic brain injury (TBI) from blast produces a number of conundrums. This review focuses on five fundamental questions including: (1) What are the physical correlates for blast TBI in humans? (2) Why is there limited evidence of traditional pulmonary injury from blast in current military field epidemiology? (3) What are the primary blast brain injury mechanisms in humans? (4) If TBI can present with clinical symptoms similar to those of Post-Traumatic Stress Disorder (PTSD), how do we clinically differentiate blast TBI from PTSD and other psychiatric conditions? (5) How do we scale experimental animal models to human response? The preponderance of the evidence from a combination of clinical practice and experimental models suggests that blast TBI from direct blast exposure occurs on the modern battlefield. Progress has been made in establishing injury risk functions in terms of blast overpressure time histories, and there is strong experimental evidence in animal models that mild brain injuries occur at blast intensities that are similar to the pulmonary injury threshold. Enhanced thoracic protection from ballistic protective body armor likely plays a role in the occurrence of blast TBI by preventing lung injuries at blast intensities that could cause TBI. Principal areas of uncertainty include the need for a more comprehensive injury assessment for mild blast injuries in humans, an improved understanding of blast TBI pathophysiology of blast TBI in animal models and humans, the relationship between clinical manifestations of PTSD and mild TBI from blunt or blast trauma including possible synergistic effects, and scaling between animals models and human exposure to blasts in wartime and terrorist attacks. Experimental methodologies, including location of the animal model relative to the shock or blast source, should be carefully designed to provide a realistic blast experiment with conditions comparable to blasts on humans. If traditional blast scaling is

  16. What is geothermal steam worth?

    International Nuclear Information System (INIS)

    Thorhallsson, S.; Ragnarsson, A.

    1992-01-01

    Geothermal steam is obtained from high-temperature boreholes, either directly from the reservoir or by flashing. The value of geothermal steam is similar to that of steam produced in boilers and lies in its ability to do work in heat engines such as turbines and to supply heat for a wide range of uses. In isolated cases the steam can be used as a source of chemicals, for example the production of carbon dioxide. Once the saturated steam has been separated from the water, it can be transported without further treatment to the end user. There are several constraints on its use set by the temperature of the reservoir and the chemical composition of the reservoir fluid. These constraints are described (temperature of steam, scaling in water phase, gas content of steam, well output) as are the methods that have been adopted to utilize this source of energy successfully. Steam can only be transported over relatively short distances (a few km) and thus has to be used close to the source. Examples are given of the pressure drop and sizing of steam mains for pipelines. The path of the steam from the reservoir to the end user is traced and typical cost figures given for each part of the system. The production cost of geothermal steam is estimated and its sensitivity to site-specific conditions discussed. Optimum energy recovery and efficiency is important as is optimizing costs. The paper will treat the steam supply system as a whole, from the reservoir to the end user, and give examples of how the site-specific conditions and system design have an influence on what geothermal steam is worth from the technical and economic points of view

  17. Facility to separate water and steam

    International Nuclear Information System (INIS)

    Loesel, G.

    1977-01-01

    The water/steam mixture from the pressure vessel e.g. of a BWR is separated by means of centrifugal separators untilizing the natural separation of steam. The steam is supplied to a steam drying vessel and the water to a water collecting tank. These vessels may be combined to a common vessel or connected through additional pipes. From the water collecting tank, arranged below the steam dryer, a feedwater pipe runs back to the pressure vessel. By construction out of individual components cleaning, decontamination, and operating control are essentially simplified. (RW) 891 RW [de

  18. Low pressure steam expansion pretreatment as a competitive approach to improve diosgenin yield and the production of fermentable sugar from Dioscorea zingiberensis C.H. Wright.

    Science.gov (United States)

    Wei, Mi; Tong, Yao; Wang, Hongbo; Wang, Lihua; Yu, Longjiang

    2016-04-01

    Development of efficient pretreatment methods which can disrupt the peripheral lignocellulose and even the parenchyma cells is of great importance for production of diosgenin from turmeric rhizomes. It was found that low pressure steam expansion pretreatment (LSEP) could improve the diosgenin yield by more than 40% compared with the case without pretreatment, while simultaneously increasing the production of fermentable sugar by 27.37%. Furthermore, little inhibitory compounds were produced in LSEP process which was extremely favorable for the subsequent biotransformation of fermentable sugar to other valuable products such as ethanol. Preliminary study showed that the ethanol yield when using the fermentable sugar as carbon source was comparable to that using glucose. The liquid residue of LSEP treated turmeric tuber after diosgenin production can be utilized as a quality fermentable carbon source. Therefore, LSEP has great potential in industrial application in diosgenin clean production and comprehensive utilization of turmeric tuber. Copyright © 2016 Elsevier Ltd. All rights reserved.

  19. Nitrided steel with increased reliability for steam turbine blades of low pressure cylinders; Vysokoazotistaya stal` s povishennoj nadezhdnostni dlya lopatok ha tsilindrov niskogo davleniya parnikh turbin

    Energy Technology Data Exchange (ETDEWEB)

    Andreev, Ch; Lengarski, P [Bylgarska Akademiya na Naukite, Sofia (Bulgaria). Inst. po Metaloznanie i Tekhnologiya na Metalite

    1996-12-31

    A new type of steel has been developed, containing 0.11-0.20% N and less than 0.05% C, the sum of both components being within the range 0.16-0.26%. The metal has an austenite-martensite structure with 10-30% austenite content. Samples obtained by counter-pressure casting have been investigated with respect to the influence of the thermal treatment on mechanical properties. The best properties are obtained when applying hardening by heating at 1050{sup o} C and cooling at 550{sup o} C: fluidity limit R{sub 0}.2>=850 MPa, relative elongation A>=15%, relative shortening Z>=50%, impact viscosity KCU >= 588 kJ/m{sup 2} at critical temperature of brittleness <-40{sup o} C. These properties are combined with high corrosion and wear resistance and make the steel suitable for steam turbine blades. 5 refs., 2 figs., 4 tabs.

  20. Horizontal steam generator thermal-hydraulics

    Energy Technology Data Exchange (ETDEWEB)

    Ubra, O. [SKODA Praha Company, Prague (Czechoslovakia); Doubek, M. [Czech Technical Univ., Prague (Czechoslovakia)

    1995-09-01

    Horizontal steam generators are typical components of nuclear power plants with pressure water reactor type VVER. Thermal-hydraulic behavior of horizontal steam generators is very different from the vertical U-tube steam generator, which has been extensively studied for several years. To contribute to the understanding of the horizontal steam generator thermal-hydraulics a computer program for 3-D steady state analysis of the PGV-1000 steam generator has been developed. By means of this computer program, a detailed thermal-hydraulic and thermodynamic study of the horizontal steam generator PGV-1000 has been carried out and a set of important steam generator characteristics has been obtained. The 3-D distribution of the void fraction and 3-D level profile as functions of load and secondary side pressure have been investigated and secondary side volumes and masses as functions of load and pressure have been evaluated. Some of the interesting results of calculations are presented in the paper.

  1. Analysis of steam-generator tube-rupture events combined with auxiliary-feedwater control-system failure for Three Mile Island-Unit 1 and Zion-Unit 1 pressurized water reactors

    International Nuclear Information System (INIS)

    Nassersharif, B.

    1986-01-01

    A steam-generator tube-rupture (SGTR) event combined with loss of all offsite alternating-current power and failure of the auxiliary-feedwater (AFW) control system has been investigated for the Three Mile Island-Unit 1 (TMI-1) and Zion-Unit 1 (Zion-1) pressurized water reactors. The Transient Reactor Analysis Code was used to simulate the accident sequence for each plant. The objectives of the study were to predict the plant transient response with respect to tube-rupture flow termination, extent of steam generator overfill, and thermal-hydraulic conditions in the steam lines. Two transient cases were calculated: (1) a TMI-1 SGTR and runaway-AFW transient, and (2) a Zion-1 SGTR and runaway-AFW transient. Operator actions terminated the tube-rupture flow by 1342 s (22.4 min) and 1440 s (24.0 min) for TMI-1 and Zion-1, respectively, but AFW injection was continued. The damaged steam generator (DSG) overfilled by 1273 s (21.2 min) for the TMI-1 calculation and by 1604 s (26.7 min) for the Zion-1 calculation. The DSG steam lines were completely filled by 1500 s (25 min) and 2000 s (33.3 min) for TMI-1 and Zion-1, respectively. The maximum subcooling in the steam lines was approx.63 K (approx.113 0 F) for TMI-1 and approx.44 K (approx.80 0 F) for Zion-1

  2. Friction pressure drop and heat transfer coefficient of two-phase flow in helically coiled tube once-through steam generator for integrated type marine water reactor

    International Nuclear Information System (INIS)

    Nariai, Hideki; Kobayashi, Michiyuki; Matsuoka, Takeshi.

    1982-01-01

    Two-phase friction pressure drop and heat transfer coefficients in a once-through steam generator with helically coiled tubes were investigated with the model test rig of an integrated type marine water reactor. As the dimensions of the heat transfer tubes and the thermal-fluid conditions are almost the same as those of real reactors, the data applicable directly to the real reactor design were obtained. As to the friction pressure drop, modified Kozeki's prediction which is based on the experimental data by Kozeki for coiled tubes, agreed the best with the experimental data. Modified Martinelli-Nelson's prediction which is based on Martinelli-Nelson's multiplier using Ito's equation for single-phase flow in coiled tube, agreed within 30%. The effect of coiled tube on the average heat transfer coefficients at boiling region were small, and the predictions for straight tube could also be applied to coiled tube. Schrock-Grossman's correlation agreed well with the experimental data at the pressures of lower than 3.5 MPa. It was suggested that dryout should be occurred at the quality of greater than 90% within the conditions of this report. (author)

  3. 46 CFR 61.15-5 - Steam piping.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 2 2010-10-01 2010-10-01 false Steam piping. 61.15-5 Section 61.15-5 Shipping COAST... Periodic Tests of Piping Systems § 61.15-5 Steam piping. (a) Main steam piping shall be subjected to a... removed and the piping thoroughly examined. (b) All steam piping subject to pressure from the main boiler...

  4. Primary blast survival and injury risk assessment for repeated blast exposures.

    Science.gov (United States)

    Panzer, Matthew B; Bass, Cameron R Dale; Rafaels, Karin A; Shridharani, Jay; Capehart, Bruce P

    2012-02-01

    The widespread use of explosives by modern insurgents and terrorists has increased the potential frequency of blast exposure in soldiers and civilians. This growing threat highlights the importance of understanding and evaluating blast injury risk and the increase of injury risk from exposure to repeated blast effects. Data from more than 3,250 large animal experiments were collected from studies focusing on the effects of blast exposure. The current study uses 2,349 experiments from the data collection for analysis of the primary blast injury and survival risk for both long- and short-duration blasts, including the effects from repeated exposures. A piecewise linear logistic regression was performed on the data to develop survival and injury risk assessment curves. New injury risk assessment curves uniting long- and short-duration blasts were developed for incident and reflected pressure measures and were used to evaluate the risk of injury based on blast over pressure, positive-phase duration, and the number of repeated exposures. The risk assessments were derived for three levels of injury severity: nonauditory, pulmonary, and fatality. The analysis showed a marked initial decrease in injury tolerance with each subsequent blast exposure. This effect decreases with increasing number of blast exposures. The new injury risk functions showed good agreement with the existing experimental data and provided a simplified model for primary blast injury risk. This model can be used to predict blast injury or fatality risk for single exposure and repeated exposure cases and has application in modern combat scenarios or in setting occupational health limits. .Copyright © 2012 by Lippincott Williams & Wilkins

  5. Characteristic and kinetics of corn stalk pyrolysis in a high pressure reactor and steam gasification of its char

    OpenAIRE

    Zhang, Guojie; Sun, Yinghui; Shi, Yuliang; Jia, Yong; Xu, Ying; Zhao, Peiyu; Zhang, Yongfa

    2016-01-01

    Pyrolysis characteristics of the corn stalk were investigated at different temperatures and pressures, with focus mainly on the releasing profiles and forming behaviors of gas. The results show that a higher pressure was conducive to the yield of H2 and CH4, and was not conducive to the generation of CO. Combustible gas components increase with the increase temperature. H2 concentration was affected significantly by temperature, was mainly released at higher temperatures (>500 °C). At 700 °C,...

  6. Analysis of multiple-tube ruptures in both steam generators for the Three Mile Island-1 pressurized water reactor

    International Nuclear Information System (INIS)

    Nassersharif, B.

    1985-01-01

    The operator guidelines were followed for both transients described. Both transients resulted in SG overfill and the tube-rupture flow did not terminate in either transient. The following statements can be deducted from the results of the calculations: the tube-rupture flow could not be stopped for either case during 2600 s (43 min) of transient time; each accident scenario resulted in SG overfill; both SGs overfilled by 1600 s (27 min) and 1800 s (30 min) for Cases 1 and 2, respectively; conditions for isolation of the SGs were not reached; and core subcooling was not lost in either case but the upper head was voided in Case 2. Comparison of the cooldown rates in the two cases after 1200 s (20 min) shows that these rates are equal (i.e., restart of the RCPs did not change the primary-system cooldown rate). However, in Case 2, a steam bubble was formed in the upper head, which did not disappear during the simulated time. One of the immediate actions in the guidelines was to fill both SGs to 95% level. This step was almost unnecessary because the tube-rupture flow was large enough that the 15.4 - mK/s (100 - 0 F h) cooldown-rate limit was exceeded and AFW could not be injected. Also, the guidelines did not address the SG overfill issue

  7. A theoretical critical heat flux model for low-pressure, low-mass-flux, and low-steam quality conditions

    International Nuclear Information System (INIS)

    Weihsiao Ho; Kuanchywan Tu; Baushei Pei; Chinjang Chang

    1993-01-01

    The critical heat flux (CHF) is the maximum heat flux just before a boiling crisis; its importance as a measurement of nuclear reactor power capability design as well as in the safety of reactors has been recognized. With emphasis on CHF behavior under subcooled and low-quality (i.e., 2 ·s), an improved model that uses the sublayer dry out theory has been developed. Based on experimental observations of CHF, the model assumes that CHF under such conditions is of the departure from nucleate boiling type. Based on the postulation that CHF is triggered by Helmholtz instability in the sublayer steam-liquid system, the model was developed by a simple energy balance of liquid sublayer evaporation as the vapor blanket tends to disturb the balance between the buoyancy force and the drag force exerted upon it. The model is compared with the well-known Biasi et al. correlation as well as the Atomic Energy of Canada Limited lookup table against 102 uniformly heated round tube CHF data and 34 nonuniformly heated round tube CHF data. The comparison shows that the model provides better accuracy and a reasonable agreement between the predicted values and experimental CHF data

  8. Containments for consolidated nuclear steam systems

    International Nuclear Information System (INIS)

    Jabsen, F.S.

    1978-01-01

    A containment system for a consolidated nuclear steam system incorporating a nuclear core, steam generator and reactor coolant pumps within a single pressure vessel is described which is designed to provide radiation shielding and pressure suppression. Design details, including those for the dry well and wet well of the containment, are given. (UK)

  9. The testing of a steam-water separating device used for vertical steam generators

    International Nuclear Information System (INIS)

    Ding Xunshen; Cui Baoyuan; Xue Yunkui; Liu Shixun

    1989-01-01

    The air-water screening tests of a steam-water separating device used for vertical steam generators at low pressure are introduced. The article puts emphasis on the qualification test of the steam-water separating device at hot conditions in a high temperature and pressure water test rig. The performance of the comprehensive test of the steam-water separating device indicates that the humidity of the steam at the drier exit is much less than the specified amount of 0.25%

  10. Numerical analysis of blast flow-field of baffle type muzzle brake

    Energy Technology Data Exchange (ETDEWEB)

    Kim, D.H. [Graduate School, Chungnam National University, Taejon (Korea); Ko, S. [Chungnam National University, Taejon (Korea)

    1998-11-01

    A three-dimensional unsteady, inviscid blast flow-field of a baffle type muzzle brake has been simulated by solving the Euler equation. The blast flow-field includes the effect of the free air blast, precursor blast flow and the propellant blast gas flow. Chimera grid scheme was used to generate 9 multi-block volume grids for the complex geometry. The evolution of the blast flow-field is presented by showing the contours of pressure, density and Mach number for certain time step. The comparison of the calculated and measured peak pressures on the surfaces of the muzzle brake is also presented. (author). 4 refs., 5 figs., 1 tab.

  11. Model of reverse steam generator

    International Nuclear Information System (INIS)

    Malasek, V.; Manek, O.; Masek, V.; Riman, J.

    1987-01-01

    The claim of Czechoslovak discovery no. 239272 is a model designed for the verification of the properties of a reverse steam generator during the penetration of water, steam-water mixture or steam into liquid metal flowing inside the heat exchange tubes. The design may primarily be used for steam generators with a built-in inter-tube structure. The model is provided with several injection devices configured in different heat exchange tubes, spaced at different distances along the model axis. The design consists in that between the pressure and the circumferential casings there are transverse partitions and that in one chamber consisting of the circumferential casings, pressure casing and two adjoining partitions there is only one passage of the injection device through the inter-tube space. (Z.M.). 1 fig

  12. Steam generators in PWR's

    International Nuclear Information System (INIS)

    Michel, R.

    1974-01-01

    The steam generator of the PWR operates according to the principle of natural circulation. It consists of a U-shaped tube bundle whose free ends are welded to a bottom plate. The tube bundle is surrounded by a cylinder jacket which has slots closely above the bottom or tube plate. The feed water mixed with boiling water enters the tube bundle through these slots. Because of its buoyancy, the steam-water mixture flows upwards. Below the tube plate there are chambers for distributing and collecting pressurized water separated by means of a partition wall. By omitting some tubes, a free alloy is created so that the tubes in the center get sufficient water, too. By asymmetrical arrangement of the partition wall it is further possible to limit the tube alloy only to the inlet side for pressurized water. The flow over the tube plate is thus improved on the inlet side. (DG) [de

  13. An experimental study of high pressure steam condensation in a vertical tube of passive secondary condensation system

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sang Jae; No, Hee Cheon [KAIST, Taejon (Korea, Republic of)

    1998-07-01

    To investigate the physical parameters of PSCS (Passive Secondary Condensation System) which is a passive residual heat removal system of CP-1300, the high pressure condensation experiments are performed in a small scale experimental facility. The experimental parameters are the local heat flux and the transfer coefficient and the pressure drop in a condensation heat trasnfer. The film condensation heat transfer coefficients in a vertical tube are calculated from the measured wall temperature difference and compared with the analytical models. A new analytical condensation model is developed based on the annular film flow model. The present model gives marginally better results than those from the Shah model in comparison with the experimental data in the database. Also, experimental data are compared with the results of the RELAP5/MOD3.2 thermal hydraulic code. The RELAP5/MOD3.2 underpredicts the condensation heat transfer coefficients of the present experiment by 50 %.

  14. Experimental study of centrifugal pump performance under steam-water two-phase flow conditions at elevated pressures

    International Nuclear Information System (INIS)

    Chan, A.M.C.; Barreca, S.L.; Hartlen, R.T.

    1991-01-01

    The performance of a centrifugal pump under two-phase flow conditions was studied in a closed loop. System voids of increasing magnitude were attained by draining water from the loop in steps. The operating temperature/pressure were varied from 110 degrees C/0.15 MPa to 260 degrees C/4.7 MPa. Only tests in the first quadrant were conducted. In this paper the head-flow characteristics and pump head degradation data are presented and discussed

  15. A computational model of blast loading on the human eye.

    Science.gov (United States)

    Bhardwaj, Rajneesh; Ziegler, Kimberly; Seo, Jung Hee; Ramesh, K T; Nguyen, Thao D

    2014-01-01

    Ocular injuries from blast have increased in recent wars, but the injury mechanism associated with the primary blast wave is unknown. We employ a three-dimensional fluid-structure interaction computational model to understand the stresses and deformations incurred by the globe due to blast overpressure. Our numerical results demonstrate that the blast wave reflections off the facial features around the eye increase the pressure loading on and around the eye. The blast wave produces asymmetric loading on the eye, which causes globe distortion. The deformation response of the globe under blast loading was evaluated, and regions of high stresses and strains inside the globe were identified. Our numerical results show that the blast loading results in globe distortion and large deviatoric stresses in the sclera. These large deviatoric stresses may be indicator for the risk of interfacial failure between the tissues of the sclera and the orbit.

  16. Steam generators

    International Nuclear Information System (INIS)

    Hayden, R.L.J.

    1979-01-01

    Steam generators for nuclear reactors are designed so that deposition of solids on the surface of the inlet side of the tubesheet or the inlet header with the consequent danger of corrosion and eventual tube failure is obviated or substantially reduced. (U.K.)

  17. Steam generation: fossil-fired systems: utility boilers; industrial boilers; boiler auxillaries; nuclear systems: boiling water; pressurized water; in-core fuel management; steam-cycle systems: condensate/feedwater; circulating water; water treatment

    International Nuclear Information System (INIS)

    Anon.

    1982-01-01

    A survey of development in steam generation is presented. First, fossil-fired systems are described. Progress in the design of utility and industrial boilers as well as in boiler auxiliaries is traced. Improvements in coal pulverizers, burners that cut pollution and improve efficiency, fans, air heaters and economisers are noted. Nuclear systems are then described, including the BWR and PWR reactors, in-core fuel management techniques are described. Finally, steam-cycle systems for fossil-fired and nuclear power plants are reviewed. Condensate/feedwater systems, circulating water systems, cooling towers, and water treatment systems are discussed

  18. Process and device for accelerating condensation of the steam produced during an accident from the pressure vessel of a water cooled nuclear reactor

    International Nuclear Information System (INIS)

    Schnitker, W.

    1980-01-01

    In case of an accident, the steam from the PWR is taken away via lances under the water surface of the condensation area. In order to accelerate condensation, water is added via pipes projecting sideways into the lances. The kinetic energy of the steam carries the water over and produces a fog. (DG) [de

  19. Process and device for accelerating condensation of the steam produced during an accident from the pressure vessel of a water cooled nuclear reactor

    International Nuclear Information System (INIS)

    Schnitker, W.

    1981-01-01

    In case of an accident, the steam from the PWR is taken away via lances under the water surface of the condensation area. In order to accelerate condensation, water is added via pipes projecting sideways into the lances. The kinetic energy of the steam carries the water over and produces a fog. (orig./PW)

  20. Periodical in-service inspection as part of individual program of quality assurance of steam generators and pressurizers of WWER 440 nuclear power plant

    International Nuclear Information System (INIS)

    Kawalec, M.

    1982-01-01

    The manufacturers of equipment for nuclear power plants in the Czechoslovak Socialist Republic are obligated to process so-called individual programs of quality assurance in order to secure the quality of selected equipment in nuclear power. These programmes should include the evaluation of the design of the individual equipments with regard to the implementation of in-service inspection. The main problems are discussed related to the processing of the program of quality assurance for the steam generator and pressurizer. To solve these problems it is necessary that the general project designer should make a classification of the components according to safety categories and that the manufacturers should determine the weak points of the design on the basis of an analysis of the design of individual component nodes. On the basis of such an analysis it is then necessary to evaluate the existing design of the scale of in-service inspections and to decide whether or not new inspection methods should be added. (Z.M.)

  1. Low-cost blast wave generator for studies of hearing loss and brain injury: blast wave effects in closed spaces.

    Science.gov (United States)

    Newman, Andrew J; Hayes, Sarah H; Rao, Abhiram S; Allman, Brian L; Manohar, Senthilvelan; Ding, Dalian; Stolzberg, Daniel; Lobarinas, Edward; Mollendorf, Joseph C; Salvi, Richard

    2015-03-15

    Military personnel and civilians living in areas of armed conflict have increased risk of exposure to blast overpressures that can cause significant hearing loss and/or brain injury. The equipment used to simulate comparable blast overpressures in animal models within laboratory settings is typically very large and prohibitively expensive. To overcome the fiscal and space limitations introduced by previously reported blast wave generators, we developed a compact, low-cost blast wave generator to investigate the effects of blast exposures on the auditory system and brain. The blast wave generator was constructed largely from off the shelf components, and reliably produced blasts with peak sound pressures of up to 198dB SPL (159.3kPa) that were qualitatively similar to those produced from muzzle blasts or explosions. Exposure of adult rats to 3 blasts of 188dB peak SPL (50.4kPa) resulted in significant loss of cochlear hair cells, reduced outer hair cell function and a decrease in neurogenesis in the hippocampus. Existing blast wave generators are typically large, expensive, and are not commercially available. The blast wave generator reported here provides a low-cost method of generating blast waves in a typical laboratory setting. This compact blast wave generator provides scientists with a low cost device for investigating the biological mechanisms involved in blast wave injury to the rodent cochlea and brain that may model many of the damaging effects sustained by military personnel and civilians exposed to intense blasts. Copyright © 2015 Elsevier B.V. All rights reserved.

  2. Steam generator tube performance

    International Nuclear Information System (INIS)

    Tatone, O.S.; Tapping, R.L.; Stipan, L.

    1992-03-01

    A survey of steam generator operating experience for 1986 has been carried out for 184 pressurized water and pressurized heavy-water reactors, and 1 water-cooled, graphite-moderated reactor. Tubes were plugged at 75 of the reactors (40.5%). In 1986, 3737 tubes were plugged (0.14% of those in service) and 3148 tubes were repaired by sleeving. A small number of reactors accounted for the bulk of the plugged tubes, a phenomenon consistent with previous years. For 1986, the available tubesheet sludge data for 38 reactors has been compiled into tabular form, and sludge/deposit data will be incorporated into all future surveys

  3. Erosion corrosion in wet steam

    International Nuclear Information System (INIS)

    Tavast, J.

    1988-03-01

    The effect of different remedies against erosion corrosion in wet steam has been studied in Barsebaeck 1. Accessible steam systems were inspected in 1984, 1985 and 1986. The effect of hydrogen peroxide injection of the transport of corrosion products in the condensate and feed water systems has also been followed through chemical analyses. The most important results of the project are: - Low alloy chromium steels with a chromium content of 1-2% have shown excellent resistance to erosion corrosion in wet steam. - A thermally sprayed coating has shown good resistance to erosion corrosion in wet steam. In a few areas with restricted accessibility minor attacks have been found. A thermally sprayed aluminium oxide coating has given poor results. - Large areas in the moisture separator/reheater and in steam extraction no. 3 have been passivated by injection of 20 ppb hydrogen peroxide to the high pressure steam. In other inspected systems no significant effect was found. Measurements of the wall thickness in steam extraction no. 3 showed a reduced rate of attack. - The injection of 20 ppb hydrogen peroxide has not resulted in any significant reduction of the iron level result is contrary to that of earlier tests. An increase to 40 ppb resulted in a slight decrease of the iron level. - None of the feared disadvantages with hydrogen peroxide injection has been observed. The chromium and cobalt levels did not increase during the injection. Neither did the lifetime of the precoat condensate filters decrease. (author)

  4. Coolant Mixing in a Pressurized Water Reactor: Deboration Transients, Steam-Line Breaks, and Emergency Core Cooling Injection

    International Nuclear Information System (INIS)

    Prasser, Horst-Michael; Grunwald, Gerhard; Hoehne, Thomas; Kliem, Soeren; Rohde, Ulrich; Weiss, Frank-Peter

    2003-01-01

    The reactor transient caused by a perturbation of boron concentration or coolant temperature at the inlet of a pressurized water reactor (PWR) depends on the mixing inside the reactor pressure vessel (RPV). Initial steep gradients are partially lessened by turbulent mixing with coolant from the unaffected loops and with the water inventory of the RPV. Nevertheless the assumption of an ideal mixing in the downcomer and the lower plenum of the reactor leads to unrealistically small reactivity inserts. The uncertainties between ideal mixing and total absence of mixing are too large to be acceptable for safety analyses. In reality, a partial mixing takes place. For realistic predictions it is necessary to study the mixing within the three-dimensional flow field in the complicated geometry of a PWR. For this purpose a 1:5 scaled model [the Rossendorf Coolant Mixing Model (ROCOM) facility] of the German PWR KONVOI was built. Compared to other experiments, the emphasis was put on extensive measuring instrumentation and a maximum of flexibility of the facility to cover as much as possible different test scenarios. The use of special electrode-mesh sensors together with a salt tracer technique provided distributions of the disturbance within downcomer and core entrance with a high resolution in space and time. Especially, the instrumentation of the downcomer gained valuable information about the mixing phenomena in detail. The obtained data were used to support code development and validation. Scenarios investigated are the following: (a) steady-state flow in multiple coolant loops with a temperature or boron concentration perturbation in one of the running loops, (b) transient flow situations with flow rates changing with time in one or more loops, such as pump startup scenarios with deborated slugs in one of the loops or onset of natural circulation after boiling-condenser-mode operation, and (c) gravity-driven flow caused by large density gradients, e.g., mixing of cold

  5. Wood Modification at High Temperature and Pressurized Steam: a Relational Model of Mechanical Properties Based on a Neural Network

    Directory of Open Access Journals (Sweden)

    Hong Yang

    2015-07-01

    Full Text Available Thermally modified wood has high dimensional stability and biological durability.But if the process parameters of thermal modification are not appropriate, then there will be a decline in the physical properties of wood.A neural network algorithm was employed in this study to establish the relationship between the process parameters of high-temperature and high-pressure thermal modification and the mechanical properties of the wood. Three important parameters: temperature, relative humidity, and treatment time, were considered as the inputs to the neural network. Back propagation (BP neural network and radial basis function (RBF neural network models for prediction were built and compared. The comparison showed that the RBF neural network model had advantages in network structure, convergence speed, and generalization capacity. On this basis, the inverse model, reflecting the relationship between the process parameters and the mechanical properties of wood, was established. Given the desired mechanical properties of the wood, the thermal modification process parameters could be inversely optimized and predicted. The results indicated that the model has good learning ability and generalization capacity. This is of great importance for the theoretical and applicational studies of the thermal modification of wood.

  6. LTC vacuum blasting machine (concrete): Baseline report

    International Nuclear Information System (INIS)

    1997-01-01

    The LTC shot blast technology was tested and is being evaluated at Florida International University (FIU) as a baseline technology. In conjunction with FIU's evaluation of efficiency and cost, this report covers the evaluation conducted for safety and health issues. It is a commercially available technology and has been used for various projects at locations throughout the country. The LTC 1073 Vacuum Blasting Machine uses a high-capacity, direct-pressure blasting system which incorporates a continuous feed for the blast media. The blast media cleans the surface within the contained brush area of the blast. It incorporates a vacuum system which removes dust and debris from the surface as it is blasted. The safety and health evaluation during the testing demonstration focused on two main areas of exposure: dust and noise. Dust exposure during maintenance activities was minimal, but due to mechanical difficulties dust monitoring could not be conducted during operation. Noise exposure was significant. Further testing for each of these exposures is recommended because of the outdoor environment where the testing demonstration took place. This may cause the results to be inaccurate. It is feasible that the dust and noise levels will be higher in an enclosed environment. In addition, other safety and health issues found were ergonomics, heat stress, tripping hazards, electrical hazards, lockout/tagout, and arm-hand vibration

  7. Calculation of driling and blasting parameters in blasting performance

    OpenAIRE

    Dambov, Risto; Karanakova Stefanovska, Radmila; Dambov, Ilija

    2015-01-01

    In all mining technology drilling and blasting parameters and works are one of the main production processes at each mine. The parameters of drilling and blasting and explosives consumption per ton of blasting mass are define economic indicators of any blasting no matter for what purpose and where mining is performed. The calculation of rock blasting should always have in mind that the methodology of calculation of all drilling and blasting parameters in blasting performance are performed for...

  8. DYNAMIC TIME HISTORY ANALYSIS OF BLAST RESISTANT DOOR USING BLAST LOAD MODELED AS IMPACT LOAD

    Directory of Open Access Journals (Sweden)

    Y. A. Pranata

    2012-06-01

    Full Text Available A blast resistant single door was designed to withstand a 0.91 bar blast pressure and 44 ms blast duration. The analysis was done using Dynamic Time History Analysis using Blast Load modeled as Impact Load for given duration. The material properties used have been modified to accommodate dynamic effects. The analysis was done using dynamic finite element method (fem for time of the blast duration, and the maximum/minimum internal forces and displacement were taken from the time history output, in order to know the behavior under blast load and estimate the safety margin of the door. Results obtained from this research indicated that the maximum z-displacement is 1.709 mm, while in the term of serviceability, the permitted is 25 mm. The maximum reaction force is 73,960 N, while the maximum anchor capacity is 82,069 N. On blast condition, the maximum frame stress is 71.71 MPa, the maximum hinge shear stress is 45.28 MPa. While on rebound condition, the maximum frame stress is 172.11 MPa, the maximum hinge shear stress is 29.46 MPa. The maximum door edge rotation is 0.44 degree, which is not exceed the permitted boundary (1.2 degree. Keywords: Dynamic time history, blast resistant door, single door, finite element method.

  9. Plant for the delivery of long-distance steam combined with a nuclear power plant

    International Nuclear Information System (INIS)

    Schueller, K.H.

    1976-01-01

    It is proposed that long-distance steam should not be directly discharged in order to avoid each posibility of spreading radioactively contaminated steam. As a heat transmitter, a surface heat exchanger should be chosen, the heating steam of the nuclear power station heating pressurized water whose pressure is higher then that of the heating steam. Long-distance steam generation then results from expanding the pressurized water. The plant is described in detail. (UWI) [de

  10. CANDU steam generator life management

    International Nuclear Information System (INIS)

    Tapping, R.L.; Nickerson, J.; Spekkens, P.; Maruska, C.

    1998-01-01

    Steam generators are a critical component of a nuclear power reactor, and can contribute significantly to station unavailability, as has been amply demonstrated in Pressurized Water Reactors (PWRs). CANDU steam generators are not immune to steam generator degradation, and the variety of CANDU steam generator designs and tube materials has led to some unexpected challenges. However, aggressive remedial actions, and careful proactive maintenance activities, have led to a decrease in steam generator-related station unavailability of Canadian CANDUs. AECL and the CANDU utilities have defined programs that will enable existing or new steam generators to operate effectively for 40 years. Research and development work covers corrosion and mechanical degradation of tube bundles and internals, chemistry, thermalhydraulics, fouling, inspection and cleaning, as well as provision for specially tool development for specific problem solving. A major driving force is development of CANDU-specific fitness-for-service guidelines, including appropriate inspection and monitoring technology to measure steam generator condition. Longer-range work focuses on development of intelligent on-line monitoring for the feedwater system and steam generator. New designs have reduced risk of corrosion and fouling, are more easily inspected and cleaned, and are less susceptible to mechanical damage. The Canadian CANDU utilities have developed programs for remedial actions to combat degradation of performance (Gentilly-2, Point Lepreau, Bruce A/B, Pickering A/B), and have developed strategic plans to ensure that good future operation is ensured. The research and development program, as well as operating experience, has identified where improvements in operating practices and/or designs can be made in order to ensure steam generator design life at an acceptable capacity factory. (author)

  11. An innovative modular device and wireless control system enabling thermal and pressure sensors using FPGA on real-time fault diagnostics of steam turbine functional deterioration

    Science.gov (United States)

    Devi, S.; Saravanan, M.

    2018-03-01

    It is necessary that the condition of the steam turbines is continuously monitored on a scheduled basis for the safe operation of the steam turbines. The review showed that steam turbine fault detection and operation maintenance system (STFDOMS) is gaining importance recently. In this paper, novel hardware architecture is proposed for STFDOMS that can be communicated through the GSM network. Arduino is interfaced with the FPGA so as to transfer the message. The design has been simulated using the Verilog programming language and implemented in hardware using FPGA. The proposed system is shown to be a simple, cost effective and flexible and thereby making it suitable for the maintenance of steam turbines. This system forewarns the experts to access to data messages and take necessary action in a short period with great accuracy. The hardware developed is promised as a real-time test bench, specifically for investigations of long haul effects with different parameter settings.

  12. Blast effects physical properties of shock waves

    CERN Document Server

    2018-01-01

    This book compiles a variety of experimental data on blast waves. The book begins with an introductory chapter and proceeds to the topic of blast wave phenomenology, with a discussion Rankine-Hugoniot equations and the Friedlander equation, used to describe the pressure-time history of a blast wave. Additional topics include arrival time measurement, the initiation of detonation by exploding wires, a discussion of TNT equivalency, and small scale experiments. Gaseous and high explosive detonations are covered as well. The topics and experiments covered were chosen based on the comparison of used scale sizes, from small to large. Each characteristic parameter of blast waves is analyzed and expressed versus scaled distance in terms of energy and mass. Finally, the appendix compiles a number of polynomial laws that will prove indispensable for engineers and researchers.

  13. Signal analysis of steam line acoustics

    International Nuclear Information System (INIS)

    Martin, C. Samuel

    2003-01-01

    The vibration of nuclear steam piping is usually associated with pressure fluctuations emanating from flow disturbances such as steam generator nozzles, bends, or other pipe fittings. Flow separation at pipe tees and within steam chest manifolds or headers generate pressure fluctuations that propagate both upstream to steam generators as well as downstream to the steam turbine. Steady-state acoustic oscillations at various frequencies occur within the piping, possibly exciting structural vibrations. This paper focuses on the assessment of the origin of the disturbances using signal analyses of two dynamic pressure recordings from pressure transducers located along straight runs in the steam piping. The technique involves performing the cross spectrum to two dynamic pressure signals in piping between (1) the steam generator and steam chest header, and (2) between the header and steam turbine outlet. If, at a specified frequency, no causality occurs between the two signals then the cross spectra magnitude will be negligible. Of interest here is the value of the phase between the two signals for frequencies for which the magnitude of the cross spectrum is not negligible. It is shown in the paper that the direction of the dominant waves at all frequencies can be related to the phase angle from the cross spectrum. It has to be realized that pressure waves emanating from one source such as a steam generator will propagate along uniform steam pipes with little transformation or attenuation, but will be reflected at fittings and at inlets and outlets. Hence, the eventual steady-state time record at a given location in the piping is a result of not only the disturbance, but also reflections of earlier pulsations. Cross-spectral analyses has been employed to determine the direction of the dominant acoustic waves in the piping for various frequencies for which there are signals. To prove the technique, synthetic spectra are generated comprised of harmonic waves moving both

  14. Electricity from geothermal steam

    Energy Technology Data Exchange (ETDEWEB)

    Wheatcroft, E L.E.

    1959-01-01

    The development of the power station at Wairakei geothermal field is described. Wairakei is located at the center of New Zealand's volcanic belt, which lies within a major graben which is still undergoing some degree of downfaulting. A considerable number of wells, some exceeding 610 m, have been drilled. Steam and hot water are produced from both deep and shallow wells, which produce at gauge pressures of 1.5 MPa and 0.6 MPa, respectively. The turbines are fed by low, intermediate, and high pressure mains. The intermediate pressure turbine bank was installed as a replacement for a heavy water production facility which had originally been planned for the development. Stage 1 includes a 69 MW plant, and stage 2 will bring the capacity to 150 MW. A third stage, which would bring the output up to 250 MW had been proposed. The second stage involves the installation of more high pressure steam turbines, while the third stage would be powered primarily by hot water flashing. Generation is at 11 kV fed to a two-section 500 MVA board. Each section of the board feeds through a 40 MVA transformer to a pair of 220 V transmission lines which splice into the North Island grid. Other transformers feed 400 V auxiliaries and provide local supply.

  15. Nuclear reactor steam depressurization valve

    International Nuclear Information System (INIS)

    Moore, G.L.

    1991-01-01

    This patent describes improvement in a nuclear reactor plant, an improved steam depressurization valve positioned intermediate along a steam discharge pipe for controlling the venting of steam pressure from the reactor through the pipe. The improvement comprises: a housing including a domed cover forming a chamber and having a partition plate dividing the chamber into a fluid pressure activation compartment and a steam flow control compartment, the valve housing being provided with an inlet connection and an outlet connection in the steam flow control compartment, and a fluid duct in communication with a source of fluid pressure for operating the valve; a valve set mounted within the fluid flow control compartment comprising a cylindrical section surrounding the inlet connection with one end adjoining the connection and having a radially projecting flange at the other end with a contoured extended valve sealing flange provided with an annular valve sealing member, and a valve cylinder traversing the partition plate and reciprocally movable within an opening in the partition plate with one terminal and extending into the fluid pressure activation compartment and the other terminal end extending into the steam flow control compartment coaxially aligned with the valve seat surrounding the inlet connection, the valve cylinder being surrounded by two bellow fluid seals and provided with guides to inhibit lateral movement, an end of the valve cylinder extending into the fluid flow control compartment having a radially projecting flange substantially conterminous with the valve seat flange and having a contoured surface facing and complimentary to the contoured valve seating surface whereby the two contoured valve surfaces can meet in matching relationship, thus providing a pressure actuated reciprocatable valve member for making closing contact with the valve seat and withdrawing therefrom for opening fluid flow through the valve

  16. Effectiveness of eye armor during blast loading.

    Science.gov (United States)

    Bailoor, Shantanu; Bhardwaj, Rajneesh; Nguyen, Thao D

    2015-11-01

    Ocular trauma is one of the most common types of combat injuries resulting from the interaction of military personnel with improvised explosive devices. Ocular blast injury mechanisms are complex, and trauma may occur through various injury mechanisms. However, primary blast injuries (PBI) are an important cause of ocular trauma that may go unnoticed and result in significant damage to internal ocular tissues and visual impairment. Further, the effectiveness of commonly employed eye armor, designed for ballistic and laser protection, in lessening the severity of adverse blast overpressures (BOP) is unknown. In this paper, we employed a three-dimensional (3D) fluid-structure interaction computational model for assessing effectiveness of the eye armor during blast loading on human eyes and validated results against free field blast measurements by Bentz and Grimm (2013). Numerical simulations show that the blast waves focused on the ocular region because of reflections from surrounding facial features and resulted in considerable increase in BOP. We evaluated the effectiveness of spectacles and goggles in mitigating the pressure loading using the computational model. Our results corroborate experimental measurements showing that the goggles were more effective than spectacles in mitigating BOP loading on the eye. Numerical results confirmed that the goggles significantly reduced blast wave penetration in the space between the armor and the eyes and provided larger clearance space for blast wave expansion after penetration than the spectacles. The spectacles as well as the goggles were more effective in reducing reflected BOP at higher charge mass because of the larger decrease in dynamic pressures after the impact. The goggles provided greater benefit of reducing the peak pressure than the spectacles for lower charge mass. However, the goggles resulted in moderate, sustained elevated pressure loading on the eye, that became 50-100% larger than the pressure loading

  17. How to compute the power of a steam turbine with condensation, knowing the steam quality of saturated steam in the turbine discharge

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez Albarran, Manuel Jaime; Krever, Marcos Paulo Souza [Braskem, Sao Paulo, SP (Brazil)

    2009-07-01

    To compute the power and the thermodynamic performance in a steam turbine with condensation, it is necessary to know the quality of the steam in the turbine discharge and, information of process variables that permit to identifying with high precision the enthalpy of saturated steam. This paper proposes to install an operational device that will expand the steam from high pressure point on the shell turbine to atmosphere, both points with measures of pressure and temperature. Arranging these values on the Mollier chart, it can be know the steam quality value and with this data one can compute the enthalpy value of saturated steam. With the support of this small instrument and using the ASME correlations to determine the equilibrium temperature and knowing the discharge pressure in the inlet of surface condenser, the absolute enthalpy of the steam discharge can be computed with high precision and used to determine the power and thermodynamic efficiency of the turbine. (author)

  18. High-speed measurement of firearm primer blast waves

    OpenAIRE

    Courtney, Michael; Daviscourt, Joshua; Eng, Jonathan; Courtney, Amy

    2012-01-01

    This article describes a method and results for direct high-speed measurements of firearm primer blast waves employing a high-speed pressure transducer located at the muzzle to record the blast pressure wave produced by primer ignition. Key findings are: 1) Most of the lead styphnate based primer models tested show 5.2-11.3% standard deviation in the magnitudes of their peak pressure. 2) In contrast, lead-free diazodinitrophenol (DDNP) based primers had standard deviations of the peak blast p...

  19. Proceedings of the seventh annual symposium on explosives and blasting research

    International Nuclear Information System (INIS)

    Anon.

    1991-01-01

    Papers from this symposium dealt with the following topics: advanced primer designs, seismic effects of blasting, systems for velocity of detonation measurement and pressure measurement, toxic fumes from explosions, blast performance, blasting for rock fragmentation, computer-aided blast design, characteristics of liquid oxygen explosives, and correlations of performance of explosives with ground vibration, partitioning of energy, and firing time scatter effects. Papers have been indexed separately for inclusion on the data base

  20. Steam generator life management

    International Nuclear Information System (INIS)

    King, P.; McGillivray, R.; Reinhardt, W.; Millman, J.; King, B.; Schneider, W.

    2003-01-01

    'Full-Text:' Steam Generator Life Management responsibility embodies doing whatever is necessary to maintain the steam generation equipment of a nuclear plant in effective, reliable service. All comes together in that most critical deliverable, namely the submission of the documentation which wins approval for return to service after an outage program. Life management must address all aspects of SG reliability over the life of the plant. Nevertheless, the life management activities leading up to return to service approval is where all of it converges. Steam Generator Life Management activities entail four types of work, all equally important in supporting the objective of successful operation. These activities are i) engineering functions; including identification of inspection and maintenance requirements, outage planning and scope definition plus engineering assessment, design and analysis as necessary to support equipment operation, ii) fitness of service work; including the expert evaluation of degradation mechanisms, disposition of defects for return to service or not, and the fitness for service analysis as required to justify ongoing operation with acceptable defects, iii) inspection work; including large scale eddy current inspection of tubing, the definition of defect size and character, code inspections of pressure vessel integrity and visual inspections for integrity and iv) maintenance work; including repairs, retrofits, cleaning and modifications, all as necessary to implement the measures defined during activities i) through iii). The paper discusses the approach and execution of the program for the achievement of the above objectives and particularly of items i) and ii). (author)

  1. Optimal design of marine steam turbine

    International Nuclear Information System (INIS)

    Liu Chengyang; Yan Changqi; Wang Jianjun

    2012-01-01

    The marine steam turbine is one of the key equipment in marine power plant, and it tends to using high power steam turbine, which makes the steam turbine to be heavier and larger, it causes difficulties to the design and arrangement of the steam turbine, and the marine maneuverability is seriously influenced. Therefore, it is necessary to apply optimization techniques to the design of the steam turbine in order to achieve the minimum weight or volume by means of finding the optimum combination of design parameters. The math model of the marine steam turbine design calculation was established. The sensitivities of condenser pressure, power ratio of HP turbine with LP turbine, and the ratio of diameter with height at the end stage of LP turbine, which influence the weight of the marine steam turbine, were analyzed. The optimal design of the marine steam turbine, aiming at the weight minimization while satisfying the structure and performance constraints, was carried out with the hybrid particle swarm optimization algorithm. The results show that, steam turbine weight is reduced by 3.13% with the optimization scheme. Finally, the optimization results were analyzed, and the steam turbine optimization design direction was indicated. (authors)

  2. 饱和蒸汽压式波纹管疏水阀热动元件实验研究%Experimental study on the saturated vapor pressure type thermostatic bellows for steam traps

    Institute of Scientific and Technical Information of China (English)

    李树勋; 徐登伟; 把桥环

    2011-01-01

    针对液体膨胀式波纹管蒸汽疏水阀排量不稳定、漏汽率高等问题,分析波纹管热动元件的热工特性.基于Riedel蒸汽压方程和气液平衡方程,建立饱和蒸汽压式波纹管热动元件的热力学模型,设计相应实验系统,进行不同参数下的实验研究.结果表明,饱和蒸汽压式波纹管热动元件伸长量是相变温度的单值函数,近似呈指数关系;采用不同混合比、刚度及填充方式,可调节疏水阀的排水过冷度.%In view of the instabilities of displacement and high steam leakage rate for the liquid-expansion type ther-mostatic bellows steam traps, thermodynamic characteristical of thermostatic bellows was analyzed. Based on the Riedel equation and the vapor-liquid equilibrium equation, thermodynamic model of the saturated vapor pressure type thermostatic bellows was set up, corresponding experimental system was designed, and experimental studies with different parameters was carried out. The experimental results agree well with the theoretical analysis. The results show that the elongation A/I of the saturated vapor pressure type thermostatic bellows is monodrome function of phase transition temperature T, and relationship between the elongation A/I and the phase change temperature t is an exponential function. The subcooled temperature of steam trap can be adjusted by using different mixture ratio, different bellows' stiffness and different sufficient attire method. This paper establishes theoretical and experimental foundation for improving the performance of thermostatic bellows steam traps.

  3. 7 CFR 305.23 - Steam sterilization treatment schedules.

    Science.gov (United States)

    2010-01-01

    ... 7 Agriculture 5 2010-01-01 2010-01-01 false Steam sterilization treatment schedules. 305.23... Steam sterilization treatment schedules. Treatment schedule Temperature( °F) Pressure Exposure period (minutes) Directions T303-b-1 10 lbs 20 Use 28″ vacuum. Steam sterilization is not practical for treatment...

  4. Model for transient simulation in a PWR steam circuit

    International Nuclear Information System (INIS)

    Mello, L.A. de.

    1982-11-01

    A computer code (SURF) was developed and used to simulate pressure losses along the tubes of the main steam circuit of a PWR nuclear power plant, and the steam flow through relief and safety valves when pressure reactors its thresholds values. A thermodynamic model of turbines (high and low pressure), and its associated components are simulated too. The SURF computer code was coupled to the GEVAP computer code, complementing the simulation of a PWR nuclear power plant main steam circuit. (Author) [pt

  5. French steam generator design developments

    International Nuclear Information System (INIS)

    Ginier, R.; Campan, J.L.; Pontier, M.; Leridon, A.; Remond, A.; Castello, G.; Holcblat, A.; Paurobally, H.

    1986-01-01

    From the outset of the French nuclear power program, a significant R and D effort has been invested in improvement of the design and operation of Pressurized Water Reactors including a special committment to improving steam generators. The steam generator enhancement program has spawned a wide variety of specific R and D resources, e.g., low temperature hydraulic models for investigation of areas with single-phase flow, and freon-filled models for simulation of areas of steam generators experiencing two-phase flow (tube bundles and moisture separators). For the moisture separators, a large scale research program using freon-filled models and highly sophisticated instrumentation was used. Tests at reactor sites during startup of both 900 MWe and 1300 MWe have been used to validate the assumptions made on the basis of loop tests. These tests also demonstrated the validity of using freon to simulate two-phase flow conditions. The wealth of knowledge accumulated by the steam generator R and D program has been used to develop a new design of steam generators for the N4 plants. The current R and D effort is aimed at qualifying the N4 steam generator model and developing more comprehensive models. One prong of the R and D effort is the Megeve program. Megeve is a 25 MW steam generator which simulates operating conditions of the N4 model. The other prong is Clotaire, a freon-filled steam generator model which will be used to qualify thermal/hydraulic design codes used for multidimensional calculations for design of tube bundles

  6. Cleaning device for steam units in a nuclear power plant

    International Nuclear Information System (INIS)

    Sasamuro, Takemi.

    1978-01-01

    Purpose: To prevent radioactive contamination upon dismantling and inspection of steam units such as a turbine to a building containing such units and the peripheral area. Constitution: A steam generator indirectly heated by steam supplied from steam generating source in a separate system containing no radioactivity is provided to produce cleaning steam. A cleaning steam pipe is connected by way of a stop valve between separation valve of a nuclear power plant steam pipe and a high pressure turbine. Upon cleaning, the separation valve is closed, and steam supplied from the cleaning steam pipe is flown into a condenser. The water thus condensated is returned by way of a feed water heater and a condenser to a water storage tank. (Nakamura, S.)

  7. Severity parameters for steam cracking

    NARCIS (Netherlands)

    Golombok, M.; Bijl, J.L.M.; Kornegoor, M.

    2001-01-01

    There are several ways to measure severity in steam cracking which are all a function of residence time, temperature, and pressure. Many measures of severity are not practicable for experimental purposes. Our experimental study shows that methane make is the best measure of severity because it is an

  8. Analysis of experimental characteristics of multistage steam-jet electors of steam turbines

    Science.gov (United States)

    Aronson, K. E.; Ryabchikov, A. Yu.; Brodov, Yu. M.; Brezgin, D. V.; Zhelonkin, N. V.; Murmanskii, I. B.

    2017-02-01

    A series of questions for specification of physical gas dynamics model in flow range of steam-jet unit and ejector computation methodology, as well as functioning peculiarities of intercoolers, was formulated based on analysis of experimental characteristics of multistage team-jet steam turbines. It was established that coefficient defining position of critical cross-section of injected flow depends on characteristics of the "sound tube" zone. Speed of injected flow within this tube may exceed that of sound, and pressure jumps in work-steam decrease at the same time. Characteristics of the "sound tube" define optimal axial sizes of the ejector. According to measurement results, the part of steam condensing in the first-stage coolant constitutes 70-80% of steam amount supplied into coolant and is almost independent of air content in steam. Coolant efficiency depends on steam pressure defined by operation of steam-jet unit of ejector of the next stage after coolant of steam-jet stage, temperature, and condensing water flow. As a rule, steam entering content of steam-air mixture supplied to coolant is overheated with respect to saturation temperature of steam in the mixture. This should be taken into account during coolant computation. Long-term operation causes changes in roughness of walls of the ejector's mixing chamber. The influence of change of wall roughness on ejector characteristic is similar to the influence of reverse pressure of the steam-jet stage. Until some roughness value, injection coefficient of the ejector stage operating in superlimiting regime hardly changed. After reaching critical roughness, the ejector switches to prelimiting operating regime.

  9. Analysis of ways to control the supply of the blast, and their impact on gas-dynamic processes in the blast furnace

    Directory of Open Access Journals (Sweden)

    Віктор Петрович Кравченко

    2016-07-01

    Full Text Available The article presents the analysis of two methods of control over hot blast supply into a blast furnace with constant pressure and constant amount (consumption. The analysis of these two methods was performed with the aim of determining their influence upon changes in gas pressure in the blast furnace top. The blast furnace was considered as a unity of vessels (furnace hearth, the top and gas-dynamic resistance (a column of charge materials. A differential equation was obtained, with regard to the dynamic balance of gas flow at the inlet and outlet of the top; the equation relates the pressure and gas consumption at the top to the pressure and hot blast consumption at the inlet and outlet of the furnace and to the resistance of the column of charge materials. The column of charge materials is considered as n-th number of channels through which gas flow inside the furnace moves and which resist to the flow. By the analysis of this equation at steady state (automatic stabilization of gas pressure in the top, the conditions were obtained to be satisfied with the specified value of gas pressure in the top. This value is equal to a half of the sum of the value of hot blast pressure at the inlet into the furnace and the value of pressure inside the collector of blast furnace gas. This conclusion is verified by the operation practice of blast furnaces in Ukraine. While analyzing the second method of controlling the supply of blast supply-stabilization of consumption (amount of hot blast supplied into the furnace it has been shown that the method could be realized in condition of stabilization of the amount of blast furnace gas, going out of the furnace. As the resistance of the column of charge materials constantly changes it is necessary to change the hot blast pressure in order to ensure the constant amount of blast, supplied into the furnace. It is often connected with possible substantial pressure fluctuations of hot blast at the inlet of the

  10. A drier unit for steam separators

    International Nuclear Information System (INIS)

    Peyrelongue, J.-P.

    1973-01-01

    Description is given of a drier unit adapted to equip a water separator mounted in a unit for treating a wet steam fed from a high pressure enclosure, so as to dry and contingently superheat said steam prior to injecting same into a turbine low pressure stage. This drier unit is constituted by at least a stack of separating sheets maintained in parallel relationship and at a slight angle with respect to the horizontal so as to allow the water provided by wet steam to flow toward a channel communicating with a manifold, and by means for guiding the steam between the sheets and evenly distributing it. This can be applied to steam turbines in nuclear power stations [fr

  11. Testing installation for a steam generator

    International Nuclear Information System (INIS)

    Dubourg, M.

    1985-01-01

    The invention proposes a testing installation for a steam generator associated to a boiler, comprising a testing exchanger connected to a feeding circuit in secondary fluid and to a circuit to release the steam produced, and comprising a heating-tube bundle connected to a closed circuit of circulation of a primary coolant at the same temperature and at the pressure than the primary fluid. The heating-tube bundle of the testing exchanger has the same height than the primary bundle of the steam generator and the testing exchanger is at the same level and near the steam generator and is fed by the same secondary fluid such as it is subject to the same operation phases during a long period. The in - vention applies, more particularly, to the steam generators of pressurized water nuclear power plants [fr

  12. Simulation of main steam and feedwater system of full scope simulator for Qinshan 300 MW Nuclear Power Unit

    International Nuclear Information System (INIS)

    Zhao Xiaoyu

    1996-01-01

    The simulation of main steam and feedwater system is the most important and maximal part in secondary circuit model, including all of main steam and feedwater's thermal-hydraulic properties, except heat-exchange of secondary side of steam generator. It simulates main steam header, steam power in each stage of turbine, moisture separator-reheater, deaerator, condenser, high pressure and low pressure heater, auxiliary feedwater and main steam bypass in full scope

  13. Experimental investigation of blast mitigation and particle-blast interaction during the explosive dispersal of particles and liquids

    Science.gov (United States)

    Pontalier, Q.; Loiseau, J.; Goroshin, S.; Frost, D. L.

    2018-05-01

    The attenuation of a blast wave from a high-explosive charge surrounded by a layer of inert material is investigated experimentally in a spherical geometry for a wide range of materials. The blast wave pressure is inferred from extracting the blast wave velocity with high-speed video as well as direct measurements with pressure transducers. The mitigant consists of either a packed bed of particles, a particle bed saturated with water, or a homogeneous liquid. The reduction in peak blast wave overpressure is primarily dependent on the mitigant to explosive mass ratio, M/ C, with the mitigant material properties playing a secondary role. Relative peak pressure mitigation reduces with distance and for low values of M/ C (compaction, deformation, and fracture of the powders plays an important role. The difference in scaled arrival time of the blast and material fronts increases with M/ C and scaled distance, with solid particles giving the largest separation between the blast wave and cloud of particles. Surrounding a high-explosive charge with a layer of particles reduces the positive-phase blast impulse, whereas a liquid layer has no influence on the impulse in the far field. Taking the total impulse due to the blast wave and material impact into account implies that the damage to a nearby structure may actually be augmented for a range of distances. These results should be taken into consideration in the design of explosive mitigant systems.

  14. Thorough Chemical Decontamination with the MEDOC Process : Batch Treatment of Dismantled Pieces or Loop Treatment of Large Components Such as the BR3 Steam Generator and Pressurizer

    International Nuclear Information System (INIS)

    Ponnet, M.; Klein, M.; Massaut, V.; Davain, H.; Aleton, G.

    2003-01-01

    The dismantling of the BR3-PWR reactor leads to the production of large masses of contaminated metallic pieces, including structural materials, primary pipings, tanks and heat exchangers. One of our main objectives is to demonstrate that we can minimize the volume of radioactive waste in an economical way, by the use of alternative waste routes, such as the clearance of materials after thorough decontamination. The SCKoCEN uses its own developed chemical decontamination process, so-called MEDOC (Metal Decontamination by Oxidation with Cerium), based on the use of cerium IV as strong oxidant in sulphuric acid with continuous regeneration using ozone. An industrial installation has been designed and constructed in close collaboration with Framatome-ANP (France). This installation started operation in September 1999 for the treatment of the metallic pieces arising from the dismantling of the BR3 reactor. Since then, more than 25 tons of contaminated material including primary pipes have been treated batchwise with success. 75 % of material could be directly cleared after treatment (Activity lower than 0.1 Bq/g for 60Co) and the other 25% free released after melting activity. The SCKoCEN performed in April 2002 the closed loop decontamination of the BR3 Steam Generator by connection of the MEDOC plant after few adaptations. The decontamination was done within 30 cycles in 3 weeks with consecutive steps like decontamination steps (injection of the solution into the SG) and regeneration steps with ozone. In total, 60 hours of decontamination at 70 C and 130 hours of regeneration were needed to reach the objectives. The tube bundle (600 m2) was attacked and about 10 (micro)m representing more than 41 kg of stainless steel and 2.06 GBq of 60Co was dissolved into the solution. The residual contamination measurements made directly into the water box are still going on, however it seems that the objective to reach the free release criteria after melting is achieved. The next

  15. Evaluation of high pressure water blast with rotating spray bar for removing paint and rubber deposits from airport runways, and review of runway slipperiness problems created by rubber contamination

    Science.gov (United States)

    Horne, W. B.; Griswold, G. D.

    1975-01-01

    A high pressure water blast with rotating spray bar treatment for removing paint and rubber deposits from airport runways is studied. The results of the evaluation suggest that the treatment is very effective in removing above surface paint and rubber deposits to the point that pavement skid resistance is restored to trafficked but uncontaminated runway surface skid resistance levels. Aircraft operating problems created by runway slipperiness are reviewed along with an assessment of the contributions that pavement surface treatments, surface weathering, traffic polishing, and rubber deposits make in creating or alleviating runway slipperiness. The results suggest that conventional surface treatments for both portland cement and asphaltic concrete runways are extremely vulnerable to rubber deposit accretions which can produce runway slipperiness conditions for aircraft operations as or more slippery than many snow and ice-covered runway conditions. Pavement grooving surface treatments are shown to be the least vulnerable to rubber deposits accretion and traffic polishing of the surface treatments examined.

  16. A phase I study of low-pressure hyperbaric oxygen therapy for blast-induced post-concussion syndrome and post-traumatic stress disorder.

    Science.gov (United States)

    Harch, Paul G; Andrews, Susan R; Fogarty, Edward F; Amen, Daniel; Pezzullo, John C; Lucarini, Juliette; Aubrey, Claire; Taylor, Derek V; Staab, Paul K; Van Meter, Keith W

    2012-01-01

    This is a preliminary report on the safety and efficacy of 1.5 ATA hyperbaric oxygen therapy (HBOT) in military subjects with chronic blast-induced mild to moderate traumatic brain injury (TBI)/post-concussion syndrome (PCS) and post-traumatic stress disorder (PTSD). Sixteen military subjects received 40 1.5 ATA/60 min HBOT sessions in 30 days. Symptoms, physical and neurological exams, SPECT brain imaging, and neuropsychological and psychological testing were completed before and within 1 week after treatment. Subjects experienced reversible middle ear barotrauma (5), transient deterioration in symptoms (4), and reversible bronchospasm (1); one subject withdrew. Post-treatment testing demonstrated significant improvement in: symptoms, neurological exam, full-scale IQ (+14.8 points; pSPECT coefficient of variation in all white matter and some gray matter ROIs after the first HBOT, and in half of white matter ROIs after 40 HBOT sessions, and SPECT statistical parametric mapping analysis (diffuse improvements in regional cerebral blood flow after 1 and 40 HBOT sessions). Forty 1.5 ATA HBOT sessions in 1 month was safe in a military cohort with chronic blast-induced PCS and PTSD. Significant improvements occurred in symptoms, abnormal physical exam findings, cognitive testing, and quality-of-life measurements, with concomitant significant improvements in SPECT.

  17. Numerical simulation of effect of catalyst wire-mesh pressure drop characteristics on flow distribution in catalytic parallel plate steam reformer

    DEFF Research Database (Denmark)

    Sigurdsson, Haftor Örn; Kær, Søren Knudsen

    2012-01-01

    Steam reforming of hydrocarbons using a catalytic plate-type-heat-exchanger (CPHE) reformer is an attractive method of producing hydrogen for a fuel cell-based micro combined-heat-and-power system. In this study the flow distribution in a CPHE reformer, which uses a coated wire-mesh catalyst...

  18. Blast densification trials for oilsands tailings

    Energy Technology Data Exchange (ETDEWEB)

    Port, A. [Klohn Crippen Berger Ltd., Vancouver, BC (Canada); Martens, S. [Klohn Crippen Berger Ltd., Calgary, AB (Canada); Eaton, T. [Shell Canada Ltd., Calgary, AB (Canada)

    2010-07-01

    The Shell Canada Muskeg River Mine External Tailings Facility (ETF) is an upstream constructed tailings facility located near Fort McMurray, Alberta. Raises have incrementally stepped out over the beach since construction of the starter dam and deposition within standing water has left some parts of the beach in a loose state. In order to assess the effectiveness of blast densification, a blast densification trial program that was conducted in 2006 at the ETF. The primary purpose of the test program was to determine the effectiveness of blast densification in tailings containing layers and zones of bitumen. The paper described the site characterization and explosive compaction trial program, with particular reference to test layout; drilling methodology; and blasting and timing sequence. The paper also described the instrumentation, including the seismographs; high pressure electric piezometers; low pressure electric piezometers; vibrating wire piezometers; inclinometers; settlement gauges; and surveys. Trial observations and post-trial observations were also presented. It was concluded that controlled blasting techniques could be used to safely induce liquefaction in localized areas within the tailings deposit, with a resulting increase in the tailings density. 5 refs., 1 tab., 14 figs.

  19. Modelling human eye under blast loading.

    Science.gov (United States)

    Esposito, L; Clemente, C; Bonora, N; Rossi, T

    2015-01-01

    Primary blast injury (PBI) is the general term that refers to injuries resulting from the mere interaction of a blast wave with the body. Although few instances of primary ocular blast injury, without a concomitant secondary blast injury from debris, are documented, some experimental studies demonstrate its occurrence. In order to investigate PBI to the eye, a finite element model of the human eye using simple constitutive models was developed. The material parameters were calibrated by a multi-objective optimisation performed on available eye impact test data. The behaviour of the human eye and the dynamics of mechanisms occurring under PBI loading conditions were modelled. For the generation of the blast waves, different combinations of explosive (trinitrotoluene) mass charge and distance from the eye were analysed. An interpretation of the resulting pressure, based on the propagation and reflection of the waves inside the eye bulb and orbit, is proposed. The peculiar geometry of the bony orbit (similar to a frustum cone) can induce a resonance cavity effect and generate a pressure standing wave potentially hurtful for eye tissues.

  20. Spalling of concrete walls under blast load

    International Nuclear Information System (INIS)

    Kot, C.A.

    1977-01-01

    A common effect of the detonation of explosives in close proximity of concrete shield walls is the spalling (scabbing) of the back face of the wall. Spalling is caused by the free surface reflection of the shock wave induced in the wall by high pressure air blast and occurs whenever the dynamic tensile rupture strength is exceeded. While a complex process, reasonable analytical spall estimates can be obtained for brittle materials with low tensile strengths, such as concrete, by assuming elastic material behavior and instantaneous spall formation. Specifically, the spall thicknesses and velocities for both normal and oblique incidence of the shock wave on the back face of the wall are calculated. The complex exponential decay wave forms of the air blast are locally approximated by simple power law expressions. Variations of blast wave strength with distance to the wall, charge weight and angle of incidence are taken into consideration. The shock wave decay in the wall is also accounted for by assuming elastic wave propagation. For explosions close-in to the wall, where the reflected blast wave pressures are sufficiently high, multiple spall layers are formed. Successive spall layers are of increasing thickness, at the same time the spall velocities decrease. The spall predictions based on elastic theory are in overall agreement with experimntal results and provide a rapid means of estimating spalling trends of concrete walls subjected to air blast. (Auth.)

  1. Linking blast physics to biological outcomes in mild traumatic brain injury: Narrative review and preliminary report of an open-field blast model.

    Science.gov (United States)

    Song, Hailong; Cui, Jiankun; Simonyi, Agnes; Johnson, Catherine E; Hubler, Graham K; DePalma, Ralph G; Gu, Zezong

    2018-03-15

    Blast exposures are associated with traumatic brain injury (TBI) and blast-induced TBIs are common injuries affecting military personnel. Department of Defense and Veterans Administration (DoD/VA) reports for TBI indicated that the vast majority (82.3%) has been mild TBI (mTBI)/concussion. mTBI and associated posttraumatic stress disorders (PTSD) have been called "the invisible injury" of the current conflicts in Iraq and Afghanistan. These injuries induce varying degrees of neuropathological alterations and, in some cases, chronic cognitive, behavioral and neurological disorders. Appropriate animal models of blast-induced TBI will not only assist the understanding of physical characteristics of the blast, but also help to address the potential mechanisms. This report provides a brief overview of physical principles of blast, injury mechanisms related to blast exposure, current blast animal models, and the neurological behavioral and neuropathological findings related to blast injury in experimental settings. We describe relationships between blast peak pressures and the observed injuries. We also report preliminary use of a highly reproducible and intensity-graded blast murine model carried out in open-field with explosives, and describe physical and pathological findings in this experimental model. Our results indicate close relationships between blast intensities and neuropathology and behavioral deficits, particularly at low level blast intensities relevant to mTBI. Copyright © 2016 Elsevier B.V. All rights reserved.

  2. Steam-water relative permeability

    Energy Technology Data Exchange (ETDEWEB)

    Ambusso, W.; Satik, C.; Home, R.N. [Stanford Univ., CA (United States)

    1997-12-31

    A set of relative permeability relations for simultaneous flow of steam and water in porous media have been measured in steady state experiments conducted under the conditions that eliminate most errors associated with saturation and pressure measurements. These relations show that the relative permeabilities for steam-water flow in porous media vary approximately linearly with saturation. This departure from the nitrogen/water behavior indicates that there are fundamental differences between steam/water and nitrogen/water flows. The saturations in these experiments were measured by using a high resolution X-ray computer tomography (CT) scanner. In addition the pressure gradients were obtained from the measurements of liquid phase pressure over the portions with flat saturation profiles. These two aspects constitute a major improvement in the experimental method compared to those used in the past. Comparison of the saturation profiles measured by the X-ray CT scanner during the experiments shows a good agreement with those predicted by numerical simulations. To obtain results that are applicable to general flow of steam and water in porous media similar experiments will be conducted at higher temperature and with porous rocks of different wetting characteristics and porosity distribution.

  3. Scale testing of a partially confined blast chamber

    CSIR Research Space (South Africa)

    Grundling, W

    2012-10-01

    Full Text Available in pressure readings. A scale model of the blast chamber, Emily, was constructed with the addition of a pendulum plate hanging concentrically covering 65% of the open area. PURPOSE OF SCALED BLAST CHAMBER The purpose of this particular test is to evaluate... PHASE Illustrated in Figure 3 and 4 are the results obtained during testing of the scaled blast chamber. In both cases the pressure dissipates over time, showing pulsating behaviour as the shockwaves reflect off the chamber walls. By looking...

  4. Water regime of steam power plants

    International Nuclear Information System (INIS)

    Oesz, Janos

    2011-01-01

    The water regime of water-steam thermal power plants (secondary side of pressurized water reactors (PWR); fossil-fired thermal power plants - referred to as steam power plants) has changed in the past 30 years, due to a shift from water chemistry to water regime approach. The article summarizes measures (that have been realised by chemists of NPP Paks) on which the secondary side of NPP Paks has become a high purity water-steam power plant and by which the water chemistry stress corrosion risk of heat transfer tubes in the VVER-440 steam generators was minimized. The measures can also be applied to the water regime of fossil-fired thermal power plants with super- and subcritical steam pressure. Based on the reliability analogue of PWR steam generators, water regime can be defined as the harmony of construction, material(s) and water chemistry, which needs to be provided in not only the steam generators (boiler) but in each heat exchanger of steam power plant: - Construction determines the processes of flow, heat and mass transfer and their local inequalities; - Material(s) determines the minimal rate of general corrosion and the sensitivity for local corrosion damage; - Water chemistry influences the general corrosion of material(s) and the corrosion products transport, as well as the formation of local corrosion environment. (orig.)

  5. Steam Digest 2002

    Energy Technology Data Exchange (ETDEWEB)

    2003-11-01

    Steam Digest 2002 is a collection of articles published in the last year on steam system efficiency. DOE directly or indirectly facilitated the publication of the articles through it's BestPractices Steam effort. Steam Digest 2002 provides a variety of operational, design, marketing, and program and program assessment observations. Plant managers, engineers, and other plant operations personnel can refer to the information to improve industrial steam system management, efficiency, and performance.

  6. Future development of large steam turbines

    International Nuclear Information System (INIS)

    Chevance, A.

    1975-01-01

    An attempt is made to forecast the future of the large steam turbines till 1985. Three parameters affect the development of large turbines: 1) unit output; and a 2000 to 2500MW output may be scheduled; 2) steam quality: and two steam qualities may be considered: medium pressure saturated or slightly overheated steam (light water, heavy water); light enthalpie drop, high pressure steam, high temperature; high enthalpic drop; and 3) the quality of cooling supply. The largest range to be considered might be: open system cooling for sea-sites; humid tower cooling and dry tower cooling. Bi-fluid cooling cycles should be also mentioned. From the study of these influencing factors, it appears that the constructor, for an output of about 2500MW should have at his disposal the followings: two construction technologies for inlet parts and for high and intermediate pressure parts corresponding to both steam qualities; exhaust sections suitable for the different qualities of cooling supply. The two construction technologies with the two steam qualities already exist and involve no major developments. But, the exhaust section sets the question of rotational speed [fr

  7. Proceedings of the fourteenth annual symposium on explosives and blasting research

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-12-31

    Subjects covered include: ground vibration effects on structures; open-pit blast vibration prediction; effects of velocity of detonation and gas pressurization on fragmentation in layered rock; thermal ignition for emulsion powder explosives and emulsion matrix; effect of cut-off pressure on energy partition and blast design; new burden and spacing formulae for optimum blasting; calculated risk of experiencing lightning caused unplanned detonation; predicting explosive toxic fumes; and stemming techniques for loading angled holes charged with Anfo.

  8. Circumferential cracking of steam generator tubes

    International Nuclear Information System (INIS)

    Karwoski, K.J.

    1997-04-01

    On April 28, 1995, the U.S. Nuclear Regulatory Commission (NRC) issued Generic Letter (GL) 95-03, open-quote Circumferential Cracking of Steam Generator Tubes.close-quote GL 95-03 was issued to obtain information needed to verify licensee compliance with existing regulatory requirements regarding the integrity of steam generator tubes in domestic pressurized-water reactors (PWRs). This report briefly describes the design and function of domestic steam generators and summarizes the staff's assessment of the responses to GL 95-03. The report concludes with several observations related to steam generator operating experience. This report is intended to be representative of significant operating experience pertaining to circumferential cracking of steam generator tubes from April 1995 through December 1996. Operating experience prior to April 1995 is discussed throughout the report, as necessary, for completeness

  9. Considerations in selecting tubing materials for CANDU steam generators

    International Nuclear Information System (INIS)

    Hemmings, R.L.

    1978-01-01

    Corrosion resistance is the major consideration in selecting tubing material for CANDU steam generators. Corrosion, and additional considerations, lead to the following steam generator tubing material recommendations: for CANDU-BPHWR's (boiling pressurized heavy water reactors) low-cobalt Incoloy-800; for CANDU-PHWR's (pressurized, non-boiling, heavy water reactors), low-cobalt Monel-400

  10. Distinguishing Realistic Military Blasts from Firecrackers in Mitigation Studies of Blast Induced Traumatic Brain Injury

    Energy Technology Data Exchange (ETDEWEB)

    Moss, W C; King, M J; Blackman, E G

    2011-01-21

    In their Contributed Article, Nyein et al. (1,2) present numerical simulations of blast waves interacting with a helmeted head and conclude that a face shield may significantly mitigate blast induced traumatic brain injury (TBI). A face shield may indeed be important for future military helmets, but the authors derive their conclusions from a much smaller explosion than typically experienced on the battlefield. The blast from the 3.16 gm TNT charge of (1) has the following approximate peak overpressures, positive phase durations, and incident impulses (3): 10 atm, 0.25 ms, and 3.9 psi-ms at the front of the head (14 cm from charge), and 1.4 atm, 0.32 ms, and 1.7 psi-ms at the back of a typical 20 cm head (34 cm from charge). The peak pressure of the wave decreases by a factor of 7 as it traverses the head. The blast conditions are at the threshold for injury at the front of the head, but well below threshold at the back of the head (4). The blast traverses the head in 0.3 ms, roughly equal to the positive phase duration of the blast. Therefore, when the blast reaches the back of the head, near ambient conditions exist at the front. Because the headform is so close to the charge, it experiences a wave with significant curvature. By contrast, a realistic blast from a 2.2 kg TNT charge ({approx} an uncased 105 mm artillery round) is fatal at an overpressure of 10 atm (4). For an injury level (4) similar to (1), a 2.2 kg charge has the following approximate peak overpressures, positive phase durations, and incident impulses (3): 2.1 atm, 2.3 ms, and 18 psi-ms at the front of the head (250 cm from charge), and 1.8 atm, 2.5 ms, and 16.8 psi-ms at the back of the head (270 cm from charge). The peak pressure decreases by only a factor of 1.2 as it traverses the head. Because the 0.36 ms traversal time is much smaller than the positive phase duration, pressures on the head become relatively uniform when the blast reaches the back of the head. The larger standoff implies

  11. Experimental investigation of blast mitigation and particle-blast interaction during the explosive dispersal of particles and liquids

    Science.gov (United States)

    Pontalier, Q.; Loiseau, J.; Goroshin, S.; Frost, D. L.

    2018-04-01

    The attenuation of a blast wave from a high-explosive charge surrounded by a layer of inert material is investigated experimentally in a spherical geometry for a wide range of materials. The blast wave pressure is inferred from extracting the blast wave velocity with high-speed video as well as direct measurements with pressure transducers. The mitigant consists of either a packed bed of particles, a particle bed saturated with water, or a homogeneous liquid. The reduction in peak blast wave overpressure is primarily dependent on the mitigant to explosive mass ratio, M/C, with the mitigant material properties playing a secondary role. Relative peak pressure mitigation reduces with distance and for low values of M/C (pressure levels in the mid-to-far field. Solid particles are more effective at mitigating the blast overpressure than liquids, particularly in the near field and at low values of M/C, suggesting that the energy dissipation during compaction, deformation, and fracture of the powders plays an important role. The difference in scaled arrival time of the blast and material fronts increases with M/C and scaled distance, with solid particles giving the largest separation between the blast wave and cloud of particles. Surrounding a high-explosive charge with a layer of particles reduces the positive-phase blast impulse, whereas a liquid layer has no influence on the impulse in the far field. Taking the total impulse due to the blast wave and material impact into account implies that the damage to a nearby structure may actually be augmented for a range of distances. These results should be taken into consideration in the design of explosive mitigant systems.

  12. Modelling and simulation of the steam line, the high and low pressure turbines and the pressure regulator for the SUN-RAH nucleo electric university simulator; Modelado y simulacion de la linea de vapor, las turbinas de alta y de baja presion y el regulador de presion para el simulador universitario de nucleo electricas SUN RAH

    Energy Technology Data Exchange (ETDEWEB)

    Lopez R, A. [DEPFI, Campus Morelos, en IMTA Jiutepec, Morelos, UNAM (Mexico)]. e-mail: andyskamx@yahoo.com.mx

    2003-07-01

    In the following article the development of a simulator that allows to represent the dynamics of the following systems: steam line, nozzle, vapor separator, reheater, high pressure turbine, low pressure turbine, power generator and the pressure regulator of a nucleo electric power station. We start from the supposition that this plant will be modeled from a nuclear reactor type BWR (Boiling Water Reactor), using models of reduced order that represent the more important dynamic variables of the physical processes that happen along the steam line until the one generator. To be able to carry out the simulation in real time the Mat lab mathematical modeling software is used, as well as the specific simulation tool Simulink. It is necessary to point out that the platform on which the one is executed the simulator is the Windows operating system, to allow the intuitive use that only this operating system offers. The above-mentioned obeys to that the objective of the simulator it is to help the user to understand some of the dynamic phenomena that are present in the systems of a nuclear plant, and to provide a tool of analysis and measurement of variables to predict the desirable behavior of the same ones. The model of a pressure controller for the steam lines, the high pressure turbine and the low pressure turbine is also presented that it will be the one in charge of regulating the demand of the system according to the characteristics and critic restrictions of safety and control, assigned according to those wanted parameters of performance of this system inside the nucleo electric plant. This simulator is totally well defined and it is part of the University student nucleo electric simulator with Boiling Water Reactor (SUN-RAH), an integral project and of greater capacity. (Author)

  13. Method of determining the enthalpy and moisture content of wet steam

    International Nuclear Information System (INIS)

    Silvestri, G.J. Jr.

    1991-01-01

    This patent describes a nuclear powered multi-stage steam turbine system wherein steam at higher than atmospheric pressure is introduced into the turbine system at a high pressure turbine element and thereafter flows through a series of turbine elements at successively decreasing pressures, wherein portions of the steam are extracted from the turbine elements at a plurality of lower pressure points and the steam is finally exhausted at a lowest pressure point, the method of determining moisture content and enthalpy of steam at a selected pressure point. It comprises sampling a small quantity of steam at the selected pressure point; super heating the steam sample to a single-phase state by reducing its pressure and bottling it in a closed measuring chamber whereby the flow energy of the sample is converted into internal energy; measuring the pressure of the steam sample within the chamber; determining the sonic velocity of the steam sample by passing a sound wave through the sample from a transmitter to a receiver located at a known distance from the transmitter and measuring the time required for the sound wave to travel from transmitter to receiver; and utilizing the measured pressure and sonic velocity of the steam sample to calculate the moisture content and enthalpy of the steam at the selected pressure point

  14. Validation of Hiriart equation to compute steam production by the lip pressure method; Validacion de la ecuacion de Hiriart para calculo de gasto de vapor por el metodo de presion de labio

    Energy Technology Data Exchange (ETDEWEB)

    Flores Armenta, Magaly [Gerencia de Proyectos Geotermoelectricos de la Comision Federal de Electricidad, Morelia (Mexico)

    1996-09-01

    Mainly in new geothermal wells, it is necessary to evaluate the production in a very fast, simple and not expensive way, to know the convenience to install surface equipment, such as silencers and separators, to drive the steam to the commercial gathering system. In practice, one of the most known methods is the lip pressure one, which requires a simple set of installations. The objective of this paper is to validate the steam flow rate calculated by the lip pressure method, with respect to the ASME method. The ASME method is known for its accuracy, and is done by measuring the steam and liquid after a high pressure separator, by an orifice plate of known diameter and a triangular weir. Results of the validation show up the feasibility of application of the lip pressure method by using a simple adjustment equation. Percentage of mistake results less than 1%, without any notable influence of the production enthalpy. That equation to be applied in a general case, is as follows: Q{nu} =(20642)(F*P*D{sup 2}/{radical}h-2000). For the particular case of the Los Azufres geothermal field, the equation is: Q{nu}= 810*P*D{sup 2} [Espanol] En los pozos geotermicos, principalmente en los nuevos, es necesario evaluar su produccion de manera rapida, sencilla y economica, para determinar la conveniencia de instalar equipo superficial, como separadores, silenciadores, etc., que permita la integracion del vapor al sistema comercial de generacion electrica. Para fines practicos uno de los metodos mas conocidos es el de presion de labio, que solo requiere un arreglo sencillo de instalaciones superficiales. En este documento se validan y ajustan los calculos de produccion de vapor por ese metodo de presion de labio, con respecto a las mediciones exactas efectuadas con el metodo ASME. Este ultimo es reconocido internacionalmente por su precision, y se lleva a cabo separando la mezcla obtenida en superficie en un recipiente a presion para medir el vapor a traves de una placa de orificio

  15. Steam generator with perfected dryers

    International Nuclear Information System (INIS)

    Fenet, J.C.

    1987-01-01

    This steam generator has vertically superposed array of steam dryers. These dryers return the steam flow of 180 0 . The return of the water is made by draining channels to the steam production zone [fr

  16. Steam separator-superheater with drawing of a fraction of the dried steam

    International Nuclear Information System (INIS)

    Bessouat, Roger; Marjollet, Jacques.

    1976-01-01

    This invention concerns a vertical separator-superheater of the steam from a high pressure expansion turbine before it is admitted to an expansion turbine at a lower pressure, by heat exchange with steam under a greater pressure, and drawing of a fraction of the dried steam before it is superheated. Such drawing off is necessary in the heat exchange systems of light water nuclear reactors. Its purpose is to provide a separator-superheater that provides an even flow of non superheated steam and a regular distribution of the steam to be superheated to the various superheating bundles, with a significantly uniform temperature of the casing, thereby preventing thermal stresses and ensuring a minimal pressure drop. The vertical separator-superheater of the invention is divided into several vertical sections comprising as from the central area, a separation area of the steam entrained water and a superheater area and at least one other vertical section with only a separation area of the steam entrained water [fr

  17. High speed drying of saturated steam

    International Nuclear Information System (INIS)

    Marty, C.; Peyrelongue, J.P.

    1993-01-01

    This paper describes the development of the drying process for the saturated steam used in the PWR nuclear plant turbines in order to prevent negative effects of water on turbine efficiency, maintenance costs and equipment lifetime. The high speed drying concept is based on rotating the incoming saturated steam in order to separate water which is more denser than the steam; the water film is then extracted through an annular slot. A multicellular modular equipment has been tested. Applications on high and low pressure extraction of various PWR plants are described (Bugey, Loviisa)

  18. HTGR steam generator development

    International Nuclear Information System (INIS)

    Schuetzenduebel, W.G.; Hunt, P.S.; Weber, M.

    1976-01-01

    More than 40 gas-cooled reactor plants have produced in excess of 400 reactor years of operating experience which have proved a reasonably high rate of gas-cooled reactor steam generator availability. The steam generators used in these reactors include single U-tube and straight-tube steam generators as well as meander type and helically wound or involute tube steam generators. It appears that modern reactors are being equipped with helically wound steam generators of the once-through type as the end product of steam generator evolution in gas-cooled reactor plants. This paper provides a general overview of gas-cooled reactor steam generator evolution and operating experience and shows how design criteria and constraints, research and development, and experience data are factored into the design/development of modern helically wound tube steam generators for the present generation of gas-cooled reactors

  19. Steam Digest 2001

    Energy Technology Data Exchange (ETDEWEB)

    2002-01-01

    Steam Digest 2001 chronicles BestPractices Program's contributions to the industrial trade press for 2001, and presents articles that cover technical, financial and managerial aspects of steam optimization.

  20. Steam generator tube extraction

    International Nuclear Information System (INIS)

    Delorme, H.

    1985-05-01

    To enable tube examination on steam generators in service, Framatome has now developed a process for removing sections of steam generator tubes. Tube sections can be removed without being damaged for treating the tube section expanded in the tube sheet

  1. Steam sterilization does not require saturated steam

    NARCIS (Netherlands)

    van Doornmalen Gomez Hoyos, J. P.C.M.; Paunovic, A.; Kopinga, K.

    2017-01-01

    The most commonly applied method to sterilize re-usable medical devices in hospitals is steam sterilization. The essential conditions for steam sterilization are derived from sterilization in water. Microbiological experiments in aqueous solutions have been used to calculate various time–temperature

  2. Steam explosions in light water reactors

    International Nuclear Information System (INIS)

    1981-01-01

    The report deals with a postulated accident caused by molten fuel falling into the lower plenum of the containment of a reactor. The analysis which is presented in the report shows that the thermal energy released in the resulting steam explosion is not enough to destroy the pressure vessel or the containment. The report was prepared for the Swedish Governmental Committee on steam explosion in light water reactors. It includes statements issued by internationally well-known specialists. (G.B.)

  3. Corrosion of PWR steam generators

    International Nuclear Information System (INIS)

    Garnsey, R.

    1979-01-01

    Some designs of pressurized water reactor (PWR) steam generators have experienced a variety of corrosion problems which include stress corrosion cracking, tube thinning, pitting, fatigue, erosion-corrosion and support plate corrosion resulting in 'denting'. Large international research programmes have been mounted to investigate the phenomena. The operational experience is reviewed and mechanisms which have been proposed to explain the corrosion damage are presented. The implications for design development and for boiler and feedwater control are discussed. (author)

  4. The Invisibility of Steam

    Science.gov (United States)

    Greenslade, Thomas B., Jr.

    2014-01-01

    Almost everyone "knows" that steam is visible. After all, one can see the cloud of white issuing from the spout of a boiling tea kettle. In reality, steam is the gaseous phase of water and is invisible. What you see is light scattered from the tiny droplets of water that are the result of the condensation of the steam as its temperature…

  5. Strategies for steam

    International Nuclear Information System (INIS)

    Hennagir, T.

    1996-01-01

    This article is a review of worldwide developments in the steam turbine and heat recovery steam generator markets. The Far East is driving the market in HRSGs, while China is driving the market in orders placed for steam turbine prime movers. The efforts of several major suppliers are discussed, with brief technical details being provided for several projects

  6. Steam Digest: Volume IV

    Energy Technology Data Exchange (ETDEWEB)

    2004-07-01

    This edition of the Steam Digest is a compendium of 2003 articles on the technical and financial benefits of steam efficiency, presented by the stakeholders of the U.S. Department of Energy's BestPractices Steam effort.

  7. Steam Digest Volume IV

    Energy Technology Data Exchange (ETDEWEB)

    None

    2004-07-01

    This edition of the Steam Digest is a compendium of 2003 articles on the technical and financial benefits of steam efficiency, presented by the stakeholders of the U.S. Department of Energy's BestPractices Steam effort.

  8. Large nuclear steam turbine plants

    International Nuclear Information System (INIS)

    Urushidani, Haruo; Moriya, Shin-ichi; Tsuji, Kunio; Fujita, Isao; Ebata, Sakae; Nagai, Yoji.

    1986-01-01

    The technical development of the large capacity steam turbines for ABWR plants was partially completed, and that in progress is expected to be completed soon. In this report, the outline of those new technologies is described. As the technologies for increasing the capacity and heightening the efficiency, 52 in long blades and moisture separating heaters are explained. Besides, in the large bore butterfly valves developed for making the layout compact, the effect of thermal efficiency rise due to the reduction of pressure loss can be expected. As the new technology on the system side, the simplification of the turbine system and the effect of heightening the thermal efficiency by high pressure and low pressure drain pumping-up method based on the recent improvement of feed water quality are discussed. As for nuclear steam turbines, the actual records of performance of 1100 MW class, the largest output at present, have been obtained, and as a next large capacity machine, the development of a steam turbine of 1300 MWe class for an ABWR plant is in progress. It can be expected that by the introduction of those new technologies, the plants having high economical efficiency are realized. (Kako, I.)

  9. Steam generator operation and maintenance

    International Nuclear Information System (INIS)

    Lee, C.K.

    2004-01-01

    capacity due to excessive pressure drop across the tube support plates. OTSG owners group has developed both mechanical and chemical cleaning process and an upgraded secondary water chemistry in resolving these problems. The OTSG performance has been greatly improved since OTSG plants implemented chemical cleaning and morpholine water chemistry. The SGOG project officially ended December 31, 1986. A six-year Steam Generator Reliability Program (SGRP) under the EPRI base program began January 1, 1987. SGRP continued to address the generic steam generator problems facing nuclear utilities. In order to develop appropriate strategies to cope with the tube degradation problems, SGRP has performed the statistical evaluations to model the progression of damage mechanism aimed at accurate prediction of the defect growth rate of various mechanisms such that long term trends can be developed. Analysis of the behavior of group of plants indicate that insights on the potential behavior of a specific plant may be developed from the observed behavior at other plants. SGRP has provided utilities with tube inservice inspection guidelines (ISI Guideline) including ISI Performance Demonstration program to help utilities to improve tube inspection accuracy and sensitivity. SGRP has also updated the secondary Chemistry Guidelines and worked on the advanced amine application guidelines to better protect the steam generator tube corrosion

  10. A three-dimensional laboratory steam injection model allowing in situ saturation measurements. [Comparing steam injection and steam foam injection with nitrogen and without nitrogen

    Energy Technology Data Exchange (ETDEWEB)

    Demiral, B.M.R.; Pettit, P.A.; Castanier, L.M.; Brigham, W.E.

    1992-08-01

    The CT imaging technique together with temperature and pressure measurements were used to follow the steam propagation during steam and steam foam injection experiments in a three dimensional laboratory steam injection model. The advantages and disadvantages of different geometries were examined to find out which could best represent radial and gravity override flows and also fit the dimensions of the scanning field of the CT scanner. During experiments, steam was injected continuously at a constant rate into the water saturated model and CT scans were taken at six different cross sections of the model. Pressure and temperature data were collected with time at three different levels in the model. During steam injection experiments, the saturations obtained by CT matched well with the temperature data. That is, the steam override as observed by temperature data was also clearly seen on the CT pictures. During the runs where foam was present, the saturation distributions obtained from CT pictures showed a piston like displacement. However, the temperature distributions were different depending on the type of steam foam process used. The results clearly show that the pressure/temperature data alone are not sufficient to study steam foam in the presence of non-condensible gas.

  11. Characteristics of steam jet impingement on annulus

    International Nuclear Information System (INIS)

    Yoon, Sang H.; Kim, Won J.; Suh, Kune Y.; Song, Chul H.

    2004-01-01

    The steam jet impingement occurs when the steam through the cold leg from the steam generator strikes the inner reactor barrel during the reflood phase of a loss-of-coolant accident (LOCA), which is a characteristic behavior for the APR1400 (Advanced Power Reactor 1400 MWe). In the cold leg break LOCA, the steam and water flows in the downcomer are truly multidimensional. The azimuthal velocity distribution of the steam flow has an important bearing on the thermal hydraulic phenomena such as the emergency coolant water direct bypass, sweepout, steam condensation, and so forth. The investigation of jet flow is required to determine the steam path and momentum reduction rate after the impingement. For the observation of the steam behavior near the break, the computational fluid dynamic (CFD) analysis has been carried out using CFX5.6. The flow visualization and analysis demonstrate the velocity profiles of the steam flow in the annulus region for the same boundary conditions. Pursuant to the CFD results, the micro-Pitot tubes were positioned at varying angles, and corrected for their sensitivity. The experiments were carried out to directly measure the pressure differential and to visualize the flow utilizing a smoke injection method. Results from this study are slated to be applied to MARS, which is a thermal hydraulic system code for the best-estimate analysis. The current one- or two-dimensional analysis in MARS was known to distort the local flow behavior. To enhance prediction capability of MARS, it is necessary to inspect the steam path in the break flow and mechanically simulate the momentum variation. The present experimental and analytical results can locally be applied to developing the engineering models of specific and essential phenomena. (author)

  12. Remote operated vehicle with carbon dioxide blasting (ROVCO2)

    International Nuclear Information System (INIS)

    Resnick, A.M.

    1995-01-01

    The Remote Operated Vehicle with Carbon Dioxide Blasting (ROVCO 2 ), as shown in a front view, is a six-wheeled remote land vehicle used to decontaminate concrete floors. The remote vehicle has a high pressure Cryogenesis blasting subsystem, Oceaneering Technologies (OTECH) developed a CO 2 xY Orthogonal Translational End Effector (COYOTEE) subsystem, and a vacuum/filtration and containment subsystem. Figure 2 shows a block diagram with the various subsystems labeled

  13. Investigations of primary blast-induced traumatic brain injury

    Science.gov (United States)

    Sawyer, T. W.; Josey, T.; Wang, Y.; Villanueva, M.; Ritzel, D. V.; Nelson, P.; Lee, J. J.

    2018-01-01

    The development of an advanced blast simulator (ABS) has enabled the reproducible generation of single-pulse shock waves that simulate free-field blast with high fidelity. Studies with rodents in the ABS demonstrated the necessity of head restraint during head-only exposures. When the head was not restrained, violent global head motion was induced by pressures that would not produce similar movement of a target the size and mass of a human head. This scaling artefact produced changes in brain function that were reminiscent of traumatic brain injury (TBI) due to impact-acceleration effects. Restraint of the rodent head eliminated these, but still produced subtle changes in brain biochemistry, showing that blast-induced pressure waves do cause brain deficits. Further experiments were carried out with rat brain cell aggregate cultures that enabled the conduct of studies without the gross movement encountered when using rodents. The suspension nature of this model was also exploited to minimize the boundary effects that complicate the interpretation of primary blast studies using surface cultures. Using this system, brain tissue was found not only to be sensitive to pressure changes, but also able to discriminate between the highly defined single-pulse shock waves produced by underwater blast and the complex pressure history exposures experienced by aggregates encased within a sphere and subjected to simulated air blast. The nature of blast-induced primary TBI requires a multidisciplinary research approach that addresses the fidelity of the blast insult, its accurate measurement and characterization, as well as the limitations of the biological models used.

  14. Report on US-Japan 1983 meetings on steam generators

    International Nuclear Information System (INIS)

    1984-04-01

    This is a report on a trip to Japan by personnel of the US Nuclear Regulatory Commission in 1983 to exchange information on steam generators of nuclear power plants. Steam generators of Japanese pressurized water reactors have experienced nearly all of the forms of degradation that have been experienced in US recirculating-type steam generators, except for denting and pitting. More tubes have been plugged per year of reactor operation in Japanese than in US steam generators, but much of the Japanese tube plugging is preventative rather than the result of leaks experienced. The number of leaks per reactor year is much smaller for Japanese than for US steam generators. No steam generators have been replaced in Japan while several have been replaced in the US. The Japanese experience may be related to their very stringent inspection and maintenance programs for steam generators

  15. Thermodynamics of the silica-steam system

    Energy Technology Data Exchange (ETDEWEB)

    Krikorian, Oscar H [Lawrence Radiation Laboratory, University of California, Livermore, CA (United States)

    1970-05-01

    In most nuclear cratering and cavity formation applications, the working fluid in the expanding cavity consists primarily of vaporized silica and steam. The chemical reaction products of silica and steam under these conditions are not known, although it is known that silica is very volatile in the presence of high-pressure steam under certain geologic conditions and in steam turbines. A review is made of work on the silica-steam system in an attempt to determine the vapor species that exist, and to establish the associated thermo-dynamic data. The review indicates that at 600-900 deg K and 1-100 atm steam pressure, Si(OH){sub 4} is the most likely silicon-containing gaseous species. At 600-900 deg. K and 100-1000 atm steam, Si{sub 2}O(OH){sub 6} is believed to predominate, whereas at 1350 deg K and 2000-9000 atm, a mixture of Si(OH){sub 4} and Si{sub 2}O(OH){sub 6} is consistent with the observed volatilities. In work at 1760 deg. K in which silica was reacted either with steam at 0.5 and 1 atm, or with gaseous mixtures of H{sub 2}/H{sub 2}O and O{sub 2}/H{sub 2}O at 1 atm total pressure, only part of the volatility could be accounted for by Si(OH){sub 4}. Hydrogen was found to greatly enhance the volatility of silica, and oxygen to suppress it. The species most likely to explain this behavior is believed to be SiO(OH). A number of other species may also be significant under these conditions. Thermodynamic data have been estimated for all species considered. The Si-OH bond dissociation energy is found to be {approx}117 kcal/mole in both Si(OH){sub 4} and Si{sub 2}O(OH){sub 6}. (author)

  16. Proceedings of the twenty-fourth water reactor safety information meeting. Volume 2: Reactor pressure vessel embrittlement and thermal annealing; Reactor vessel lower head integrity; Evaluation and projection of steam generator tube condition and integrity

    Energy Technology Data Exchange (ETDEWEB)

    Monteleone, S. [comp.] [Brookhaven National Lab., Upton, NY (United States)

    1997-02-01

    This three-volume report contains papers presented at the Twenty-Fourth Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, October 21--23, 1996. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Czech Republic, Finland, France, Japan, Norway, Russia and United Kingdom. This volume is divided into the following sections: reactor pressure vessel embrittlement and thermal annealing; reactor vessel lower head integrity; and evaluation and projection of steam generator tube condition and integrity. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

  17. Proceedings of the twenty-fourth water reactor safety information meeting. Volume 2: Reactor pressure vessel embrittlement and thermal annealing; Reactor vessel lower head integrity; Evaluation and projection of steam generator tube condition and integrity

    International Nuclear Information System (INIS)

    Monteleone, S.

    1997-02-01

    This three-volume report contains papers presented at the Twenty-Fourth Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, October 21--23, 1996. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Czech Republic, Finland, France, Japan, Norway, Russia and United Kingdom. This volume is divided into the following sections: reactor pressure vessel embrittlement and thermal annealing; reactor vessel lower head integrity; and evaluation and projection of steam generator tube condition and integrity. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database

  18. Preliminary evaluation of steam generator tube rupture (SGTR) accident in lead cooled reactor

    International Nuclear Information System (INIS)

    Frano, R. Lo; Forasassi, G.

    2009-01-01

    In this paper some contributions are provided to the development of a European Lead-cooled System, known as the ELSY project (within EU-6 Framework Project); that will constitute a possible reference system for a large lead-cooled reactor of GEN IV. Steam generator (SG) tubing of this system type might be subject to a variety of degradation processes, such as cracking, wall thinning and potential leakage or rupture, eventually leading to the failure of one or more SG tubes that constitute a steam generator tube rupture (SGTR) accident with possible consequences for the safety of the primary systems. It is therefore of interest for the designer to know how the SG itself, as well as the vessel and internals structures, behave under impulsive loading conditions (in form of a rapid and strong increase of pressure) that can arise as consequences of the interaction between the primary and secondary coolants (lead-water interaction). The analysed initiator event, as already mentioned, is a large break (up to a double ended guillotine break) of one (or more) SG cooling tubes that may become severe enough to determine dangerous effects on the interested structures. In order to better simulate and perform the mentioned postulated SGTR accident sequence analyses, an appropriate numerical model with the available computing resources (FEM codes) was set up at the DIMNP of Pisa University. That model was used to evaluate the effects of the propagation of the blast pressure waves inside the SG structures, taking into account also the sloshing phenomenon that could be induced by the lead primary coolant motions. Therefore the SGTR effects study may be considered as a transient and non linear problem the solution of which provides the 'time histories' of hydrodynamic pressures and stresses on the reactor pressure vessel and internals walls. (author)

  19. FY 1999 report on the feasibility study on dissemination of the technologies for effective utilization of blast furnace top pressure in China; 1999 nendo Chugoku ni okeru koro rochoatsu yuko riyo gijutsu no fukyu kanosei chosa hokokusho

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-03-01

    Described herein are the FY 1999 results of the feasibility study on the blast furnace top pressure recovery turbine (TRT) technology in China. Government of China is promoting the steel industry policies to weed out outdated, inefficient production systems while preparing the priority guidance list to promote introduction of advanced technologies, for which the government is implementing or planning the preferential measures, including exemption from the import taxes and VAT, and subsidization of interest on investment. This feasibility study predicts the commercial based investment in the case where the dry dust collector and TRT that are the principal components of a dry type TRT system are manufactured in China and its acquisition ratio is increased on the condition that a suitable number of the systems are installed and on the premise that a dissemination model project is implemented. The results indicate that the system can be introduced at a cost a little less than 60% of the predicted investment in the dissemination model project. The cost-effect ratio varies with power price, which widely varies by region in China, and the dissemination feasibility is studied over a wide range of power price. The system has high dissemination potential, because almost all of the Chinese steel makers will have a sufficient cost-effect ratio which satisfies their investment standards, if they are granted the preferential measures from the government. (NEDO)

  20. Steam generating system in LMFBR type reactors

    International Nuclear Information System (INIS)

    Kurosawa, Katsutoshi.

    1984-01-01

    Purpose: To suppress the thermal shock loads to the structures of reactor system and secondary coolant system, for instance, upon plant trip accompanying turbine trip in the steam generation system of LMFBR type reactors. Constitution: Additional feedwater heater is disposed to the pipeway at the inlet of a steam generator in a steam generation system equipped with a closed loop extended from a steam generator by way of a gas-liquid separator, a turbine and a condensator to the steam generator. The separated water at high temperature and high pressure from a gas-liquid separator is heat exchanged with coolants flowing through the closed loop of the steam generation system in non-contact manner and, thereafter, introduced to a water reservoir tank. This can avoid the water to be fed at low temperature as it is to the steam generator, whereby the thermal shock loads to the structures of the reactor system and the secondary coolant system can be suppressed. (Moriyama, K.)

  1. Fatigue life and damage evolution of martensitic steels for low-pressure steam turbine blades in the VHCF regime; Lebensdauer und Schaedigungsentwicklung martensitischer Staehle fuer Niederdruck-Dampfturbinenschaufeln bei Ermuedungsbeanspruchung im VHCF-Bereich

    Energy Technology Data Exchange (ETDEWEB)

    Kovacs, Stephan

    2014-07-01

    Low-pressure steam turbine blades are usually made of martensitic steels with Cr contents between 9 and 12%, which combine good corrosion resistance, high mechanical strength and sufficient ductility. The inhomogeneous flow field behind the vanes generates high-frequency oscillations above 1 kHz. In addition, the blades with lengths up to 1.5 m are operated at rotational speeds up to 3000 rpm, resulting in large centrifugal forces leading to the superposition of extremely high mean stresses. Also resonance oscillations during start-up and shutdown cannot be completely excluded. Currently, the components are designed using high safety factors against S-N curves with an assumed asymptotic fatigue limit above 107 load cycles. Nevertheless, fatigue cracks are observed even at high number of cycles, starting from the blade root without pre-damage by erosion or steam droplet impingement. While fatigue failure usually occurs at the surface, fatigue cracks at very high number of cycles (> 108) initiate at oxides or intermetallic inclusions below the surface. This transition between both failure mechanisms in the Very High-Cycle Fatigue (VHCF) regime is in the focus of numerous current research activities, because numbers of cycles above 108 can be attained in a viable period of time using the recently developed high-frequency testing techniques operated at 20 kHz. Also for wind turbines, gas turbines, bearings, springs, etc. VHCF issues become increasingly important. Within this work, the fatigue life and damage behavior of a martensitic Cr-steel during fatigue loading with and without high mean stresses at number of cycles to failure above 108 was analyzed. On the one hand, the studies gave insights into the relation between fatigue life and fatigue damage evolution of the investigated group of high-strength steels in the very high cycle fatigue regime (up to 2·109). In particular, the influence of high mean stresses on the VHCF behavior (fracture origin, crack growth

  2. Cheaper power generation from surplus steam generating capacities

    International Nuclear Information System (INIS)

    Gupta, K.

    1996-01-01

    Prior to independence most industries had their own captive power generation. Steam was generated in own medium/low pressure boilers and passed through extraction condensing turbines for power generation. Extraction steam was used for process. With cheaper power made available in Nehru era by undertaking large hydro power schemes, captive power generation in industries was almost abandoned except in sugar and large paper factories, which were high consumers of steam. (author)

  3. Thermal-hydraulic characteristic of the PGV-1000 steam generator

    International Nuclear Information System (INIS)

    Ubra, O.; Doubek, M.

    1995-01-01

    Horizontal steam generators are typical parts of nuclear power plants with pressure water reactor type VVER. By means of this computer program, a detailed thermal-hydraulic study of the horizontal steam generator PGV-1000 has been carried out and a special attention has been paid to the thermal-hydraulics of the secondary side. A set of important steam generator characteristics has been obtained and analyzed. Some of the interesting results of the analysis are presented in the paper. (author)

  4. Liquid-phase problems in steam turbine LP stages

    International Nuclear Information System (INIS)

    Blanc-Feraud, P.

    1978-01-01

    Wet steam formation owing to incipient condensation in final steam turbine pressure stages results in a loss of efficiency and possible rotor blading erosion. The effects of erosion are now clearly understood and quite easily counteracted, but loss of thermodynamics, mechanical and aerodynamic efficiency is still a problem. Only the final LP stages of conventional power station plant operate with wet steam, whereas nuclear plant turbines use it to produce most of their total output [fr

  5. Controlled Low-Pressure Blast-Wave Exposure Causes Distinct Behavioral and Morphological Responses Modelling Mild Traumatic Brain Injury, Post-Traumatic Stress Disorder, and Comorbid Mild Traumatic Brain Injury-Post-Traumatic Stress Disorder.

    Science.gov (United States)

    Zuckerman, Amitai; Ram, Omri; Ifergane, Gal; Matar, Michael A; Sagi, Ram; Ostfeld, Ishay; Hoffman, Jay R; Kaplan, Zeev; Sadot, Oren; Cohen, Hagit

    2017-01-01

    The intense focus in the clinical literature on the mental and neurocognitive sequelae of explosive blast-wave exposure, especially when comorbid with post-traumatic stress-related disorders (PTSD) is justified, and warrants the design of translationally valid animal studies to provide valid complementary basic data. We employed a controlled experimental blast-wave paradigm in which unanesthetized animals were exposed to visual, auditory, olfactory, and tactile effects of an explosive blast-wave produced by exploding a thin copper wire. By combining cognitive-behavioral paradigms and ex vivo brain MRI to assess mild traumatic brain injury (mTBI) phenotype with a validated behavioral model for PTSD, complemented by morphological assessments, this study sought to examine our ability to evaluate the biobehavioral effects of low-intensity blast overpressure on rats, in a translationally valid manner. There were no significant differences between blast- and sham-exposed rats on motor coordination and strength, or sensory function. Whereas most male rats exposed to the blast-wave displayed normal behavioral and cognitive responses, 23.6% of the rats displayed a significant retardation of spatial learning acquisition, fulfilling criteria for mTBI-like responses. In addition, 5.4% of the blast-exposed animals displayed an extreme response in the behavioral tasks used to define PTSD-like criteria, whereas 10.9% of the rats developed both long-lasting and progressively worsening behavioral and cognitive "symptoms," suggesting comorbid PTSD-mTBI-like behavioral and cognitive response patterns. Neither group displayed changes on MRI. Exposure to experimental blast-wave elicited distinct behavioral and morphological responses modelling mTBI-like, PTSD-like, and comorbid mTBI-PTSD-like responses. This experimental animal model can be a useful tool for elucidating neurobiological mechanisms underlying the effects of blast-wave-induced mTBI and PTSD and comorbid mTBI-PTSD.

  6. Slug flow transitions in horizontal gas/liquid two-phase flows. Dependence on channel height and system pressure for air/water and steam/water two-phase flows

    International Nuclear Information System (INIS)

    Nakamura, Hideo

    1996-05-01

    The slug flow transitions and related phenomena for horizontal two-phase flows were studied for a better prediction of two-phase flows that typically appear during the reactor loss-of-coolant accidents (LOCAs). For better representation of the flow conditions experimentally, two large-scaled facility: TPTF for high-pressure steam/water two-phase flows and large duct test facility for air/water two-phase flows, were used. The visual observation of the flow using a video-probe was performed in the TPTF experiments for good understanding of the phenomena. The currently-used models and correlations based mostly on the small-scale low-pressure experiments were reviewed and improved based on these experimental results. The modified Taitel-Dukler model for prediction of transition into slug flow from wavy flow and the modified Steen-Wallis correlation for prediction of onset of liquid entrainment from the interfacial waves were obtained. An empirical correlation for the gas-liquid interfacial friction factor was obtained further for prediction of liquid levels at wavy flow. The region of slug flow regime that is generally under influences of the channel height and system pressure was predicted well when these models and correlations were applied together. (author). 90 refs

  7. Mathematical modeling of control system for the experimental steam generator

    Science.gov (United States)

    Podlasek, Szymon; Lalik, Krzysztof; Filipowicz, Mariusz; Sornek, Krzysztof; Kupski, Robert; Raś, Anita

    2016-03-01

    A steam generator is an essential unit of each cogeneration system using steam machines. Currently one of the cheapest ways of the steam generation can be application of old steam generators came from army surplus store. They have relatively simple construction and in case of not so exploited units - quite good general conditions, and functionality of mechanical components. By contrast, electrical components and control systems (mostly based on relay automatics) are definitely obsolete. It is not possible to use such units with cooperation of steam bus or with steam engines. In particular, there is no possibility for automatically adjustment of the pressure and the temperature of the generated steam supplying steam engines. Such adjustment is necessary in case of variation of a generator load. The paper is devoted to description of improvement of an exemplary unit together with construction of the measurement-control system based on a PLC. The aim was to enable for communication between the steam generator and controllers of the steam bus and steam engines in order to construction of a complete, fully autonomic and maintenance-free microcogeneration system.

  8. Mathematical modeling of control system for the experimental steam generator

    Directory of Open Access Journals (Sweden)

    Podlasek Szymon

    2016-01-01

    Full Text Available A steam generator is an essential unit of each cogeneration system using steam machines. Currently one of the cheapest ways of the steam generation can be application of old steam generators came from army surplus store. They have relatively simple construction and in case of not so exploited units – quite good general conditions, and functionality of mechanical components. By contrast, electrical components and control systems (mostly based on relay automatics are definitely obsolete. It is not possible to use such units with cooperation of steam bus or with steam engines. In particular, there is no possibility for automatically adjustment of the pressure and the temperature of the generated steam supplying steam engines. Such adjustment is necessary in case of variation of a generator load. The paper is devoted to description of improvement of an exemplary unit together with construction of the measurement-control system based on a PLC. The aim was to enable for communication between the steam generator and controllers of the steam bus and steam engines in order to construction of a complete, fully autonomic and maintenance-free microcogeneration system.

  9. Steam turbine of WWER-1000 unit

    International Nuclear Information System (INIS)

    Drahy, J.

    1986-01-01

    The manufacture was started by Skoda of a saturated steam, 1,000 MW, 3,000 rpm turbine designed for the Temelin nuclear power plant. The turbine provides steam for heating water for district heating, this either with an output of 893 MW for a three-stage water heating at 150/60 degC, or of 570 MW for a two-stage water heating at 120/60 degC. The turbine features one high-pressure and three identical low-pressure stages. The pressure gradient between the high-pressure and the low-pressure parts was optimized with respect to the thermal efficiency of the cycle and to the thermodynamic efficiency of the low-pressure part. A value of 0.79 MPa was selected corresponding to the maximum through-flow of steam entering the turbine. This makes 5,495 t/h, the admission steam parameters are 273.3 degC and 5.8 MPa. The feed water temperature is 220.9 degC. 300 cold starts, 1,000 starts after shutdowns for 55 to 88 hours and 600 starts after shutdown for 8 hours are envisaged for the entire turbine service life. (Z.M.). 5 figs., 1 tab., 6 refs

  10. Method for operating a steam turbine of the nuclear type with electronic reheat control of a cycle steam reheater

    International Nuclear Information System (INIS)

    Luongo, M.C.

    1975-01-01

    An electronic system is provided for operating a nuclear electric power plant with electronic steam reheating control applied to the nuclear turbine system in response to low pressure turbine temperatures, and the control is adapted to operate in a plurality of different automatic control modes to control reheating steam flow and other steam conditions. Each of the modes of control permit turbine temperature variations within predetermined constraints and according to predetermined functions of time. (Official Gazette)

  11. Numerical simulation in steam injection process by a mechanistic approach

    Energy Technology Data Exchange (ETDEWEB)

    De Souza, J.C.Jr.; Campos, W.; Lopes, D.; Moura, L.S.S. [Petrobras, Rio de Janeiro (Brazil)

    2008-10-15

    Steam injection is a common thermal recovery method used in very viscous oil reservoirs. The method involves the injection of heat to reduce viscosity and mobilize oil. A steam generation and injection system consists primarily of a steam source, distribution lines, injection wells and a discarding tank. In order to optimize injection and improve the oil recovery factor, one must determine the parameters of steam flow such as pressure, temperature and steam quality. This study focused on developing a unified mathematical model by means of a mechanistic approach for two-phase steam flow in pipelines and wells. The hydrodynamic and heat transfer mechanistic model was implemented in a computer simulator to model the parameters of steam injection while trying to avoid the use of empirical correlations. A marching algorithm was used to determine the distribution of pressure and temperature along the pipelines and wellbores. The mathematical model for steam flow in injection systems, developed by a mechanistic approach (VapMec) performed well when the simulated values of pressures and temperatures were compared with the values measured during field tests. The newly developed VapMec model was incorporated in the LinVap-3 simulator that constitutes an engineering supporting tool for steam injection wells operated by Petrobras. 23 refs., 7 tabs., 6 figs.

  12. Steam generator tubesheet waterlancing at Bruce B

    Energy Technology Data Exchange (ETDEWEB)

    Persad, R. [Babcock and Wilcox Canada, Cambridge, Ontario (Canada); Eybergen, D. [Bruce Power, Tiverton, Ontario (Canada)

    2006-07-01

    High pressure water cleaning of steam generator secondary side tubesheet surfaces is an important and effective strategy for reducing or eliminating under-deposit chemical attack of the tubing. At the Bruce B station, reaching the interior of the tube bundle with a high-pressure water lance is particularly challenging due to the requirement to setup on-boiler equipment within the containment bellows. This paper presents how these and other design constraints were solved with new equipment. Also discussed is the application of new high-resolution inter-tube video probe capability to the Bruce B steam generator tubesheets. (author)

  13. Failure analysis of retired steam generator tubings

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hong Pyo; Kim, J. S.; Hwang, S. S. and others

    2005-04-15

    Degradation of steam generator leads to forced outage and extension of outage, which causes increase in repair cost, cost of purchasing replacement power and radiation exposure of workers. Steam generator tube rupture incident occurred in Uljin 4 in 2002, which made public sensitive to nuclear power plant. To keep nuclear energy as a main energy source, integrity of steam generator should be demonstrated. Quantitative relationship between ECT(eddy current test) signal and crack size is needed in assesment of integrity of steam generator in pressurized water reactor. However, it is not fully established for application in industry. Retired steam generator of Kori 1 has many kinds of crack such as circumferential and axial primary water stress corrosion crack and outer diameter stress corrosion crack(ODSCC). So, it can be used in qualifying and improving ECT technology and in condition monitoring assesment for crack detected in ISI(in service inspection). In addition, examination of pulled tube of Kori 1 retired steam generator will give information about effectiveness of non welded sleeving technology which was employed to repair defect tubes and remedial action which was applied to mitigate ODSCC. In this project, hardware such as semi hot lab. for pulled tube examination and modification transportation cask for pulled tube and software such as procedure of transportation of radioactive steam generator tube and non-destructive and destructive examination of pulled tube were established. Non-destructive and destructive examination of pulled tubes from Kori 1 retired steam generator were performed in semi hot lab. Remedial actions applied to Kori 1 retired steam generator, PWSCC trend and bulk water chemistry and crevice chemistry in Kori 1 were evaluated. Electrochemical decontamination technology for pulled tube was developed to reduce radiation exposure and enhance effectiveness of pulled tube examination. Multiparameter algorithm developed at ANL, USA was

  14. Power plant and system for accelerating a cross compound turbine in such plant, especially one having an HTGR steam supply

    International Nuclear Information System (INIS)

    Jaegtnes, K.O.; Braytenbah, A.S.

    1977-01-01

    An electric power plant having a cross compound steam turbine and a steam source that includes a high temperature gas-cooled nuclear reactor is described. The steam turbine includes high and intermediate-pressure portions which drive a first generating means, and a low-pressure portion which drives a second generating means. The steam source supplies superheat steam to the high-pressure turbine portion, and an associated bypass permits the superheat steam to flow from the source to the exhaust of the high-pressure portion. The intermediate and low-pressure portions use reheat steam; an associated bypass permits reheat steam to flow from the source to the low-pressure exhaust. An auxiliary turbine driven by steam exhausted from the high-pressure portion and its bypass drives a gas blower to propel the coolant gas through the reactor. While the bypass flow of reheat steam is varied to maintain an elevated pressure of reheat steam upon its discharge from the source, both the first and second generating means and their associated turbines are accelerated initially by admitting steam to the intermediate and low-pressure portions. The electrical speed of the second generating means is equalized with that of the first generating means, whereupon the generating means are connected and acceleration proceeds under control of the flow through the high-pressure portion. 29 claims, 2 figures

  15. Evaluation of acoustic resonance at branch section in main steam line. Part 1. Effects of steam wetness on acoustic resonance

    International Nuclear Information System (INIS)

    Uchiyama, Yuta; Morita, Ryo

    2011-01-01

    The power uprating of the nuclear power plant (NPP) is conducted in United States, EU countries and so on, and also is planned in Japan. However, the degradation phenomena such as flow-induced vibration and wall thinning may increase or expose in the power uprate condition. In U.S. NPP, the dryer had been damaged by high cycle fatigue due to acoustic-induced vibration under a 17% extended power uprating (EPU) condition. This is caused by acoustic resonance at the stub pipes of safety relief valves (SRVs) in the main steam lines (MSL). Increased velocity by uprating excites the pressure fluctuations and makes large amplitude resonance. To evaluate the acoustic resonance at the stub pipes of SRVs in actual BWR, it is necessary to clarify the acoustic characteristics in steam flow. Although there are several previous studies about acoustic resonance, most of them are not steam flow but air flow. Therefore in this study, to investigate the acoustic characteristics in steam flow, we conducted steam flow experiments in each dry and wet steam conditions, and also nearly saturated condition. We measured pressure fluctuation at the top of the single stub pipe and in main steam piping. As a result, acoustic resonance in dry steam flow could be evaluated as same as that in air flow. It is clarified that resonance amplitude of fluctuating pressure at the top of the stub pipe in wet steam was reduced to one-tenth compared with that in dry. (author)

  16. Blast Load Response of Steel Sandwich Panels with Liquid Encasement

    Energy Technology Data Exchange (ETDEWEB)

    Dale Karr; Marc Perlin; Benjamin Langhorst; Henry Chu

    2009-10-01

    We describe an experimental investigation of the response of hybrid blast panels for protection from explosive and impact forces. The fundamental notion is to dissipate, absorb, and redirect energy through plastic collapse, viscous dissipation, and inter-particle forces of liquid placed in sub-structural compartments. The panels are designed to absorb energy from an impact or air blast by elastic-plastic collapse of the panel substructure that includes fluid-filled cavities. The fluid contributes to blast effects mitigation by providing increased initial mass and resistance, by dissipation of energy through viscosity and fluid flow, and by redirecting the momentum that is imparted to the system from the impact and blast impulse pressures. Failure and deformation mechanisms of the panels are described.

  17. Condensation of steam

    International Nuclear Information System (INIS)

    Prisyazhniuk, V.A.

    2002-01-01

    An equation for nucleation kinetics in steam condensation has been derived, the equation taking into account the concurrent and independent functioning of two nucleation mechanisms: the homogeneous one and the heterogeneous one. The equation is a most general-purpose one and includes all the previously known condensation models as special cases. It is shown how the equation can be used in analyzing the process of steam condensation in the condenser of an industrial steam-turbine plant, and in working out new ways of raising the efficiency of the condenser, as well as of the steam-turbine plant as a whole. (orig.)

  18. EPRI steam generator programs

    International Nuclear Information System (INIS)

    Martel, L.J.; Passell, T.O.; Bryant, P.E.C.; Rentler, R.M.

    1977-01-01

    The paper describes the current overall EPRI steam generator program plan and some of the ongoing projects. Because of the recent occurrence of a corrosion phenomenon called ''denting,'' which has affected a number of operating utilities, an expanded program plan is being developed which addresses the broad and urgent needs required to achieve improved steam generator reliability. The goal of improved steam generator reliability will require advances in various technologies and also a management philosophy that encourages conscientious efforts to apply the improved technologies to the design, procurement, and operation of plant systems and components that affect the full life reliability of steam generators

  19. Modelling of the steam-water-countercurrent flow in the rewetting and flooding phase after loss-of-coolant accidents in pressurized water reactors

    International Nuclear Information System (INIS)

    Curca-Tivig, F.

    1990-01-01

    A new interphase momentum exchange model has been developed to simulate the Refill- Reflood Phase after LOCAs. Special phenomena of steam/water- countercurrent flow - like limitation or onset of downward-watee penetration - have been modelled and integrated into a flooding model. The interphase momentum exchange model interconnected with the flooding model has been implemented into the advanced system code RELAP5/MOD1. The new version of this code can now be utilized to predict the hot leg emergency-core-cooling (ECC) injection for German PWRs. The interfacial momentum transfer model developed includes the interphase frictional drag, the force due to virtual mass and the momenta due to interphase mass transfer. The modelling of the interfacial shear or drag accounts for the effects of phase and velocity profiles. The flooding model predicts countercurrent-flow limitation, onset of water penetration and partial delivery. The flooding correlation specifies the maximum down flow liquid velocity in case of countercurrent flow through flow restrictions for a given vapor velocity. (orig./HP) [de

  20. Steam cracking of hydrocarbons 3. Straight-run naphtha

    NARCIS (Netherlands)

    Bajus, M.; Vesely, V.; Leclercq, P.A.; Rijks, J.A.

    1980-01-01

    Steam cracking of straight-run naphtha from Romashkino crude oil was investigated in quartz and stainless steel reactors with a relatively large ratio of Inner surface to volume. The experiments were performed at atmospheric pressure at 780-800 OC for starting ratios of steam to naphtha between 0.5

  1. Sludge cleaning in the steam generators: sludge Lancing e IBL

    International Nuclear Information System (INIS)

    Montoro, E.; Gonzalez, S.; Calderon, N.

    2013-01-01

    IBERDROLA Engineering and Construction has echoed the need for plants to remove oxide deposits (sludge) located on the secondary side, on the bottom plate and into the tube bundle steam steam generators. Therefore, and with its partner SAVAC SRA has developed a specific system consisting of applying a capillary water at very high pressure applied directly to the location of these oxides. (Author)

  2. Production analysis of methanol and hydrogen of a modificated blast furnace gas using nuclear energy of the high temperature reactor

    International Nuclear Information System (INIS)

    Peschel, W.

    1985-12-01

    Modern blast furnaces are operated with a coke ration of 500 kg/t pig iron. The increase of the coke ratio to 1000 kg/t pig iron raises the content of carbon monoxide and hydrogen in the blast furnace gas. On the basis of a blast furnace gas modificated in such a way, the production of methanol and hydrogen is investigated under the coupling of current and process heat from the high temperature reactor. Moreover the different variants are discussed, for which respectively a material and energetic balance as well as an estimation of the production costs is performed. Regarding the subsequent treatment of the blast furnace gas it turns out favourably in principle to operate the blast furnace with a nitrogen-free wind consisting only of oxygen and steam. The production costs show a strong dependence on the raw material costs, whose influence is shown in a nomograph. (orig.) [de

  3. Steam injection for heavy oil recovery: Modeling of wellbore heat efficiency and analysis of steam injection performance

    International Nuclear Information System (INIS)

    Gu, Hao; Cheng, Linsong; Huang, Shijun; Li, Bokai; Shen, Fei; Fang, Wenchao; Hu, Changhao

    2015-01-01

    Highlights: • A comprehensive mathematical model was established to estimate wellbore heat efficiency of steam injection wells. • A simplified approach of predicting steam pressure in wellbores was proposed. • High wellhead injection rate and wellhead steam quality can improve wellbore heat efficiency. • High wellbore heat efficiency does not necessarily mean good performance of heavy oil recovery. • Using excellent insulation materials is a good way to save water and fuels. - Abstract: The aims of this work are to present a comprehensive mathematical model for estimating wellbore heat efficiency and to analyze performance of steam injection for heavy oil recovery. In this paper, we firstly introduce steam injection process briefly. Secondly, a simplified approach of predicting steam pressure in wellbores is presented and a complete expression for steam quality is derived. More importantly, both direct and indirect methods are adopted to determine the wellbore heat efficiency. Then, the mathematical model is solved using an iterative technique. After the model is validated with measured field data, we study the effects of wellhead injection rate and wellhead steam quality on steam injection performance reflected in wellbores. Next, taking cyclic steam stimulation as an example, we analyze steam injection performance reflected in reservoirs with numerical reservoir simulation method. Finally, the significant role of improving wellbore heat efficiency in saving water and fuels is discussed in detail. The results indicate that we can improve the wellbore heat efficiency by enhancing wellhead injection rate or steam quality. However, high wellbore heat efficiency does not necessarily mean satisfactory steam injection performance reflected in reservoirs or good performance of heavy oil recovery. Moreover, the paper shows that using excellent insulation materials is a good way to save water and fuels due to enhancement of wellbore heat efficiency

  4. Corrosion aspects in steam generators of nuclear power plants

    International Nuclear Information System (INIS)

    Visoni, E.; Santos Pinto, M. dos

    1988-01-01

    Steam generators of pressurized water reactors (PWR), transfer heat from a primary coolant system to a secondary coolant system. Primary coolant water is heated in the core and passes through the steam generator that transfer heat to the secondary coolant water. However, the steam generator is dead for ionic impurities, corrosion products and fabrication/maintenence residues. These impurities concentrate between crevice and cracks. Many types of degradation mechanisms affect the tubes. The tubes are dented, craked, ovalized, wasted, etc. This paper describes the main corrosion problems in steam generators and includes the corrective actions to considered to reduce or eliminate these corrosion problems. (author) [pt

  5. Technology of turbine plant operating with wet steam

    International Nuclear Information System (INIS)

    1989-01-01

    The technology of turbine plant operating with wet steam is a subject of continuing interest and importance, notably in view of the widespread use of wet steam cycles in nuclear power plants and the recent developments of advanced low pressure blading for both conventional and wet steam turbines. The nature of water formation in expanding steam has an important influence on the efficiency of turbine blading and on the integrity and safe operating life of blading and associated turbine and plant components. The subjects covered in this book include research, flow analysis and measurement, development and design of turbines and ancillary plant, selection of materials of construction, manufacturing methods and operating experience. (author)

  6. Steam generator operating experience update, 1982-1983

    International Nuclear Information System (INIS)

    Frank, L.

    1984-06-01

    This report is a continuation of earlier reports by the staff addressing pressurized water reactor steam generator operating experience. NUREG-0886, Steam Generator Tube Experience, published in February 1982 summarized experience in domestic and foreign plants through December 1981. This report summarizes steam generator operating experience in domestic plants for the years 1982 and 1983. Included are new problems encountered with secondary-side loose parts, sulfur-induced stress-assisted corrosion cracking, and flow-induced vibrational wear in the new preheater design steam generators. The status of Unresolved Safety Issues A3, A4, and A5 is also discussed

  7. A steam separator-superheater apparatus

    International Nuclear Information System (INIS)

    Androw, Jean; Bessouat, Roger; Peyrelongue, J.-P.

    1973-01-01

    Description is given of a separator-superheater apparatus comprising an outer enclosure containing a separating-unit and a steam superheating unit according to the main patent. The present addition relates to an improvement in that apparatus, characterized in that the separating unit and the superheating unit, mounted in two distinct portions of the outer enclosure, are divided into the same number of sub-units of each unit being identical and operating in parallel, and in that to each separator sub-unit is associated a superheater sub-unit, said sub-units being mounted in series and located in one in the other of the enclosure two portions, respectively. This can be applied to the treatment of the exhaust steam of a turbine high pressure body, prior to re-injecting said steam into the low pressure body [fr

  8. Shiraz solar power plant operation with steam engine

    International Nuclear Information System (INIS)

    Yaghoubi, M.; Azizian, K.

    2004-01-01

    The present industrial developments and daily growing need of energy, as well as economical and environmental problem caused by fossil fuels consumption, resulted certain constraint for the future demand of energy. During the past two decades great attention has been made to use renewable energy for different sectors. In this regard for the first time in Iran, design and construction of a 250 K W Solar power plant in Shiraz, Iran is being carried out and it will go to operation within next year. The important elements of this power plant is an oil cycle and a steam cycle, and several studies have been done about design and operation of this power plant, both for steady state and transient conditions. For the steam cycle, initially a steam turbine was chosen and due to certain limitation it has been replaced by a steam engine. The steam engine is able to produce electricity with hot or saturated vapor at different pressures and temperatures. In this article, the effects of installing a steam engine and changing its vapor inlet pressure and also the effects of sending hot or saturated vapor to generate electricity are studied. Various cycle performance and daily electricity production are determined. The effects of oil cycle temperature on the collector field efficiency, and daily, monthly and annual amount of electricity production is calculated. Results are compared with the steam cycle output when it contains a steam turbine. It is found that with a steam engine it is possible to produce more annual electricity for certain conditions

  9. The BLAST experiment

    NARCIS (Netherlands)

    Hasell, D.; Botto, T.; van den Brand, J.F.J.

    2009-01-01

    The Bates large acceptance spectrometer toroid (BLAST) experiment was operated at the MIT-Bates Linear Accelerator Center from 2003 until 2005. The detector and experimental program were designed to study, in a systematic manner, the spin-dependent electromagnetic interaction in few-nucleon systems.

  10. Reciprocating wear in a steam environment

    Energy Technology Data Exchange (ETDEWEB)

    Brown, L.J.; Gee, M.G. [National Physical Laboratory, Teddington, Middlesex (United Kingdom)

    2010-07-01

    Tests to simulate the wear between sliding components in steam power plant have been performed using a low frequency wear apparatus at elevated temperatures under static load, at ambient pressure, in a steam environment. The apparatus was modified to accept a novel method of steam delivery. The materials tested were pre-exposed in a flowing steam furnace at temperature for either 500 or 3000 hours to provide some simulation of long term ageing. The duration of each wear test was 50 hours and tests were also performed on as-received material for comparison purposes. Data has been compared with results of tests performed on non-oxidised material for longer durations and also on tests without steam to examine the effect of different environments. Data collected from each test consists of mass change, stub height measurement and friction coefficient as well as visual inspection of the wear track. Within this paper, it is reported that both pre-ageing and the addition of steam during testing clearly influence the friction between material surfaces. (orig.)

  11. Oil injection into the blast furnace

    Energy Technology Data Exchange (ETDEWEB)

    Dongsheng Liao; Mannila, P.; Haerkki, J.

    1997-12-31

    Fuel injection techniques have been extensively used in the commercial blast furnaces, a number of publications concerning the fuels injection have been reported. This present report only summarizes the study achievements of oil injection due to the research need the of authors, it includes the following parts: First, the background and the reasons reducing coke rate of oil injection are analyzed. Reducing coke rate and decreasing the ironmaking costs are the main deriving forces, the contents of C, H and ash are direct reasons reducing coke rate. It was also found that oil injection had great effects on the state of blast furnace, it made operation stable, center gas flow develop fully, pressure drop increase, descent speed of burden materials decrease and generation of thermal stagnation phenomena, the quality of iron was improved. Based on these effects, as an ideal mean, oil injection was often used to adjust the state of blast furnace. Secondly, combustion behavior of oil in the raceway and tuyere are discussed. The distribution of gas content was greatly changed, the location of CO, H{sub 2} generation was near the tuyere; the temperature peak shifts from near the raceway boundary to the tuyere. Oxygen concentration and blast velocity were two important factors, it was found that increasing excess oxygen ratio 0.9 to 1.3, the combustion time of oil decreases 0.5 msec, an increase of the blast velocity results in increasing the flame length. In addition, the nozzle position and oil rate had large effects on the combustion of oil. Based on these results, the limit of oil injection is also discussed, soot formation is the main reason limiting to further increase oil injection rate, it was viewed that there were three types of soot which were generated under blast furnace operating conditions. The reason generating soot is the incomplete conversion of the fuel. Finally, three methods improving combustion of oil in the raceway are given: Improvement of oil

  12. Dismantling of the 50 MW steam generator test facility

    International Nuclear Information System (INIS)

    Nakai, S.; Onojima, T.; Yamamoto, S.; Akai, M.; Isozaki, T.; Gunji, M.; Yatabe, T.

    1997-01-01

    We have been dismantling the 50MW Steam Generator Test Facility (50MWSGTF). The objectives of the dismantling are reuse of sodium components to a planned large scale thermal hydraulics sodium test facility and the material examination of component that have been operated for long time in sodium. The facility consisted of primary sodium loop with sodium heater by gas burner as heat source instead of reactor, secondary sodium loop with auxiliary cooling system (ACS) and water/steam system with steam temperature and pressure reducer instead of turbine. It simulated the 1 loop of the Monju cooling system. The rated power of the facility was 50MWt and it was about 1/5 of the Monju power plant. Several sodium removal methods are applied. As for the components to be dismantled such as piping, intermediate heat exchanger (IHX), air cooled heat exchangers (AC), sodium is removed by steam with nitrogen gas in the air or sodium is burned in the air. As for steam generators which material tests are planned, sodium is removed by steam injection with nitrogen gas to the steam generator. The steam generator vessel is filled with nitrogen and no air in the steam generator during sodium removal. As for sodium pumps, pump internal structure is pulled out from the casing and installed into the tank. After the installation, sodium is removed by the same method of steam generator. As for relatively small reuse components such as sodium valves, electromagnet flow meters (EMFs) etc., sodium is removed by alcohol process. (author)

  13. Steam feeding redundancy for turbine-drives of feed pumps at WWER-1000 NPP

    International Nuclear Information System (INIS)

    Nesterov, Yu.V.; Shmukler, B.I.

    1987-01-01

    The system of steam supply for feed pump driving turbines (T) at the South Ukrainian Unit 1 according to the centralized redundancy principle is described. T is feeded through the collector of water auxiliary sytem (CWAS) to which steam from the third steam extraction line of turbine is supplied under thenormal regime. Under the reduction of turbine load, live steam from the steam generator is supplied to CWAS through the pressure regulator, possesing 10 s speed of responce. In this case the level reduction in the steam generator makes up 170 mm

  14. Lateral blasts at Mount St. Helens and hazard zonation

    Science.gov (United States)

    Crandell, D.R.; Hoblitt, R.P.

    1986-01-01

    Lateral blasts at andesitic and dacitic volcanoes can produce a variety of direct hazards, including ballistic projectiles which can be thrown to distances of at least 10 km and pyroclastic density flows which can travel at high speed to distances of more than 30 km. Indirect effect that may accompany such explosions include wind-borne ash, pyroclastic flows formed by the remobilization of rock debris thrown onto sloping ground, and lahars. Two lateral blasts occurred at a lava dome on the north flank of Mount St. Helens about 1200 years ago; the more energetic of these threw rock debris northeastward across a sector of about 30?? to a distance of at least 10 km. The ballistic debris fell onto an area estimated to be 50 km2, and wind-transported ash and lapilli derived from the lateral-blast cloud fell on an additional lobate area of at least 200 km2. In contrast, the vastly larger lateral blast of May 18, 1980, created a devastating pyroclastic density flow that covered a sector of as much as 180??, reached a maximum distance of 28 km, and within a few minutes directly affected an area of about 550 km2. The May 18 lateral blast resulted from the sudden, landslide-induced depressurization of a dacite cryptodome and the hydrothermal system that surrounded it within the volcano. We propose that lateral-blast hazard assessments for lava domes include an adjoining hazard zone with a radius of at least 10 km. Although a lateral blast can occur on any side of a dome, the sector directly affected by any one blast probably will be less than 180??. Nevertheless, a circular hazard zone centered on the dome is suggested because of the difficulty of predicting the direction of a lateral blast. For the purpose of long-term land-use planning, a hazard assessment for lateral blasts caused by explosions of magma bodies or pressurized hydrothermal systems within a symmetrical volcano could designate a circular potential hazard area with a radius of 35 km centered on the volcano

  15. Automatic system for redistributing feedwater in a steam generator of a nuclear power plant

    International Nuclear Information System (INIS)

    Fuoto, J.S.; Crotzer, M.E.; Lang, G.E.

    1980-01-01

    A system is described for automatically redistributing a steam generator secondary tube system after a burst in the secondary tubing. This applies to a given steam generator in a system having several steam generators partially sharing a common tube system, and employs a pressure control generating an electrical signal which is compared with given values [fr

  16. Improvements to the COBRA-TF (EPRI) computer code for steam generator analysis. Final report

    International Nuclear Information System (INIS)

    Stewart, C.W.; Barnhart, J.S.; Koontz, A.S.

    1980-09-01

    The COBRA-TF (EPRI) code has been improved and extended for pressurized water reactor steam generator analysis. New features and models have been added in the areas of subcooled boiling and heat transfer, turbulence, numerics, and global steam generator modeling. The code's new capabilities are qualified against selected experimental data and demonstrated for typical global and microscale steam generator analysis

  17. Production of D-lactic acid from sugarcane bagasse using steam-explosion

    Science.gov (United States)

    Sasaki, Chizuru; Okumura, Ryosuke; Asakawa, Ai; Asada, Chikako; Nakamura, Yoshitoshi

    2012-03-01

    This study investigated the production of D-lactic acid from unutilized sugarcane bagasse using steam explosion pretreatment. The optimal steam pressure for a steaming time of 5 min was determined. By enzymatic saccharification using Meicellase, the highest recovery of glucose from raw bagasse, 73.7%, was obtained at a steam pressure of 20 atm. For residue washed with water after steam explosion, the glucose recovery increased up to 94.9% at a steam pressure of 20 atm. These results showed that washing with water is effective in removing enzymatic reaction inhibitors. After steam pretreatment (steam pressure of 20 atm), D-lactic acid was produced by Lactobacillus delbrueckii NBRC 3534 from the enzymatic hydrolyzate of steam-exploded bagasse and washed residue. The conversion rate of D-lactic acid obtained from the initial glucose concentration was 66.6% for the hydrolyzate derived from steam-exploded bagasse and 90.0% for that derived from the washed residue after steam explosion. These results also demonstrated that the hydrolyzate of steam-exploded bagasse (without washing with water) contains fermentation inhibitors and washing with water can remove them.

  18. 49 CFR 230.37 - Steam test following repairs or alterations.

    Science.gov (United States)

    2010-10-01

    ... 49 Transportation 4 2010-10-01 2010-10-01 false Steam test following repairs or alterations. 230... RAILROAD ADMINISTRATION, DEPARTMENT OF TRANSPORTATION STEAM LOCOMOTIVE INSPECTION AND MAINTENANCE STANDARDS Boilers and Appurtenances Pressure Testing of Boilers § 230.37 Steam test following repairs or alterations...

  19. Production of D-lactic acid from sugarcane bagasse using steam-explosion

    International Nuclear Information System (INIS)

    Sasaki, Chizuru; Okumura, Ryosuke; Asakawa, Ai; Asada, Chikako; Nakamura, Yoshitoshi

    2012-01-01

    This study investigated the production of D-lactic acid from unutilized sugarcane bagasse using steam explosion pretreatment. The optimal steam pressure for a steaming time of 5 min was determined. By enzymatic saccharification using Meicellase, the highest recovery of glucose from raw bagasse, 73.7%, was obtained at a steam pressure of 20 atm. For residue washed with water after steam explosion, the glucose recovery increased up to 94.9% at a steam pressure of 20 atm. These results showed that washing with water is effective in removing enzymatic reaction inhibitors. After steam pretreatment (steam pressure of 20 atm), D-lactic acid was produced by Lactobacillus delbrueckii NBRC 3534 from the enzymatic hydrolyzate of steam-exploded bagasse and washed residue. The conversion rate of D-lactic acid obtained from the initial glucose concentration was 66.6% for the hydrolyzate derived from steam-exploded bagasse and 90.0% for that derived from the washed residue after steam explosion. These results also demonstrated that the hydrolyzate of steam-exploded bagasse (without washing with water) contains fermentation inhibitors and washing with water can remove them.

  20. STEAM by Design

    Science.gov (United States)

    Keane, Linda; Keane, Mark

    2016-01-01

    We live in a designed world. STEAM by Design presents a transdisciplinary approach to learning that challenges young minds with the task of making a better world. Learning today, like life, is dynamic, connected and engaging. STEAM (Science, Technology, Environment, Engineering, Art, and Math) teaching and learning integrates information in…

  1. Steampunk: Full Steam Ahead

    Science.gov (United States)

    Campbell, Heather M.

    2010-01-01

    Steam-powered machines, anachronistic technology, clockwork automatons, gas-filled airships, tentacled monsters, fob watches, and top hats--these are all elements of steampunk. Steampunk is both speculative fiction that imagines technology evolved from steam-powered cogs and gears--instead of from electricity and computers--and a movement that…

  2. Safety Picks up "STEAM"

    Science.gov (United States)

    Roy, Ken

    2016-01-01

    This column shares safety information for the classroom. STEAM subjects--science, technology, engineering, art, and mathematics--are essential for fostering students' 21st-century skills. STEAM promotes critical-thinking skills, including analysis, assessment, categorization, classification, interpretation, justification, and prediction, and are…

  3. Steam-water separator

    International Nuclear Information System (INIS)

    Modrak, T.M.; Curtis, R.W.

    1978-01-01

    A two-stage steam-water separating device is introduced, where the second stage is made as a cyclone separator. The water separated here is collected in the first stage of the inner tube and is returned to the steam raising unit. (TK) [de

  4. Steam power plant

    International Nuclear Information System (INIS)

    Campbell, J.W.E.

    1981-01-01

    This invention relates to power plant forced flow boilers operating with water letdown. The letdown water is arranged to deliver heat to partly expanded steam passing through a steam reheater connected between two stages of the prime mover. (U.K.)

  5. Design of blast simulators for nuclear testing

    International Nuclear Information System (INIS)

    Mark, A.; Opalka, K.O.; Kitchens, C.W. Jr.

    1983-01-01

    A quasi-one-dimensional computational technique is used to model the flow of a large, complicated shock tube. The shock tube, or Large Blast Simulator, is used to simulate conventional or nuclear explosions by shaping the pressure history. Results from computations show favorable agreement when compared with data taken in the facility at Gramat, France. Such future shock tubes will include a thermal irradiation capability to better simulate a nuclear event. The computations point to the need for venting of the combustion products since the pressure history will be considerably altered as the shock propagates through these hot gases

  6. Cogeneration steam turbines from Siemens: New solutions

    Science.gov (United States)

    Kasilov, V. F.; Kholodkov, S. V.

    2017-03-01

    The Enhanced Platform system intended for the design and manufacture of Siemens AG turbines is presented. It combines organizational and production measures allowing the production of various types of steam-turbine units with a power of up to 250 MWel from standard components. The Enhanced Platform designs feature higher efficiency, improved reliability, better flexibility, longer overhaul intervals, and lower production costs. The design features of SST-700 and SST-900 steam turbines are outlined. The SST-700 turbine is used in backpressure steam-turbine units (STU) or as a high-pressure cylinder in a two-cylinder condensing turbine with steam reheat. The design of an SST-700 single-cylinder turbine with a casing without horizontal split featuring better flexibility of the turbine unit is presented. An SST-900 turbine can be used as a combined IP and LP cylinder (IPLPC) in steam-turbine or combined-cycle power units with steam reheat. The arrangements of a turbine unit based on a combination of SST-700 and SST-900 turbines or SST-500 and SST-800 turbines are presented. Examples of this combination include, respectively, PGU-410 combinedcycle units (CCU) with a condensing turbine and PGU-420 CCUs with a cogeneration turbine. The main equipment items of a PGU-410 CCU comprise an SGT5-4000F gas-turbine unit (GTU) and STU consisting of SST-700 and SST-900RH steam turbines. The steam-turbine section of a PGU-420 cogeneration power unit has a single-shaft turbine unit with two SST-800 turbines and one SST-500 turbine giving a power output of N el. STU = 150 MW under condensing conditions.

  7. Lifetime of superheated steam components

    International Nuclear Information System (INIS)

    Stoklossa, K.H.; Oude-Hengel, H.H.; Kraechter, H.J.

    1974-01-01

    The current evaluation schemes in use for judging the lifetime expectations of superheated steam components are compared with each other. The influence of pressure and temperature fluctuations, the differences in the strength of the wall, and the spread band of constant-strainrates are critically investigated. The distribution of these contributory effects are demonstrated in the hight of numerous measuring results. As an important supplement to these evaluation schemes a newly developed technique is introduced which is designed to calculate failure probabilities. (orig./RW) [de

  8. Steam explosion triggering and efficiency studies

    International Nuclear Information System (INIS)

    Buxton, L.D.; Nelson, L.S.; Benedick, W.B.

    1979-01-01

    Laboratory experiments on the thermal interaction of simulated light water reactor (LWR) fuel melts and water are summarized. Their purpose was to investigate the possibility of steam explosions occurring for a range of hypothetical accident conditions. Pressure, temperature, hot liquid motion and cold liquid motion were monitored during the experiments

  9. Atmospheric particulate emissions from dry abrasive blasting using coal slag

    Energy Technology Data Exchange (ETDEWEB)

    Bhaskar Kura; Kalpalatha Kambham; Sivaramakrishnan Sangameswaran; Sandhya Potana [University of New Orleans, New Orleans, LA (United States). Department of Civil and Environmental Engineering

    2006-08-15

    Coal slag is one of the widely used abrasives in dry abrasive blasting. Atmospheric emissions from this process include particulate matter (PM) and heavy metals, such as chromium, lead, manganese, nickel. Quantities and characteristics of PM emissions depend on abrasive characteristics and process parameters. Emission factors are key inputs to estimate emissions. Experiments were conducted to study the effect of blast pressure, abrasive feed rate, and initial surface contamination on total PM (TPM) emission factors for coal slag. Rusted and painted mild steel surfaces were used as base plates. Blasting was carried out in an enclosed chamber, and PM was collected from an exhaust duct using U.S. Environment Protection Agency source sampling methods for stationary sources. Results showed that there is significant effect of blast pressure, feed rate, and surface contamination on TPM emissions. Mathematical equations were developed to estimate emission factors in terms of mass of emissions per unit mass of abrasive used, as well as mass of emissions per unit of surface area cleaned. These equations will help industries in estimating PM emissions based on blast pressure and abrasive feed rate. In addition, emissions can be reduced by choosing optimum operating conditions. 40 refs., 5 figs., 2 tabs.

  10. Thermal hydraulic studies in steam generator test facility

    International Nuclear Information System (INIS)

    Vinod, V.; Suresh Kumar, V.A.; Noushad, I.B.; Ellappan, T.R.; Rajan, K.K.; Rajan, M.; Vaidyanathan, G.

    2005-01-01

    Full text of publication follows: A 500 MWe fast breeder reactor is being constructed at Kalpakkam, India. This is a sodium cooled reactor with two primary and two secondary sodium loops with total 8 steam generators. The typical advantage of fast breeder plants is the high operating temperature of steam cycles and the high plant efficiency. To produce this high pressure and high temperature steam, once through straight tube vertical sodium heated steam generators are used. The steam is generated from the heat produced in the reactor core and being transported through primary and secondary sodium circuits. The steam generator is a 25 m high middle supported steam generator with expansion bend and 23 m heat transfer length. Steam Generator Test Facility (SGTF) constructed at Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam aims at performing various tests on a 5.5 MWt steam generator. This vertically simulated test article is similar in all respects to the proposed 157 MWt steam generator module for the Prototype Fast Breeder Reactor (PFBR), with reduced number of tubes. Heat transfer performance tests are done with this 19 tube steam generator at various load conditions. Sodium circuit for the SGTF is equipped with oil fired heater as heat source and centrifugal sodium pump, to pump sodium at 105 m 3 /hr flow rate. Other typical components like sodium to air heat exchanger, sodium purification system and hydrogen leak detection system is also present in the sodium circuit. High pressure steam produced in the steam generator is dumped in a condenser and recycled. Important tests planned in SGTF are the heat transfer performance test, stability test, endurance test and performance test of steam generator under various transients. The controlled operation of steam generator will be studied with possible control schemes. A steady state simulation of the steam generator is done with a mathematical model. This paper gives the details of heat transfer

  11. Flow characteristics in nuclear steam turbine blade passage

    International Nuclear Information System (INIS)

    Ahn, H.J.; Yoon, W.H.; Kwon, S.B.

    1995-01-01

    The rapid expansion of condensable gas such as moist air or steam gives rise to nonequilibrium condensation. As a result of irreversibility of condensation process in the nuclear steam turbine blade passage, the entropy of the flow increases, and the efficiency of the turbine decreases. In the present study, in order to investigate the flow characteristics of moist air in two-dimensional turbine blade passage which is made from the configuration of the last stage tip section of the actual nuclear steam turbine moving blade, the static pressures along both pressure and suction sides of blade are measured by static pressure taps and the distribution of Mach number on both sides of the blade are obtained by using the measured static pressure. Also, the flow field is visualized by a Schlieren system. From the experimental results, the effects of the stagnation temperature and specific humidity on the flow properties in the two dimensional steam turbine blade passage are clearly identified

  12. Mechanical design of a sodium heated steam generator

    International Nuclear Information System (INIS)

    Chetal, S.C.

    1975-01-01

    FBTR steam generator is a once through type unit consisting of four 12.5 MW thermal modules generating a total of 74 tons per hour of steam at 125 bar and 480 0 C. This paper outlines the mechanical design of such type of steam generator with emphasis on special design problems associated with this type of sodium to water steam heat exchanger, namely, thermal cycling of transition zone where nucleate boiling changes over to film boiling, application of pressure vessel design criteria for transient pressures, thermal stress evaluation resulting from differential expansion between shell and tube in this typical configuration, sodium headers support design, thermal sleeve, design, thermal shock analysis in thick tubes, thermal stress resulting from stratification and stability of expansion bends against vibration. Some of the possible design changes for the future large size steam generator are outlined. (author)

  13. Development of plastic media blasting device for stud bolt

    International Nuclear Information System (INIS)

    Yoshihisa, Y.; Miyashita, T.

    1999-01-01

    Plastic media blasting is a mechanical cleaning method for removing paint, rust and/or anti-galling material etc on the surface of metal without damaging the metal surface. The method is suitable for cleaning the surface of reusable elements and parts such as bolts and nuts. Anti-galling material such as molybdenum disulfide is applied to fastening stud bolts used for the steam turbine rotor casing. It is necessary to remove this material when new anti-galling material is to be applied. Genden Engineering Services and Construction Co., and Morikawa Industries Corp., have developed a plastic media blasting device to clean the surface of stud bolt screw threads installed in the facility such as lower casing of the turbine. This paper reports the outline of the results. (author)

  14. FY 1999 report on the plan to implement the model project for effective utilization of blast furnace gas pressure energy and the potential survey of the spread in India; 1999 nendo Indo ni okeru koro gas atsuryoku energy yuko riyo model jigyo jisshi keikaku oyobi fukyu kanosei chosa hokokusho

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-03-01

    As to the implementation of 'the model project for effective utilization of blast furnace gas pressure energy' and the possibility of spreading the said technology, survey/study were conducted, and the results were reported. The enterprise implementing the model project, Steel Authority of India Limited (SAIL), has been investing a total of 300 billion yen in modernizing facility/equipment after the liberalization in 1992, which is a cause of a large amount of deficit. About SAIL' cash flow, it is expected that SAIL will go into the black after tax in and after 2002. Local financial institutions are careful in investing in the iron/steel industry, but they have no problems in their own financing ability. In the evaluation of economical efficiency on the commercial basis, it is difficult to implement this model if using power prices, customs duties and borrowing rates in India as conditions. However, when preferential measures such as the exemption from taxation and help to borrowing rates are taken for energy conservation facility/equipment and the cost reduction is realized by technical cooperation, etc., it will be possible to recover the investment in 5 years. As to the possibility of spreading this model, the blast furnace with a capacity of 1000{sup 3} or more is targeted, and the number of the targeted blast furnace is 23. (NEDO)

  15. A condenser for very high power steam turbines

    International Nuclear Information System (INIS)

    Gardey, Robert.

    1973-01-01

    The invention relates to a condenser for very high power steam turbines under the masonry-block supporting the low-pressure stages of the turbine, that condenser comprises two horizontal aligned water-tube bundles passing through the steam-exhaust sleeves of the low-pressure stages, on both sides of a common inlet water box. The invention can be applied in particular to the 1000-2000 MW turbines of light water nuclear power stations [fr

  16. Sodium and steam leak simulation studies for fluidized bed steam generators

    International Nuclear Information System (INIS)

    Keeton, A.R.; Vaux, W.G.; Lee, P.K.; Witkowski, R.E.

    1976-01-01

    An experimental program is described which was conducted to study the effects of sodium or steam leaking into an operating fluidized bed of metal or ceramic particles at 680 to 800 0 K. This effort was part of the early development studies for a fluidized-bed steam generator concept using helium as the fluidizing gas. Test results indicated that steam and small sodium leaks had no effect on the quality of fluidization, heat transfer coefficient, temperature distribution, or fluidizing gas pressure drop across the bed. Large sodium leaks, however, immediately upset the operation of the fluidized bed. Both steam and sodium leaks were detected positively and rapidly at an early stage of a leak by instruments specifically selected to accomplish this

  17. Blasting agents and initiation systems

    Energy Technology Data Exchange (ETDEWEB)

    Fiscor, S.

    2000-01-01

    Although blasting differs between and within each industry, as a whole, the mines and quarries are making a shift from a purely ammonium nitrate/fuel oil (ANFO) mixture to a blend of emulsion and ANFO on a straight emulsion. Non-electric (shock tube) initiation systems have provided a viable alternative to the electric detonator (blasting cap). Explosives manufacturers are seeing their roles changes to being blasting contractors or consultants rather than just suppliers. The article discusses these trends and gives examples of typical blasting techniques and amounts of blasting agent used at large USA surface coal mines. Electric caps are still used in blasting underground coal. The Ensign Bickford Co. (EBCo) is developing electronic detonators and has been field testing an electronic initiator, the DIGIDET detonator, for the last four years. When commercially available, electronic detonators will be accurate but will come with a hefty price tag. 2 photos.

  18. Experiment on the Influence Factors of Steam Distillation Rate of Crude Oil in Porous Media

    Directory of Open Access Journals (Sweden)

    Tian Guoqing

    2017-01-01

    Full Text Available To explore the influence of complexity of reservoir properties in porous media and the diversity of operating conditions on the steam distillation rate of crude oil in the process of heavy oil exploitation with steam injection, steam distillation simulation devices are used to study steam distillation rate of crude oil in porous media. Then steam distillation ratio is obtained under the condition of different core permeability, oil saturation, steam temperatures, system pressure, steam injection rates and steam distillation rates with different viscosities of crude oil. The results show that the steam distillation rate of crude oil in porous media depends mainly on the nature of the crude oil itself, for temperature and pressure are the key factors compared with the pore structure, the initial oil saturation and steam injection rate. The experimental results help estimate the amount of crude oil and the required steam in the reservoir in the steam drive process, aiming to facilitate the optimization design and operation of steam drive.

  19. Methane Steam Reforming Kinetics for a Rhodium-Based Catalyst

    DEFF Research Database (Denmark)

    Jakobsen, Jon Geest; Jakobsen, M.; Chorkendorff, Ib

    2010-01-01

    Methane steam reforming is the key reaction to produce synthesis gas and hydrogen at the industrial scale. Here the kinetics of methane steam reforming over a rhodium-based catalyst is investigated in the temperature range 500-800 A degrees C and as a function of CH4, H2O and H-2 partial pressures....... The methane steam reforming reaction cannot be modeled without taking CO and H coverages into account. This is especially important at low temperatures and higher partial pressures of CO and H-2. For methane CO2 reforming experiments, it is also necessary to consider the repulsive interaction of CO...

  20. Experimental study of steam condensation regime map for simplified spargers

    International Nuclear Information System (INIS)

    Kim, Y. S.; Yoon, Y. J.; Song, C. H.; Park, C. K.; Kang, H. S.; Jun, H. K.

    2003-01-01

    An experimental study was conducted to produce a condensation regime map for single-hole and 4-hole steam spargers using GIRLS facility. The regime map for a single-hole sparger was derived using parameters such as the frequency and magnitude of the dynamic pressure. For 4-hole sparager, the regime map was derived using the trends of sound and dynamic pressure. Using the single-hole and 4-hole data, a steam jet condensation regime map was suggested with respect to pool temperature and steam mass flux

  1. Estimation of steam-chamber extent using 4D seismic

    Energy Technology Data Exchange (ETDEWEB)

    Tanaka, M. [Waseda Univ., Waseda (Japan); Endo, K. [Japan Canada Oil Sands Ltd., Calgary, AB (Canada); Onozuka, S. [Japan Oil, Gas and Metals National Corp., Tokyo (Japan)

    2009-07-01

    The steam-assisted gravity drainage (SAGD) technique is among the most effective steam injection methods and is widely applied in Canadian oil-sand reservoirs. The SAGD technology uses hot steam to decrease bitumen viscosity and allow it to flow. Japan Canada Oil Sands Limited (JACOS) has been developing an oil-sand reservoir in the Alberta's Hangingstone area since 1997. This paper focused on the western area of the reservoir and reported on a study that estimated the steam-chamber extent generated by horizontal well pairs. It listed steam injection start time for each well of the western area. Steam-chamber distribution was determined by distinguishing high temperature and high pore-pressure zones from low temperature and high pore-pressure zones. The bitumen recovery volume in the steam-chamber zone was estimated and compared with the actual cumulative production. This paper provided details of the methodology and interpretation procedures for the quantitative method to interpret 4D-seismic data for a SAGD process. A procedure to apply a petrophysical model was demonstrated first by scaling laboratory measurements to field-scale applications, and then by decoupling pressure and temperature effects. The first 3D seismic data in this study were already affected by higher pressures and temperatures. 11 refs., 3 tabs., 12 figs.

  2. Integrated steam generation process and system for enhanced oil recovery

    Energy Technology Data Exchange (ETDEWEB)

    Betzer-Zilevitch, M. [Ex-Tar Technologies Inc., Calgary, AB (Canada)

    2010-07-01

    A method of producing steam for the extraction of heavy bitumens was presented. The direct contact steam generation (DCSG) method is used for the direct heat transfer between combustion gas and contaminated liquid phase water to generate steam. This paper presented details of experimental and field studies conducted to demonstrate the DCSG. Results of the study demonstrated that pressure and temperature are positively correlated. As pressure increases, the flow rate of the discharged mass decreases and the steam ratio decreases. As pressure increases, the condensate and distillate flow rates increases while water vapor losses in the non-condensable gases decrease. The study indicated that for a 10 bar pressurized system producing 9.6 mt per hour of 10,000 kpa steam and 9.6 mt per hour of distillate BFW, 70 percent of the combustion energy should be recovered to generate 10,000 kpa pressure steam for EOR. Combustion energy requirements were found to decrease when pressure decreases. 11 refs., 5 tabs., 8 figs.

  3. Solution of multiple circuits of steam cycle HTR system

    International Nuclear Information System (INIS)

    Li, Fu; Wang, Dengying; Hao, Chen; Zheng, Yanhua

    2014-01-01

    In order to analyze the dynamic operation performance and safety characteristics of the steam cycle high temperature gas cooled reactor (HTR) systems, it is necessary to find the solution of the whole HTR systems with all coupled circuits, including the primary circuit, the secondary circuit, and the residual heat removal system (RHRS). Considering that those circuits have their own individual fluidity and characteristics, some existing code packages for independent circuits themselves have been developed, for example THEMRIX and TINTE code for the primary circuit of the pebble bed reactor, BLAST for once through steam generator. To solve the coupled steam cycle HTR systems, a feasible way is to develop coupling method to integrate these independent code packages. This paper presents several coupling methods, e.g. the equivalent component method between the primary circuit and steam generator which reflect the close coupling relationship, the overlapping domain decomposition method between the primary circuit and the passive RHRS which reflects the loose coupling relationship. Through this way, the whole steam cycle HTR system with multiple circuits can be easily and efficiently solved by integration of several existing code packages. Based on this methodology, a code package TINTE–BLAST–RHRS was developed. Using this code package, some operation performance of HTR–PM was analyzed, such as the start-up process of the plant, and the depressurized loss of forced cooling accident when different number of residual heat removal trains is operated

  4. Blasting agent package

    Energy Technology Data Exchange (ETDEWEB)

    Fox, R.

    1971-03-17

    A protected preassembled package for blasting agents susceptible to desensitization by water consists of, in combination: (1) an inner rigid and self-supporting tube, the upper end of which is suited to be connected, or attached, to the discharge end of a loading hose for a blasting agent and the lower end of which is open; and (2) a flexible tubular liner made of water-resistant film, having a diameter greater than that of the inner tube and a length at least equal to the desired depth of its insertion into the borehole, the liner being sleeved over the length of the inner tube, the upper end of the liner being attached to the inner tube and the lower end of the liner being closed so as to prevent substantial discharge of the explosive mixture therefrom when the latter is pumped into it. (24 claims)

  5. High-temperature oxidation of Zircaloy in hydrogen-steam mixtures

    International Nuclear Information System (INIS)

    Chung, H.M.; Thomas, G.R.

    1982-09-01

    Oxidation rates of Zircaloy-4 cladding tubes have been measured in hydrogen-steam mixtures at 1200 to 1700 0 C. For a given isothermal oxidation temperature, the oxide layer thicknesses have been measured as a function of time, steam supply rate, and hydrogen overpressure. The oxidation rates in the mixtures were compared with similar data obtained in pure steam and helium-steam environments under otherwise identical conditions. The rates in pure steam and helium-steam mixtures were equivalent and comparable to the parabolic rates obtained under steam-saturated conditions and reported in the literature. However, when the helium was replaced with hydrogen of equivalent partial pressure, a significantly smaller oxidation rate was observed. For high steam-supply rates, the oxidation kinetics in a hydrogen-steam mixture were parabolic, but the rate was smaller than for pure steam or helium-steam mixtures. Under otherwise identical conditions, the ratio of the parabolic rate for hydrogen-steam to that for pure steam decreased with increasing temperature and decreasing steam-supply rate

  6. Operational monitoring of temperature and state of stress of primary collectors, their stud bolts and cover and temperatures of steam generator's pressure vessel at the nuclear power unit WWER 440

    International Nuclear Information System (INIS)

    Matal, O.; Simo, T.; Holy, F.; Vejvoda, S.

    1992-01-01

    Both primary collectors of the WWER 440 steam generator (STGE) are vertically positioned inside the STGE pressure vessel and connected in their lower part to the primary piping and closed at their upper part by primary covers. The primary cover is pushed against the primary collector flange by 20 stud bolts. Two nickel packing rings are fitted between the primary cover and collector. A leakage in the collector-cover junction could cause flow of the radioactive water into the clean secondary water. If the junction is made in accordance with the Soviet standard design the computed stresses exceed the allowable value in the stud bolts by a factor of 1.5. Therefore an improved design of the primary collector - primary cover flange joint was designed and tested on one STGE at a WWER 440 nuclear power unit in Czechoslovakia. The paper describes the system of joint properties measurement, gives some substantial characteristics of the new stud bolts and primary cover design and comments on significant measured results of state of stress and temperatures in comparison with the operational regime of the STGE. (orig.)

  7. Tunnel blasting - recent developments

    Energy Technology Data Exchange (ETDEWEB)

    White, T.E.

    1999-05-01

    While tunnelling machines are more efficient than previously, there are still areas where blasting is a more efficient method of advance. Drilling and design methods are increasingly sophisticated, as is choice of explosive. Explosive deployment must be carefully calculated so as to avoid desensitisation. Nitroglycerine may be used as slurries; bulk mixing on site of ANFO is also practised in mining in the UK. Electric detonators, Nonel tubes, and electronic detonators are also increasingly employed.

  8. Blasting technology advances

    Energy Technology Data Exchange (ETDEWEB)

    1990-04-01

    Nitroglycerine explosives are increasingly giving way to emulsions and low cost Anfo compositions, the latter sometimes incorporating emulsion in order to improve water resistance and performance. The aim is to produce an explosive which is extremely efficient, relatively safe and can be tailored to meet user requirements. Discussion is presented of emulsion delivery systems, initiation systems, electronic detonators, computer blast modelling, directional explosives, underground automation, and the development of Soviet explosives. 3 refs., 4 figs.

  9. Steam turbines for nuclear power plants

    International Nuclear Information System (INIS)

    Stastny, M.

    1983-01-01

    A three-cylinder 220 MW saturated steam turbine was developed for WWER reactors by the Skoda concern. Twenty four of these turbines are currently in operation, in production or have been ordered. A 1000 MW four-cylinder turbine is being developed. The disign of the turbines has had to overcome difficulties connected with the unfavourable effects of wet steam at extreme power values. Great attention had to be devoted to the aerodynamics of control valves and to the prevention of flow separation areas. The problem of corrosion-erosion in guide wheels and the high pressure section was resolved by the use of ferritic stainless steels. For the low pressure section it was necessary to separate the moisture and to reheat the steam in the separator-reheater. Difficulties caused by the generation of wet steam in the low pressure section by spontaneous condensation were removed. Also limited was the erosion caused by droplets resulting from the disintegration of water films on the trailing edges. (A.K.)

  10. Computational Fluid Dynamics Application to Gun Muzzle Blast - A Validation Case Study

    National Research Council Canada - National Science Library

    Cler, Daniel

    2003-01-01

    ...) simulations of these flow-fields a viable alternative. Techniques and specialized CFD codes are being developed in order to properly model the unsteady, very high-pressure flows of gun muzzle blast...

  11. Analysis of the consequences of a steam explosion

    International Nuclear Information System (INIS)

    Hassmann, K.; Peehs, M.; Zeitner, W.; Reineke, H.

    1979-01-01

    Estimation of the maximum amount of melt that may possibly come into contact with water: determiantion of the maximum mechanical energy acting on the reactor pressure vessel; determination of the maximum load withstood by pressure vessels during steam explosions without macroscopic failure; elaboration of a statement on pressure vessel integrity by comparing the calculated maximum permissible load with the actual load. (orig.) [de

  12. Optimal Operations and Resilient Investments in Steam Networks

    Energy Technology Data Exchange (ETDEWEB)

    Bungener, Stéphane L., E-mail: stephane.bungener@a3.epfl.ch [Industrial Process and Energy Systems Engineering, École Polytechnique Fédérale de Lausanne, Lausanne (Switzerland); Van Eetvelde, Greet [Environmental and Spatial Management, Faculty of Engineering and Architecture, Ghent University, Ghent (Belgium); Maréchal, François [Industrial Process and Energy Systems Engineering, École Polytechnique Fédérale de Lausanne, Lausanne (Switzerland)

    2016-01-20

    Steam is a key energy vector for industrial sites, most commonly used for process heating and cooling, cogeneration of heat and mechanical power as a motive fluid or for stripping. Steam networks are used to carry steam from producers to consumers and between pressure levels through letdowns and steam turbines. The steam producers (boilers, heat and power cogeneration units, heat exchangers, chemical reactors) should be sized to supply the consumers at nominal operating conditions as well as peak demand. First, this paper proposes an Mixed Integer Linear Programing formulation to optimize the operations of steam networks in normal operating conditions and exceptional demand (when operating reserves fall to zero), through the introduction of load shedding. Optimization of investments based on operational and investment costs are included in the formulation. Though rare, boiler failures can have a heavy impact on steam network operations and costs, leading to undercapacity and unit shutdowns. A method is therefore proposed to simulate steam network operations when facing boiler failures. Key performance indicators are introduced to quantify the network’s resilience. The proposed methods are applied and demonstrated in an industrial case study using industrial data. The results indicate the importance of oversizing key steam producing equipments and the value of industrial symbiosis to increase industrial site resilience.

  13. Optimal Operations and Resilient Investments in Steam Networks

    International Nuclear Information System (INIS)

    Bungener, Stéphane L.; Van Eetvelde, Greet; Maréchal, François

    2016-01-01

    Steam is a key energy vector for industrial sites, most commonly used for process heating and cooling, cogeneration of heat and mechanical power as a motive fluid or for stripping. Steam networks are used to carry steam from producers to consumers and between pressure levels through letdowns and steam turbines. The steam producers (boilers, heat and power cogeneration units, heat exchangers, chemical reactors) should be sized to supply the consumers at nominal operating conditions as well as peak demand. First, this paper proposes an Mixed Integer Linear Programing formulation to optimize the operations of steam networks in normal operating conditions and exceptional demand (when operating reserves fall to zero), through the introduction of load shedding. Optimization of investments based on operational and investment costs are included in the formulation. Though rare, boiler failures can have a heavy impact on steam network operations and costs, leading to undercapacity and unit shutdowns. A method is therefore proposed to simulate steam network operations when facing boiler failures. Key performance indicators are introduced to quantify the network’s resilience. The proposed methods are applied and demonstrated in an industrial case study using industrial data. The results indicate the importance of oversizing key steam producing equipments and the value of industrial symbiosis to increase industrial site resilience.

  14. Condensers for measuring steam quality at the inlet of back-pressure units of the Los Azufres, Mich., geothermal field; Condensadores para medir la calidad del vapor a la entrada de las turbinas a contrapresion del campo geotermico de Los Azufres, Mich.

    Energy Technology Data Exchange (ETDEWEB)

    Sandoval Medina, Fernando; Gonzalez Gonzalez, Rubi; Reyes Delgado, Lisette; Medina Martinez, Moises [Comision Federal de Electricidad, Gerencia de Proyectos Geotermoelectricos, Residencia de Los Azufres (Mexico)]. E-mail: fernando.sandoval@cfe.gob.mx

    2007-01-15

    Electrical conductivity is an indirect measurement of the quality of the steam supplied to power units. In the Los Azufres, Mich., geothermal field, the electrical conductivity once was measured in a discrete and periodic way by condensing steam samples through a water-cooled condenser. In an attempt to continuously measure conductivity, conductivity meters were installed where the units discharged, but the values proved unstable and unrepresentative. Thereafter, taking into account that steam quality should be measured at the steam delivery-reception point, equipment was designed and tested for continuously condensing steam. Finally it was possible to get an air-cooled condenser able to condense 500 milliliters per minute, enough to collect a representative flow of the steam and to measure its electrical conductivity. The equipment was installed in all seven back-pressure units operating in the field and to date has been operating in an optimal manner. [Spanish] La conductividad electrica es una medida indirecta de la calidad del vapor que se suministra a las unidades turbogeneradoras. En el campo geotermico de Los Azufres, Mich., la conductividad electrica se media en forma puntual y periodica, condensando muestras de vapor por medio de un serpentin enfriado con agua. Despues, ante la necesidad de medirla en forma continua, se instalaron conductivimetros en las descargas de las unidades, pero los valores resultaron muy inestables y poco representativos. Considerando, ademas, que la calidad del vapor debe medirse en el punto de entrega-recepcion, se disenaron y probaron equipos para condensar vapor de manera continua, lograndose construir un condensador enfriado por aire que logra condensar un flujo de 500 mililitros por minuto, cantidad suficiente para tener un flujo representativo del vapor que alimenta a las turbinas y medirle su conductividad electrica. Se instalaron estos equipos en las siete unidades turbogeneradoras a contrapresion que funcionan en el campo

  15. Steam regulation for 5 MW back-pressure units when a failure occurs in the Los Humeros, Pue., field, Mexico; Regulacion del vapor en caso de falla a unidades a contrapresion de 5 MW en el campo de Los Humeros, Pue., Mexico

    Energy Technology Data Exchange (ETDEWEB)

    Rosales Lopez, Cesar [Comision Federal de Electricidad, Puebla (Mexico)]. E-mail: cesar.rosales@cfe.gob.mx

    2006-07-15

    Four out of the seven back-pressure power units of 5 MW operating in the Los Humeros geothermal field, State of Puebla, Mexico, are fed by one steam pipe gathering the steam produced by nine wells. When a failure occurred in any of the units and the excedence valve had to be open to deviate the steam, a decrease in the steam flow for the remaining units was noted, along with lower electrical generation. The cause for that is analyzed and explained in this paper by comparing the interconnected steam supply system to an electric circuit. A way to maintain a uniform and continuous supply of steam in the Los Humeros field has been found. It was implemented several months ago and the problem has not reoccurred. [Spanish] Cuatro de las siete unidades de 5 MW a contrapresion que operan en el campo geotermico de Los Humeros, Puebla, son alimentadas por un solo vaporducto que reune el vapor de nueve pozos productores. Cuando ocurria una falla en alguna de estas unidades y se abria por completo la valvula de excedencia para desviar el vapor, se observaba una reduccion en el flujo de vapor que llegaba a las otras tres unidades, lo que a su vez ocasionaba que la generacion de electricidad se redujera notoriamente. En este trabajo se analiza y explica la causa de ello, mediante la comparacion de este sistema interconectado de suministro de vapor con un circuito electrico, y se explica la solucion que se encontro e implemento en el campo de Los Humeros para regular el suministro continuo y uniforme de vapor, con resultados satisfactorios a varios meses de su implementacion en las cuatro unidades interconectadas.

  16. Steam jet mill-a prospective solution to industrial exhaust steam and solid waste.

    Science.gov (United States)

    Zhang, Mingxing; Chen, Haiyan

    2018-04-20

    Bulk industrial solid wastes occupy a lot of our resources and release large amounts of toxic and hazardous substances to the surrounding environment, demanding innovative strategies for grinding, classification, collection, and recycling for economically ultrafine powder. A new technology for grinding, classification, collection, and recycling solid waste is proposed, using the superheated steam produced from the industrial exhaust steam to disperse, grind, classify, and collect the industrial solid waste. A large-scale steam jet mill was designed to operate at an inlet steam temperature 230-300 °C and an inlet pressure of 0.2-0.6 MPa. A kind of industrial solid waste fluidized-bed combustion ashes was used to grinding tests at different steam temperatures and inlet pressures. The total process for grinding, classification, and collection is drying. Two kinds of particle sizes are obtained. One particle size is d 50  = 4.785 μm, and another particle size is d 50  = 8.999 μm. For particle size d 50  = 8.999 μm, the inlet temperature is 296 °C and an inlet pressure is 0.54 MPa for the grinding chamber. The steam flow is 21.7 t/h. The yield of superfine powder is 73 t/h. The power consumption is 3.76 kW h/t. The obtained superfine powder meets the national standard S95 slag. On the basis of these results, a reproducible and sustainable industrial ecological protocol using steam produced by industrial exhaust heat coupled to solid waste recycling is proposed, providing an efficient, large-scale, low-cost, promising, and green method for both solid waste recovery and industrial exhaust heat reutilization.

  17. Deliberate ignition of hydrogen-air-steam mixtures in condensing steam environments

    International Nuclear Information System (INIS)

    Blanchat, T.K.; Stamps, D.W.

    1997-05-01

    Large scale experiments were performed to determine the effectiveness of thermal glow plug igniters to burn hydrogen in a condensing steam environment due to the presence of water sprays. The experiments were designed to determine if a detonation or accelerated flame could occur in a hydrogen-air-steam mixture which was initially nonflammable due to steam dilution but was rendered flammable by rapid steam condensation due to water sprays. Eleven Hydrogen Igniter Tests were conducted in the test vessel. The vessel was instrumented with pressure transducers, thermocouple rakes, gas grab sample bottles, hydrogen microsensors, and cameras. The vessel contained two prototypic engineered systems: (1) a deliberate hydrogen ignition system and (2) a water spray system. Experiments were conducted under conditions scaled to be nearly prototypic of those expected in Advanced Light Water Reactors (such as the Combustion Engineering (CE) System 80+), with prototypic spray drop diameter, spray mass flux, steam condensation rates, hydrogen injection flow rates, and using the actual proposed plant igniters. The lack of any significant pressure increase during the majority of the burn and condensation events signified that localized, benign hydrogen deflagration(s) occurred with no significant pressure load on the containment vessel. Igniter location did not appear to be a factor in the open geometry. Initially stratified tests with a stoichiometric mixture in the top showed that the water spray effectively mixes the initially stratified atmosphere prior to the deflagration event. All tests demonstrated that thermal glow plugs ignite hydrogen-air-steam mixtures under conditions with water sprays near the flammability limits previously determined for hydrogen-air-steam mixtures under quiescent conditions. This report describes these experiments, gives experimental results, and provides interpretation of the results. 12 refs., 127 figs., 16 tabs

  18. Measurement of mass flux in high temperature high pressure steam-water two-phase flow using a combination of Pitot tubes and a gamma densitometer

    International Nuclear Information System (INIS)

    Chan, A.M.C.; Bzovey, D.

    1990-01-01

    The design and calibration of a two-phase mass-flux measurement device making use of a Pitot-tube rake and a gamma densitometer are described. Five Pitot tubes and three chordal void-fraction measurements are used. Similar devices have been reported previously. The present device is designed for easy operation and simple data interpretation for both axisymmetric and non-axisymmetric flows under high pressure transient two-phase flow conditions. The device was calibrated using a vertical two-phase flow loop as well as a model-scale pump loop in horizontal orientation. Good agreement between the measured two-phase mass fluxes and the single-phase values was obtained in both cases. (orig.)

  19. IAEA activities on steam generator life management

    International Nuclear Information System (INIS)

    Gueorguiev, B.; Lyssakov, V.; Trampus, P.

    2002-01-01

    The International Atomic Energy Agency (IAEA) carries out a set of activities in the field of Nuclear Power Plant (NPP) life management. Main activities within this area are implemented through the Technical Working Group on Life Management of NPPs, and mostly concentrated on studies of understanding mechanisms of degradation and their monitoring, optimisation of maintenance management, economic aspects, proven practices of and approaches to plant life management including decommissioning. The paper covers two ongoing activities related to steam generator life management: the International Database on NPP Steam Generators and the Co-ordinated Research Project on Verification of WWER Steam Generator Tube Integrity (WWER is the Russian designed PWR). The lifetime assessment of main components relies on an ability to assess their condition and predict future degradation trends, which to a large extent is dependent on the availability of relevant data. Effective management of ageing and degradation processes requires a large amount of data. Several years ago the IAEA started to work on the International Database on NPP Life Management. This is a multi-module database consisting of modules such as reactor pressure vessels materials, piping, steam generators, and concrete structures. At present the work on pressure vessel materials, on piping as well as on steam generator is completed. The paper will present the concept and structure of the steam generator module of the database. In countries operating WWER NPPs, there are big differences in the eddy current inspection strategy and practice as well as in the approach to steam generator heat exchanger tube structural integrity assessment. Responding to the need for a co-ordinated research to compare eddy current testing results with destructive testing using pulled out tubes from WWER steam generators, the IAEA launched this project. The main objectives of the project are to summarise the operating experiences of WWER

  20. French steam generator

    International Nuclear Information System (INIS)

    Remond, A.

    1986-01-01

    After recalling the potential damage mode of tubes of steam generator, the author recalls the safety criteria used in France. The improvements and the process of damage prejudice and reparation for tubular bundle are presented [fr

  1. Liquid metal steam generator

    International Nuclear Information System (INIS)

    Wolowodiuk, W.

    1975-01-01

    A liquid metal heated steam generator is described which in the event of a tube failure quickly exhausts out of the steam generator the products of the reaction between the water and the liquid metal. The steam is generated in a plurality of bayonet tubes which are heated by liquid metal flowing over them between an inner cylinder and an outer cylinder. The inner cylinder extends above the level of liquid metal but below the main tube sheet. A central pipe extends down into the inner cylinder with a centrifugal separator between it and the inner cylinder at its lower end and an involute deflector plate above the separator so that the products of a reaction between the liquid metal and the water will be deflected downwardly by the deflector plate and through the separator so that the liquid metal will flow outwardly and away from the central pipe through which the steam and gaseous reaction products are exhausted. (U.S.)

  2. Steam cleaning device

    International Nuclear Information System (INIS)

    Karaki, Mikio; Muraoka, Shoichi.

    1985-01-01

    Purpose: To clean complicated and long objects to be cleaned having a structure like that of nuclear reactor fuel assembly. Constitution: Steams are blown from the bottom of a fuel assembly and soon condensated initially at the bottom of a vertical water tank due to water filled therein. Then, since water in the tank is warmed nearly to the saturation temperature, purified water is supplied from a injection device below to the injection device above the water tank on every device. In this way, since purified water is sprayed successively from below to above and steams are condensated in each of the places, the entire fuel assembly elongated in the vertical direction can be cleaned completely. Water in the reservoir goes upward like the steam flow and is drained together with the eliminated contaminations through an overflow pipe. After the cleaning has been completed, a main steam valve is closed and the drain valve is opened to drain water. (Kawakami, Y.)

  3. Steam generator water lancing

    International Nuclear Information System (INIS)

    Kamler, F.; Schneider, W.

    1992-01-01

    The tubesheet and tube support plate deposits in CANDU steam generators are notable for their hardness. Also notable is the wide variety of steam generator access situations. Because of the sludge hardness and the difficulty of the access, traditional water lancing processes which directed jets from the central tube free lane or from the periphery of the bundle have proven unsuitable. This has led to the need for some very unique inter tube water lancing devices which could direct powerful water jets directly onto the deposits. This type of process was applied to the upper broached plates of the Bruce A steam generators, which had become severely blocked. It has since been applied to various other steam generator situations. This paper describes the flexlance equipment development, qualification, and performance in the various CANDU applications. 4 refs., 2 tabs., 7 figs

  4. Steam explosions in sodium cooled breeder reactors

    International Nuclear Information System (INIS)

    Lundell, B.

    1982-01-01

    Steam explosion is considered a physical process which transport heat from molten fuel to liquid coolant so fast that the coolant starts boiling in an explosion-like manner. The arising pressure waves transform part of the thermal energy to mechanical energy. This can stress the reactor tank and threaten its hightness. The course of the explosion has not been theoretical explained. Experimental results indicate that the probability of steam explosions in a breeder reactor is small. The efficiency of the transformation of the heat of fusion into mechanical energy in substantially lower than the theoretical maximum value. The mechanical stress from the steam explosion on the reactor tank does not seem to jeopardize its tightness. (G.B.)

  5. Chemical cleaning of nuclear (PWR) steam generators

    International Nuclear Information System (INIS)

    Welty, C.S. Jr.; Mundis, J.A.

    1982-01-01

    This paper reports on a significant research program sponsored by a group of utilities (the Steam Generator Owners Group), which was undertaken to develop a process to chemically remove corrosion product deposits from the secondary side of pressurized water reactor (PWR) power plant steam generators. Results of this work have defined a process (solvent system and application methods) that is capable of removing sludge and tube-to-tube support plate crevice corrosion products generated during operation with all-volatile treatment (AVT) water chemistry. Considers a plant-specific test program that includes all materials in the steam generator to be cleaned and accounts for the physical locations (proximity and contact) of those materials. Points out that prior to applying the process in an operational unit, the utility, with the participation of the NSSR vendor, must define allowable total corrosion to the materials of construction of the unit

  6. New ferritic steels for advanced steam plants

    Energy Technology Data Exchange (ETDEWEB)

    Mayer, K.H; Koenig, H. [GEC ALSTHOM Energie GmbH, Nuremberg (Germany)

    1998-12-31

    During the last 15-20 years ferritic-martensitic 9-12 % chromium steels have been developed under international research programmes which permit inlet steam temperatures up to approx. 625 deg C and pressures up to about 300 bars, thus leading to improvements in thermal efficiency of around 8 % and a CO{sub 2} reduction of about 20 % versus conventional steam parameters. These new steels are already being applied in 13 European and 34 Japanese power stations with inlet steam temperature up to 610 deg C. This presentation will give an account of the content, scope and results of the research programmes and of the experience gained during the production of components which have been manufactured from the new steels. (orig.) 13 refs.

  7. New ferritic steels for advanced steam plants

    Energy Technology Data Exchange (ETDEWEB)

    Mayer, K H; Koenig, H [GEC ALSTHOM Energie GmbH, Nuremberg (Germany)

    1999-12-31

    During the last 15-20 years ferritic-martensitic 9-12 % chromium steels have been developed under international research programmes which permit inlet steam temperatures up to approx. 625 deg C and pressures up to about 300 bars, thus leading to improvements in thermal efficiency of around 8 % and a CO{sub 2} reduction of about 20 % versus conventional steam parameters. These new steels are already being applied in 13 European and 34 Japanese power stations with inlet steam temperature up to 610 deg C. This presentation will give an account of the content, scope and results of the research programmes and of the experience gained during the production of components which have been manufactured from the new steels. (orig.) 13 refs.

  8. Blast Testing Issues and TBI; Experimental Models that Lead to Wrong Conclusions

    Directory of Open Access Journals (Sweden)

    Charles E. Needham

    2015-04-01

    Full Text Available Over the past several years we have noticed an increase in the number of blast injury studies published in peer-reviewed biomedical journals that have utilized improperly conceived experiments. Data from these studies will lead to false conclusions and more confusion than advancement in the understanding of blast injury, particularly blast neurotrauma. Computational methods to properly characterize the blast environment have been available for decades. These methods, combined with a basic understanding of blast wave phenomena enable researchers to extract useful information from well documented experiments. This basic understanding must include the differences and interrelationships of static pressure, dynamic pressure, reflected pressure, and total or stagnation pressure in transient shockwave flows, how they relate to loading of objects, and how they are properly measured. However, it is critical that the research community effectively overcomes the confusion that has been compounded by a misunderstanding of the differences between the loading produced by a free field explosive blast and loading produced by a conventional shock tube. The principles of blast scaling have been well established for decades and when properly applied will do much to repair these problems.This paper provides guidance regarding proper experimental methods and offers insights into the implications of improperly designed and executed tests. Through application of computational methods, useful data can be extracted from well documented historical tests, and future work can be conducted in a way to maximize the effectiveness and use of valuable biological test data.

  9. Steam-water separator

    International Nuclear Information System (INIS)

    Modrak, T.M.; Curtis, R.W.

    1978-01-01

    The steam-water separator connected downstream of a steam generator consists of a vertical centrifugal separator with swirl blades between two concentric pipes and a cyclone separator located above. The water separated in the cyclone separator is collected in the inner tube of the centrifugal separator which is closed at the bottom. This design allows the overall height of the separator to be reduced. (DG) [de

  10. Bomb blast mass casualty incidents: initial triage and management of injuries.

    Science.gov (United States)

    Goh, S H

    2009-01-01

    Bomb blast injuries are no longer confined to battlefields. With the ever present threat of terrorism, we should always be prepared for bomb blasts. Bomb blast injuries tend to affect air-containing organs more, as the blast wave tends to exert a shearing force on air-tissue interfaces. Commonly-injured organs include the tympanic membranes, the sinuses, the lungs and the bowel. Of these, blast lung injury is the most challenging to treat. The clinical picture is a mix of acute respiratory distress syndrome and air embolism, and the institution of positive pressure ventilation in the presence of low venous pressures could cause systemic arterial air embolism. The presence of a tympanic membrane perforation is not a reliable indicator of the presence of a blast injury in the other air-containing organs elsewhere. Radiological imaging of the head, chest and abdomen help with the early identification of blast lung injury, head injury, abdominal injury, eye and sinus injuries, as well as any penetration by foreign bodies. In addition, it must be borne in mind that bomb blasts could also be used to disperse radiological and chemical agents.

  11. International Symposium on Military Applications of Blast Simulation (5th)

    Science.gov (United States)

    1977-06-22

    Centre d’Etudes de Grainat, Gramat , France, de- scribed their work in designing a blast simulator with a test section of 12—rn width and 7—rn height...de Gramat , Gramat , France) on the dynamic behavior of limestone. Through a series of high—pressure experiments of the type developed in the recent

  12. 1000 MW steam turbine for Temelin nuclear power station

    International Nuclear Information System (INIS)

    Drahy, J.

    1992-01-01

    Before the end 1991 the delivery was completed of the main parts (3 low-pressure sections and 1 high-pressure section, all of double-flow design) of the first full-speed (3000 r.p.m.) 1000 MW steam turbine for saturated admission steam for the Temelin nuclear power plant. Description of the turbine design and of new technologies and tools used in the manufacture are given. Basic technical parameters of the steam turbine are as follows: maximum output of steam generators 6060 th -1 ; maximum steam flow into turbine 5494.7 th -1 ; output of turbo-set 1024 MW; steam conditions before the turbine inlet: pressure 5.8 MPa, temperature 273.3 degC, steam wetness 0.5%; nominal temperature of cooling water 21 degC; temperature of feed water 220.8 degC; maximum consumption of heat from turbine for heating at 3-stage heating of heating water 60/150 degC. (Z.S.) 7 figs., 2 refs

  13. Review of steam jet condensation in a water pool

    International Nuclear Information System (INIS)

    Kim, Y. S.; Song, C. H.; Park, C. K.; Kang, H. S.; Jeon, H. G.; Yoon, Y. J.

    2002-01-01

    In the advanced nuclear power plants including APR1400, the SDVS is adopted to increase the plant safety using the concept of feed-and-bleed operation. In the case of the TLOFW, the POSRV located at the top of the pressurizer is expected to open due to the pressurization of the reactor coolant system and discharges steam and/or water mixture into the water pool, where the mixture is condensed. During the condensation of the mixture, thermal-hydraulic loads such as pressure and temperature variations are induced to the pool structure. For the pool structure design, such thermal-hydraulic aspects should be considered. Understanding the phenomena of the submerged steam jet condensation in a water pool is helpful for system designers to design proper pool structure, sparger, and supports etc. This paper reviews and evaluates the steam jet condensation in a water pool on the physical phenomena of the steam condensation including condensation regime map, heat transfer coefficient, steam plume, steam jet condensation load, and steam jet induced flow

  14. Validation results of the pre-service ultrasonic inspections of the Sizewell B pressurizer and steam generators and reactor coolant pump flywheels

    International Nuclear Information System (INIS)

    Conroy, P.J.; Leyland, K.S.

    1995-01-01

    In the UK, concern over the safety issues associated with nuclear power generation resulted in a demand for a public inquiry into the construction and operation of Sizewell ''B'', Britain's first PWR. This public inquiry was additional to the UK's normal licensing process. The onus was placed upon the UK utility, CEGB (now Nuclear Electric plc) to provide evidence to the inquiry to support the case that the plant would be constructed and operated to a sufficiently high standard of safety. Part of the evidence to the inquiry (1) relied upon the ability of ultrasonic inspections to verify that the reactor pressure vessel and other safety critical components (collectively known as IoF components), were free from defects that could threaten structural integrity. At that time, the body of evidence showed that although ultrasonic inspection had the potential to satisfy this requirement, it would be necessary to validate the procedures and key operators used in order to provide assurance that they were adequate. Inspection validation therefore became an integral part of the UK PWR nuclear power program

  15. Development of technologies on innovative-simplified nuclear power plant using high-efficiency steam injectors (5) operating characteristics of center water jet type supersonic steam injector

    International Nuclear Information System (INIS)

    Abe, Y.; Kawamoto, Y.; Iwaki, C.; Narabayashi, T.; Mori, M.; Ohmori, S.

    2005-01-01

    Next-generation reactor systems have been under development aiming at simplified system and improvement of safety and credibility. A steam injector has a function of a passive pump without large motor or turbo-machinery, and has been investigated as one of the most important component of the next-generation reactor. Its performance as a pump depends on direct contact condensation phenomena between a supersonic steam and a sub-cooled water jet. As previous studies of the steam injector, there are studies about formulation of operating characteristic of steam injector and analysis of jet structure in steam injector by Narabayashi etc. And as previous studies of the direct contact condensation, there is the study about the direct contact condensation in steam atmosphere. However the study about the turbulent heat transfer under the great shear stress is not enough investigated. Therefore it is necessary to examine in detail about the operating characteristic of the steam injector. The present paper reports the observation results of the water jet behavior in the super sonic steam injector by using the video camera and the high-speed video camera. And the measuring results of the temperature and the pressure distribution in the steam injector are reported. From observation results by video camera, it is cleared that the water jet is established at the center of the steam injector right after steam supplied and the operation of the steam injector depends on the throat diameter. And from observation results by high-speed video camera, it is supposed that the columned water jet surface is established in the mixing nozzle and the water jet surface movement exists. And from temperature measuring results, it is supposed that the steam temperature at the mixing nozzle is changed between about 80 degree centigrade and about 60 degree centigrade. Then from the pressure measuring results, it is confirmed that the pressure at the diffuser depends on each the throat diameter and

  16. Highly purified water production technology. The influence of water purity on steam quality

    International Nuclear Information System (INIS)

    Ganter, J.

    1975-01-01

    The fundamental question related to high-pressure steam generation, intended for powering steam turbines, concerns steam production conditions based on constant quality standards. The characteristics of water (salinity, silica concentration) are indicated for a given steam quality as a function of the pressure. Two processes for the purification of feedwater for high pressure boilers are described: a treatment using precoated cellulose or resin filters and a treatment using mixed-bed ion exchangers. When ultrapure water is required, the demineralized water is filtred using microfiltration and ultrafiltration processes [fr

  17. Workshop proceedings: U-bend tube cracking in steam generators

    Science.gov (United States)

    Shoemaker, C. E.

    1981-06-01

    A design to reduce the rate of tube failure in high pressure feedwater heaters, a number of failed drawn and stress relieved Monel 400 U-bend tubes removed from three high pressure feedwater heaters was examined. Steam extracted from the turbine is used to preheat the boiler feedwater in fossil fuel fired steam plants to improve thermal efficiency. This is accomplished in a series of heaters between the condenser hot well and the boiler. The heaters closest to the boiler handle water at high pressure and temperature. Because of the severe service conditions, high pressure feedwater heaters are frequently tubed with drawn and stress relieved Monel 400.

  18. Numerical simulation of muzzle blast

    NARCIS (Netherlands)

    Tyler-Street, M.

    2014-01-01

    Structural design methods for naval ships include environmental, operational and military load cases. One of the operational loads acting on a typical naval vessel is the muzzle blast from a gun. Simulating the muzzle blast load acting on a ship structure with CFD and ALE methods leads to large

  19. Performance analysis of a potassium-steam two stage vapour cycle

    International Nuclear Information System (INIS)

    Mitachi, Kohshi; Saito, Takeshi

    1983-01-01

    It is an important subject to raise the thermal efficiency in thermal power plants. In present thermal power plants which use steam cycle, the plant thermal efficiency has already reached 41 to 42 %, steam temperature being 839 K, and steam pressure being 24.2 MPa. That is, the thermal efficiency in a steam cycle is facing a limit. In this study, analysis was made on the performance of metal vapour/steam two-stage Rankine cycle obtained by combining a metal vapour cycle with a present steam cycle. Three different combinations using high temperature potassium regenerative cycle and low temperature steam regenerative cycle, potassium regenerative cycle and steam reheat and regenerative cycle, and potassium bleed cycle and steam reheat and regenerative cycle were systematically analyzed for the overall thermal efficiency, the output ratio and the flow rate ratio, when the inlet temperature of a potassium turbine, the temperature of a potassium condenser, and others were varied. Though the overall thermal efficiency was improved by lowering the condensing temperature of potassium vapour, it is limited by the construction because the specific volume of potassium in low pressure section increases greatly. In the combinatipn of potassium vapour regenerative cycle with steam regenerative cycle, the overall thermal efficiency can be 58.5 %, and also 60.2 % if steam reheat and regenerative cycle is employed. If a cycle to heat steam with the bled vapor out of a potassium vapour cycle is adopted, the overall thermal efficiency of 63.3 % is expected. (Wakatsuki, Y.)

  20. Numerical study on over reading coefficient in wet steam flow measurement

    International Nuclear Information System (INIS)

    Bai Xuesong; Yuan Dewen; Yan Xiao; Peng Xingjian

    2013-01-01

    This paper investigated the flow process of wet steam in Venturi under interested conditions with CFD simulation software. The effect of pressure, mass flow rate, throat radius on over reading factor was analyzed. This paper aims to improve the wet steam over reading model and the prediction accuracy in wet steam. The results prove that the mass flow has a small effect on over reading coefficient, while the effect that throat radius has on over reading coefficient increases as the pressure rises. (authors)

  1. Check of condition of steam generators, volume compensators and turbine condensers in nuclear power plants

    International Nuclear Information System (INIS)

    Matal, O.; Klinga, J.; Holy, F.; Sobotka, J.

    1989-01-01

    A negative pressure leak detector is described designed for leak testing of tubes in steam generators and steam turbine condensers. The principle, operation and use are described of inflatable bags and an inflatable platform. The bags are designed for insulating and sealing spaces in nuclear reactor components while the inflatable platform is used in pressurizer inspections and repairs. Their properties, and other facilities for detecting leaks in steam generator tubes are briefly described. (M.D.). 3 figs

  2. Dynamic Modeling of Steam Condenser and Design of PI Controller Based on Grey Wolf Optimizer

    OpenAIRE

    Shu-Xia Li; Jie-Sheng Wang

    2015-01-01

    Shell-and-tube condenser is a heat exchanger for cooling steam with high temperature and pressure, which is one of the main kinds of heat exchange equipment in thermal, nuclear, and marine power plant. Based on the lumped parameter modeling method, the dynamic mathematical model of the simplified steam condenser is established. Then, the pressure PI control system of steam condenser based on the Matlab/Simulink simulation platform is designed. In order to obtain better performance, a new meta...

  3. Effects of low-level blast exposure on the nervous system: Is there really a controversy?

    Directory of Open Access Journals (Sweden)

    Gregory A Elder

    2014-12-01

    Full Text Available High-pressure blast waves can cause extensive CNS injury in humans. However, in combat settings such as Iraq and Afghanistan, lower level exposures associated with mild TBI (mTBI or subclinical exposure have been much more common. Yet controversy exists concerning what traits can be attributed to low-level blast, in large part due to the difficulty of distinguishing blast-related mTBI from post-traumatic stress disorder (PTSD. We describe how TBI is defined in humans and the problems posed in using current definitions to recognize blast-related mTBI. We next consider the problem of applying definitions of human mTBI to animal models, in particular that TBI severity in humans is defined in relation to alteration of consciousness at the time of injury, which typically cannot be assessed in animals. However, based on outcome assessments a condition of low-level blast exposure can be defined in animals that likely approximates human mTBI or subclinical exposure. We review blast injury modeling in animals noting that inconsistencies in experimental approach have contributed to uncertainty over the effects of low-level blast. Yet animal studies show that low-level blast pressure waves are transmitted to the brain. In brain low-level blast exposures cause behavioral, biochemical, pathological and physiological effects on the nervous system including the induction of PTSD-related behavioral traits in the absence of a psychological stressor. We review the relationship of blast exposure to chronic neurodegenerative diseases noting the paradoxical lowering of Abeta by blast, which along with other observations suggest that blast-related TBI is pathophysiologically distinct from non-blast TBI. Human neuroimaging studies show that blast-related mTBI is associated with a variety of chronic effects that are unlikely to be explained by co-morbid PTSD. We conclude that abundant evidence supports low-level blast as having long-term effects on the nervous system.

  4. Control blasting of reinforced concrete

    International Nuclear Information System (INIS)

    Nagase, Tetsuo

    1981-01-01

    With the need of decommissioning nuclear power plants, it is urgently required to establish its methods and standards. In Shimizu Construction Co., Ltd., experimental feasibility studies have been made on explosive demolition method i.e. the controlled blasting for the massive concrete structures peculiar to nuclear power plants, considering low radiation exposure, safety and high efficiency. As such, four techniques of line drilling, cushion blasting, pre-splitting and guide hole blasting, respectively, are described with photographs. Assuming the selective demolition of activated concrete structures, the series of experiments showed the good results of clear-cut surfaces and the effect of blasting was confined properly. Moreover, the scattering of debris in blasting was able to be entirely prevented by the use of rubber belts. The generation of gas and dust was also little due to the small amount of the charge used. (J.P.N.)

  5. NCBI BLAST+ integrated into Galaxy.

    Science.gov (United States)

    Cock, Peter J A; Chilton, John M; Grüning, Björn; Johnson, James E; Soranzo, Nicola

    2015-01-01

    The NCBI BLAST suite has become ubiquitous in modern molecular biology and is used for small tasks such as checking capillary sequencing results of single PCR products, genome annotation or even larger scale pan-genome analyses. For early adopters of the Galaxy web-based biomedical data analysis platform, integrating BLAST into Galaxy was a natural step for sequence comparison workflows. The command line NCBI BLAST+ tool suite was wrapped for use within Galaxy. Appropriate datatypes were defined as needed. The integration of the BLAST+ tool suite into Galaxy has the goal of making common BLAST tasks easy and advanced tasks possible. This project is an informal international collaborative effort, and is deployed and used on Galaxy servers worldwide. Several examples of applications are described here.

  6. Computer codes for simulation of Angra 1 reactor steam generator

    International Nuclear Information System (INIS)

    Pinto, A.C.

    1978-01-01

    A digital computer code is developed for the simulation of the steady-state operation of a u-tube steam generator with natural recirculation used in Pressurized Water Reactors. The steam generator is simulated with two flow channel separated by a metallic wall, with a preheating section with counter flow and a vaporizing section with parallel flow. The program permits the changes in flow patterns and heat transfer correlations, in accordance with the local conditions along the vaporizing section. Various sub-routines are developed for the determination of steam and water properties and a mathematical model is established for the simulation of transients in the same steam generator. The steady state operating conditions in one of the steam generators of ANGRA 1 reactor are determined utilizing this programme. Global results obtained agree with published values [pt

  7. Targeted steam injection using horizontal wells with limited entry perforations

    Energy Technology Data Exchange (ETDEWEB)

    Boone, T. J.; Youck, D. G.; Sun, S. [Imperial Oil Resources, Calgary, AB (Canada)

    1998-12-31

    An experimental horizontal well using limited-entry perforations as a method for distributing steam to different zones was used to replace ten vertical injection wells. The well was located between rows of vertical wells in a reservoir that has been subjected to more than ten years of operation under cyclic steam stimulation. The limited-entry perforations enabled steam to be targeted at the cold regions of the reservoir. This paper presents an assessment of the well based on theoretical calculations, measured injection pressures and rates and 3-D seismic imaging. All the data collected during the experiment support the conclusion that effective steam distribution along the well has been achieved. It was also concluded that this technology has significant potential for SAGD applications as a mechanism for achieving improved steam distribution at a much reduced cost. 5 refs., 8 figs.

  8. Optimizing cast blasting efficiency using ANFO with liners

    Energy Technology Data Exchange (ETDEWEB)

    Madsen, A.

    2007-01-15

    As part of a five research project funded by the National Science Foundation, Peabody Energy studied three experimental cast blasts conducted at the North Antelope Rochelle mine site on July 24,28 and 31 2005. The initial purpose of this research project was to determine the influence that blast initiation sequence have on: NOx production; Face Displacement; Highwall damage; Explosive performance; Vibration emissions; Displacement; Surface swell; and Cast benefit. Two new discoveries on velocity of detonation (VoD) and pressure of detonation (PoD) were made as a result of this research project. Furthermore, a relationship between surface swell velocity and face velocity was also noted. 7 figs., 3 tabs.

  9. Simulation of a main steam line break with steam generator tube rupture using trace

    Energy Technology Data Exchange (ETDEWEB)

    Gallardo, S.; Querol, A.; Verdu, G. [Departamento de Ingenieria Quimica Y Nuclear, Universitat Politecnica de Valencia, Camino de Vera s/n, 46022, Valencia (Spain)

    2012-07-01

    A simulation of the OECD/NEA ROSA-2 Project Test 5 was made with the thermal-hydraulic code TRACE5. Test 5 performed in the Large Scale Test Facility (LSTF) reproduced a Main Steam Line Break (MSLB) with a Steam Generator Tube Rupture (SGTR) in a Pressurized Water Reactor (PWR). The result of these simultaneous breaks is a depressurization in the secondary and primary system in loop B because both systems are connected through the SGTR. Good approximation was obtained between TRACE5 results and experimental data. TRACE5 reproduces qualitatively the phenomena that occur in this transient: primary pressure falls after the break, stagnation of the pressure after the opening of the relief valve of the intact steam generator, the pressure falls after the two openings of the PORV and the recovery of the liquid level in the pressurizer after each closure of the PORV. Furthermore, a sensitivity analysis has been performed to know the effect of varying the High Pressure Injection (HPI) flow rate in both loops on the system pressures evolution. (authors)

  10. Blast from the past

    Science.gov (United States)

    Carlowicz, Michael

    1996-02-01

    Forget dynamite or hydraulic and mechanical drills. Industrial and federal researchers have started boring holes with rocket fuel. In a cooperative arrangement between Sandia National Laboratory, Global Environmental Solutions, and Universal Tech Corp., scientists and engineers extracted fuel from 200 rocket motors and used it as a mining explosive. In a demonstration completed last fall, researchers used 4950 kg of solid rocket propellant to move more than 22,500 metric tons of rock from the Lone Star Quarry in Prairie, Oklahoma. They found that the fuel improved blast energy and detonation velocity over traditional explosives, and it required fewer drill holes.

  11. Trouble shooting problem blasts

    Energy Technology Data Exchange (ETDEWEB)

    Konya, K.D. [Society of Explosive Engineers, Montville, OH (United States)

    1996-02-01

    There have been many occurrences of water-gel or emulsion explosives or heavy Anfo`s (ammonium nitrate and fuel oil) either not detonating in some blast holes in surface coal mining operations, or producing yellow or reddish-brown fumes upon reaction. The explosives rely on entrapped gas to detonate properly. It is important to control the size of the bubble and the dispersion of gas bubbles throughout the explosive. The article discusses and compares the use of microspheres and of gassing agents for detonation control.

  12. Dismounted Complex Blast Injury.

    Science.gov (United States)

    Andersen, Romney C; Fleming, Mark; Forsberg, Jonathan A; Gordon, Wade T; Nanos, George P; Charlton, Michael T; Ficke, James R

    2012-01-01

    The severe Dismounted Complex Blast Injury (DCBI) is characterized by high-energy injuries to the bilateral lower extremities (usually proximal transfemoral amputations) and/or upper extremity (usually involving the non-dominant side), in addition to open pelvic injuries, genitourinary, and abdominal trauma. Initial resuscitation and multidisciplinary surgical management appear to be the keys to survival. Definitive treatment follows general principals of open wound management and includes decontamination through aggressive and frequent debridement, hemorrhage control, viable tissue preservation, and appropriate timing of wound closure. These devastating injuries are associated with paradoxically favorable survival rates, but associated injuries and higher amputation levels lead to more difficult reconstructive challenges.

  13. Steam explosion studies review

    International Nuclear Information System (INIS)

    Hwang, Moon Kyu; Kim, Hee Dong

    1999-03-01

    When a cold liquid is brought into contact with a molten material with a temperature significantly higher than the liquid boiling point, an explosive interaction due to sudden fragmentation of the melt and rapid evaporation of the liquid may take place. This phenomenon is referred to as a steam explosion or vapor explosion. Depending upon the amount of the melt and the liquid involved, the mechanical energy released during a vapor explosion can be large enough to cause serious destruction. In hypothetical severe accidents which involve fuel melt down, subsequent interactions between the molten fuel and coolant may cause steam explosion. This process has been studied by many investigators in an effort to assess the likelihood of containment failure which leads to large scale release of radioactive materials to the environment. In an effort to understand the phenomenology of steam explosion, extensive studies has been performed so far. The report presents both experimental and analytical studies on steam explosion. As for the experimental studies, both small scale tests which involve usually less than 20 g of high temperature melt and medium/large scale tests which more than 1 kg of melt is used are reviewed. For the modelling part of steam explosions, mechanistic modelling as well as thermodynamic modelling is reviewed. (author)

  14. New low pressure exhaust modules for the MAN steam turbine product line. High performance bladings for highest efficiency levels; Neue Niederdruck-Module fuer die MAN-Dampfturbinenproduktlinie. Hochentwickelte Beschaufelungen fuer hoechste Leistungsdichten und Wirkungsgrade

    Energy Technology Data Exchange (ETDEWEB)

    Schwarz, M.A.; Behnke, K.; Klemm, H. [MAN TURBO AG, Oberhausen (Germany)

    2008-07-01

    Currently it can be observed that in the case of generator drives as well as 'mechanical drives' smaller units are demanded with a steam turbine capacity of up to 150 MW and clearly higher efficiencies. MAN TURBO is meeting the challenge through realisation of a comprehensive development project aiming at the extension of the application range of the current steam turbine series.

  15. Remote operated vehicle with carbon dioxide blasting (ROVCO{sub 2})

    Energy Technology Data Exchange (ETDEWEB)

    Resnick, A.M. [Oceaneering International, Inc., Upper Marlboro, MD (United States)

    1995-10-01

    The Remote Operated Vehicle with Carbon Dioxide Blasting (ROVCO{sub 2}), as shown in a front view is a six-wheeled remote land vehicle used to decontaminate concrete floors. The remote vehicle has a high pressure Cryogenesis blasting subsystem, Oceaneering Technologies (OTECH) developed a CO{sub 2} xY Orthogonal Translational End Effector (COYOTEE) subsystem, and a vacuum/filtration and containment subsystem. The cryogenesis subsystem performs the actual decontamination work and consists of the dry ice supply unit, the blasting nozzle, the remotely controlled electric and pneumatic valves, and the vacuum work-head. The COYOTEE subsystem positions the blasting work-head within a planar work space and the vacuum subsystem provides filtration and containment of the debris generated by the CO{sub 2} blasting. It employs a High Efficiency Particulate Air (HEPA) filtration unit to separate contaminants for disposal. All of the above systems are attached to the vehicle subsystem via the support structure.

  16. The use of computer blast simulations to improve blast quality

    International Nuclear Information System (INIS)

    Favreau, R.F.; Kuzyk, G.W.; Babulic, P.J.; Tienkamp, N.J.

    1989-01-01

    Atomic Energy of Canada Limited is constructing an Underground Research Laboratory (URL) as part of a comprehensive program to evaluate the concept of nuclear fuel waste disposal deep in crystalline rock formations. Careful blasting methods have been used to minimize damage to the excavation surfaces. Good wall quality is desirable in any excavation. In excavations required for nuclear waste disposal, the objective will be to minimize blast-induced fractures which may complicate the sealing requirements necessary to control subsequent movement of groundwater around a sealed disposal vault. The construction of the URL has provided an opportunity for the development of controlled blasting methods, especially for drilling accuracy and optimization of explosive loads in the perimeter and cushion holes. The work has been assisted by the use of blast simulations with the mathematical model Blaspa. This paper reviews the results of a recent project to develop a controlled method of full-face blasting, and compares the observed field results with the results of a blast simulator called Blaspa. Good agreement is found between the two, and the Blaspa results indicate quantitatively how the blasting may induce damage in the final excavation surface. In particular, the rock in the final wall may be stressed more severely by the cushion holes than by the perimeter holes. Bootleg of the rock between the perimeter and cushion rows occurs when the burst-out velocity imparted to it by the explosive loads in the perimeter holes is inadequate. In practice, these findings indicate that quantitative rock stress and rock burst-out velocity criteria can be established to minimize wall damage and bootleg. Thus, blast simulations become an efficient way to design controlled blasting and to optimize quality of the excavation surface

  17. HIGH PRODUCTIVITY VACUUM BLASTING SYSTEM

    International Nuclear Information System (INIS)

    McPhee, William S.

    1999-01-01

    The objective of this project is to improve the productivity and lower the expense of existing vacuum blasting technology. This technology is used to remove radioactive contamination, PCBs, and lead-based paint and provides worker protection by continuously recycling the material and dust for the decontamination tasks. The proposed work would increase the cleaning rate and provide safe and cost-effective decontamination of the DOE sites. This work focuses on redesigning and improving existing vacuum blasting technology including blast head nozzles, ergonomic handling of the blast head by reducing its weight; brush-ring design, vacuum level regulator, efficiency of the dust separator, and operational control sensors. The redesign is expected to enhance the productivity and economy of the vacuum blasting system by at least 50% over current vacuum blasting systems. There are three phases in the project. Phase I consists of developing and testing mathematical models. Phase II consists of pre-prototype design and fabrication and pre-prototype unit testing. Phase III consists of prototype design and field verification testing. In phase I, mathematical models are developed and analyzed for the nozzle, blast head, wind curtain, and dust separator, first as individual devices and then combined as an integrated model. This allows study of respective airflow and design parameters. The Contractor shall, based on the results of the mathematical modeling studies, design experimental models of the components and test these models. In addition, the Contractor shall develop sensors to detect the relationship of the blast head to the blast surfaces and controls to minimize the dependency on an operator's skill and judgment to obtain optimum positioning, as well as real-time characterization sensors to determine as the blast head is moving the depth to which coatings must be removed, thereby improving production and minimizing waste. In phase II, the Contractor shall design and

  18. Dual turbine power plant and method of operating such plant, especially one having an HTGR steam supply

    International Nuclear Information System (INIS)

    Braytenbah, A.S.; Jaegtnes, K.O.

    1977-01-01

    A power plant including dual steam turbine-generators connected to pass superheat and reheat steam from a steam generator which derives heat from the coolant gas of a high temperature gas-cooled nuclear reactor is described. Associated with each turbine is a bypass line to conduct superheat steam in parallel with a high pressure turbine portion, and a bypass line to conduct superheat steam in parallel with a lower pressure turbine portion. Auxiliary steam turbines pass a portion of the steam flow to the reheater of the steam generator and drive gas blowers which circulate the coolant gas through the reactor and the steam source. Apparatus and method are disclosed for loading or unloading a turbine-generator while the other produces a steady power output. During such loading or unloading, the steam flows through the turbine portions are coordinated with the steam flows through the bypass lines for protection of the steam generator, and the pressure of reheated steam is regulated for improved performance of the gas blowers. 33 claims, 5 figures

  19. Digital simulation for nuclear once-through steam generators

    International Nuclear Information System (INIS)

    Chen, A.T.

    1976-01-01

    Mathematical models for calculating the dynamic response of the Oconee type once through steam generator (OTSG) and the integral economizer once through steam generator (IEOTSG) was developed and presented in this dissertation. Linear and nonlinear models of both steam generator types were formulated using the state variable, lumped parameter approach. Transient and frequency responses of system parameters were calculated for various perturbations from both the primary coolant side and the secondary side. Transients of key parameters, including primary outlet temperature, superheated steam outlet temperature, boiling length/subcooled length and steam pressure, were generated, compared and discussed for both steam generator types. Frequency responses of delta P/sub s//deltaT/sub pin/ of the linear OTSG model were validated by using the dynamic testing results obtained at the Oconee I nuclear power station. A sensitivity analysis in both the time and the frequency domains was performed. It was concluded that the mathematical and computer models developed in this dissertation for both the OTSG and the IEOTSG are suitable for overall plant performance evaluation and steam generator related component/system design analysis for nuclear plants using either type of steam generator

  20. 1000 MW steam turbine for nuclear power station

    International Nuclear Information System (INIS)

    Drahy, J.

    1987-01-01

    Skoda Works started the manufacture of the 1000 MW steam turbine for the Temelin nuclear power plant. The turbine will use saturated steam at 3,000 r.p.m. It will allow steam supply to heat water for district heating, this of an output of 893 MW for a three-stage water heating at a temperature of 150/60 degC or of 570 MW for a two-stage heating at a temperature of 120/60 degC. The turbine features one high-pressure and three identical low-pressure stages. The pressure gradient between the high-pressure and the low-pressure parts was optimized as concerns the thermal efficiency of the cycle and the thermodynamic efficiency of the low-pressure part. A value of 0.79 MPa was selected corresponding to the maximum flow rate of the steam entering the turbine. This is 5,495 t/h, the admission steam parameters are 273.3 degC and 5.8 MPa. The feed water temperature is 220.9 degC. It is expected that throughout the life of the turbine, there will be 300 cold starts, 1,000 starts following shutdown for 55 to 88 hours, and 600 starts following shutdown for 8 hours. (Z.M.). 8 figs., 1 ref

  1. Blast tests of expedient shelters in the DICE THROW event

    International Nuclear Information System (INIS)

    Kearny, C.H.; Chester, C.V.

    1978-03-01

    To determine the worst blast environments that eight types of expedient shelters can withstand, we subjected a total of 18 shelters to the 1-kiloton blast effects of Defense Nuclear Agency's DICE THROW main event. These expedient shelters included two Russian and two Chinese types. The best shelter tested was a Small-Pole Shelter that had a box-like room of Russian design with ORNL-designed expedient blast entries and blast doors added. It was undamaged at the 53-psi peak overpressure range; the pressure rise inside was only 1.5 psi. An unmodified Russian Pole-Covered Trench Shelter was badly damaged at 6.8 psi. A Chinese ''Man'' Shelter, which skillfully uses very small poles to attain protective earth arching, survived 20 psi, undamaged. Two types of expedient shelters built of materials found in and around most American homes gave good protection at overpressures up to about 6 psi. Rug-Covered Trench Shelters were proved unsatisfactory. Water storage pits lined with ordinary plastic trash bags were proven practical at up to 53 psi, as were triangular expedient blast doors made of poles

  2. Centrifugal shot blast system

    International Nuclear Information System (INIS)

    1998-02-01

    This report describes a demonstration of Concrete cleaning, Inc., modified centrifugal shot blast technology to remove the paint coating from concrete flooring. This demonstration is part of the Chicago Pile-5 (CP-5) Large-Scale Demonstration Project (LSDP) sponsored by the US Department of Energy (DOE), office of Science and Technology (OST), Deactivation and Decommissioning Focus Area (DDFA). The objective of the LSDP is to select and demonstrate potentially beneficial technologies at the Argonne National Laboratory-East (ANL) CP-5 Research Reactor. The purpose of the LSDP is to demonstrate that using innovative and improved decontamination and decommissioning (D and D) technologies from various sources can result in significant benefits, such as decreased cost and increased health and safety, as compared with baseline D and D technologies. Potential markets exist for the innovative centrifugal shot blast system at the following sites: Fernald Environmental Management Project, Los Alamos, Nevada, Oak Ridge Y-12 and K-25, Paducah, Portsmouth Gaseous Diffusion site, and the Savannah River Site. This information is based on a revision to the OST Linkage Tables dated August 4, 1997

  3. Steam generator tube rupture effects on a LOCA

    International Nuclear Information System (INIS)

    LaChance, J.L.

    1979-01-01

    A problem currently experienced in commercial operating pressurized water reactors (PWR) in the United States is the degradation of steam generator tubes. Safety questions have arisen concerning the effect of these degraded tubes rupturing during a postulated loss-of-coolant accident (LOCA). To determine the effect of a small number of tube ruptures on the behavior of a large PWR during a postulated LOCA, a series of computer simulations was performed. The primary concern of the study was to determine whether a small number (10 or less of steam generator tubes rupturing at the beginning surface temperatures. Additional reflood analyses were performed to determine the system behavior when from 10 to 60 tubes rupture at the beginning of core reflood. The FLOOD4 code was selected as being the most applicable code for use in this study after an extensive analysis of the capabilities of existing codes to perform simulations of a LOCA with concurrent steam generator tube ruptures. The results of the study indicate that the rupturing of 10 or less steam generator tubes in any of the steam generators during a 200% cold leg break will not result in a significant increase in the peak cladding temperature. However, because of the vaporization of the steam generator secondary water in the primary side of the steam generator, a significant increase in the core pressure occurs which retards the reflooding process

  4. Chemical tailoring of steam to remediate underground mixed waste contaminents

    Science.gov (United States)

    Aines, Roger D.; Udell, Kent S.; Bruton, Carol J.; Carrigan, Charles R.

    1999-01-01

    A method to simultaneously remediate mixed-waste underground contamination, such as organic liquids, metals, and radionuclides involves chemical tailoring of steam for underground injection. Gases or chemicals are injected into a high pressure steam flow being injected via one or more injection wells to contaminated soil located beyond a depth where excavation is possible. The injection of the steam with gases or chemicals mobilizes contaminants, such as metals and organics, as the steam pushes the waste through the ground toward an extraction well having subatmospheric pressure (vacuum). The steam and mobilized contaminants are drawn in a substantially horizontal direction to the extraction well and withdrawn to a treatment point above ground. The heat and boiling action of the front of the steam flow enhance the mobilizing effects of the chemical or gas additives. The method may also be utilized for immobilization of metals by using an additive in the steam which causes precipitation of the metals into clusters large enough to limit their future migration, while removing any organic contaminants.

  5. Nuclear steam generator sludge lance method and apparatus

    International Nuclear Information System (INIS)

    Shirey, R.A.; Murray, D.E.

    1991-01-01

    This paper describes a sludge lancing system for removing sludge deposits from an interior region of a steam generator. It comprises: a peripheral fluid injection means for injecting a fluid at a high pressure about a periphery of the steam generator, the peripheral fluid injection means comprising at least one elongated fluid conduit, at least one injection nozzle and a joint positioned at a predetermined point along the elongated fluid conduit for permitting the peripheral fluid injection means to bend to a predetermined angle at the joint within the steam generator; a reciprocable fluid injection means for injecting a fluid at a high pressure toward the sludge deposits and dislodging the sludge deposits; and a supporting means positioned within the interior of the steam generator for supporting the reciprocable fluid injection means throughout the reciprocation of the reciprocable fluid injection means

  6. Steam generator tube integrity program. Phase I report

    International Nuclear Information System (INIS)

    Alzheimer, J.M.; Clark, R.A.; Morris, C.J.; Vagins, M.

    1979-09-01

    The results are presented of the pressure tests performed as part of Phase I of the Steam Generator Tube Integrity (SGTI) program at Battelle Pacific Northwest Laboratory. These tests were performed to establish margin-to-failure predictions for mechanically defected Pressurized Water Reactor (PWR) steam generator tubing under operating and accident conditions. Defect geometries tested were selected because they simulate known or expected defects in PWR steam generators. These defect geometries are Electric Discharge Machining (EDM) slots, elliptical wastage, elliptical wastage plus through-wall slot, uniform thinning, denting, denting plus uniform thinning, and denting plus elliptical wastage. All defects were placed in tubing representative of that currently used in PWR steam generators

  7. A detection of the coarse water droplets in steam turbines

    Directory of Open Access Journals (Sweden)

    Bartoš Ondřej

    2014-03-01

    Full Text Available The aim of this paper is to introduce a novel method for the detection of coarse water droplets in a low pressure part of steam turbines. The photogrammetry method has been applied for the measurement of coarse droplets in the low-pressure part of a steam turbine. A new probe based on this measurement technique was developed and tested in the laboratory and in a steam turbine in the Počerady power-plant. The probe was equipped with state-of-the-art instrumentation. The paper contains results from laboratory tests and the first preliminary measurements in a steam turbine. Possible applications of this method have been examined.

  8. Evaluation of Novel Polyunsaturated Fatty Acid Derived Lipid Mediators of Inflammation to Ameliorate the Deleterious Effects of Blast Over Pressure on Eye and Brain Visual Processing Centers in Rats

    Science.gov (United States)

    2015-08-01

    CD68 (immune cell infiltration) (Naskar, 2002; Nakazawa, 2006; Bailes , 2010). It would also be interesting to look at chronic time points far beyond...and cGMP-PDE) (Cao, 2001; Nakazawa, 2006; Rapoport, 2008; Bailes , 2010; Haung, 2012). Plasma collected from blasted rats could also be screened for...On 05 August 2015, we received a letter of full rejection, mainly due to minimal outcome measures, subtle neuronal injury effects, and lack of

  9. Steam separator uprating by elimination of capacity-limiting mechanisms

    International Nuclear Information System (INIS)

    Parkinson, J.R.; Pruster, W.P.; Kidwell, J.H.; Schneider, W.G.

    1985-01-01

    Advanced steam/water separation equipment for nuclear steam generator application is required for new equipment manufacture and also for retrofit. For new equipment applications, the desire for higher capacity is driven by competitiveness which requires maximum throughput in the most compact package. For retrofit applications, which have arisen due to the poor performance of some of the original equipment, the need is for high capacity separators which can fit into the existing pressure vessel envelope and not only correct the performance problem, but also allow for uprated plant output. This paper describes the development of such advanced steam separators

  10. Steam Generator Owners Group PWR secondary water chemistry guidelines

    International Nuclear Information System (INIS)

    Welty, C.S. Jr.; Green, S.J.

    1985-01-01

    In 1981 the Steam Generator Owners Group (SGOG), a group of domestic and foreign pressurized water reactor (PWR) owners, developed and issued the PWR secondary water chemistry guidelines. The guidelines were prepared in response to the growing recognition that a majority of the problems causing reduced steam generator reliability (e.g., denting, wasteage, pitting, etc.) were related to secondary (steam) side water purity. The guidelines were subsequently issued as an Electric Power Research Institute (EPRI) report. In 1984 they were revised to reflect industry experience in adopting the original issuance and to incorporate new information on causes of corrosion damage. The guidelines have been endorsed and their adoption recommended by the SGOG

  11. Dynamic and control of a once through steam generator

    International Nuclear Information System (INIS)

    Gomes, Arivaldo Vicente

    1979-01-01

    This paper presents a non linear distributed parameter model for the dynamics and feedback control of a large countercurrent heat exchanger used as a once through steam generator for a breeder reactor power plant. A convergent, implicit method has been developed to solve simultaneously the equations of conservation of mass, momentum and energy. The model, applicable to heat exchanger systems in general, has been used specifically to study the performance of a once-through steam generator with respect to its load following ability and stability of throttle steam temperature and pressure. (author)

  12. A reflux capsule steam generator for sodium cooled reactors

    International Nuclear Information System (INIS)

    Lantz, E.

    Pressurized water reactor plants at numerous sites have sustained significant leakage through their steam generators. The consequent shutdowns for repairs and replacements have damaged their economics. This experience suggests that if steam generators for liquid metal fast breeder reactors (LMFBR's) continue to be built as presently designed some of them will have similar problems. Because of their larger capital investment, the consequent damage to the economics of LMFBR's could be more serious. Reflux capsules provide a way to separate sodium from water and to reduce thermal stresses in steam generators for sodium cooled reactors. Their use would also eliminate the need for a primary heat exchanger and a secondary sodium loop pump. (author)

  13. Steam generator arrangement

    International Nuclear Information System (INIS)

    Ssinegurski, E.

    1981-01-01

    A steam flow path arrangement for covering the walls of the rear gas pass of a steam generator is disclosed. The entire flow passes down the sidewalls with a minor portion then passing up through the rear wall to a superheater inlet header at an intermediate elevation. The major portion of the flow passes up the front wall and through hanger tubes to a roof header. From there the major portion passes across the roof and down the rear wall to the superheater inlet header at the intermediate elevation

  14. The effect of steam cycle conditions upon the economics and design of a sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Philpott, E.F.; Pounder, F.; Willby, C.R.

    1978-01-01

    The paper studies the effect of variation of steam and feedwater conditions upon the economics, design and layout of a sodium-cooled fast reactor. The parameters investigated are steam temperature and pressure, feedwater temperature, and boiler recirculation ratio. The paper also includes an assessment of the effects of associating the fast reactor with saturated steam cycle conditions. (author)

  15. Experimental and analytical study of intermittency in direct contact condensation of steam in a cross-flow of water

    NARCIS (Netherlands)

    Clerx, N.; Geld, van der C.W.M.

    2009-01-01

    The topology of a condensing steam jet, at low steam mass fluxes, injected in a cross-flow of water has been investigatedexperimentally for various conditions (system pressure around 3 bar). The intermittent character of the steam pocket growthand collapse clearly appeared from the high speed

  16. Anatomical manifestations of primary blast ocular trauma observed in a postmortem porcine model.

    Science.gov (United States)

    Sherwood, Daniel; Sponsel, William E; Lund, Brian J; Gray, Walt; Watson, Richard; Groth, Sylvia L; Thoe, Kimberly; Glickman, Randolph D; Reilly, Matthew A

    2014-02-24

    We qualitatively describe the anatomic features of primary blast ocular injury observed using a postmortem porcine eye model. Porcine eyes were exposed to various levels of blast energy to determine the optimal conditions for future testing. We studied 53 enucleated porcine eyes: 13 controls and 40 exposed to a range of primary blast energy levels. Eyes were preassessed with B-scan and ultrasound biomicroscopy (UBM) ultrasonography, photographed, mounted in gelatin within acrylic orbits, and monitored with high-speed videography during blast-tube impulse exposure. Postimpact photography, ultrasonography, and histopathology were performed, and ocular damage was assessed. Evidence for primary blast injury was obtained. While some of the same damage was observed in the control eyes, the incidence and severity of this damage in exposed eyes increased with impulse and peak pressure, suggesting that primary blast exacerbated these injuries. Common findings included angle recession, internal scleral delamination, cyclodialysis, peripheral chorioretinal detachments, and radial peripapillary retinal detachments. No full-thickness openings of the eyewall were observed in any of the eyes tested. Scleral damage demonstrated the strongest associative tendency for increasing likelihood of injury with increased overpressure. These data provide evidence that primary blast alone (in the absence of particle impact) can produce clinically relevant ocular damage in a postmortem model. The blast parameters derived from this study are being used currently in an in vivo model. We also propose a new Cumulative Injury Score indicating the clinical relevance of observed injuries.

  17. Model for small arms fire muzzle blast wave propagation in air

    Science.gov (United States)

    Aguilar, Juan R.; Desai, Sachi V.

    2011-11-01

    Accurate modeling of small firearms muzzle blast wave propagation in the far field is critical to predict sound pressure levels, impulse durations and rise times, as functions of propagation distance. Such a task being relevant to a number of military applications including the determination of human response to blast noise, gunfire detection and localization, and gun suppressor design. Herein, a time domain model to predict small arms fire muzzle blast wave propagation is introduced. The model implements a Friedlander wave with finite rise time which diverges spherically from the gun muzzle. Additionally, the effects in blast wave form of thermoviscous and molecular relaxational processes, which are associated with atmospheric absorption of sound were also incorporated in the model. Atmospheric absorption of blast waves is implemented using a time domain recursive formula obtained from numerical integration of corresponding differential equations using a Crank-Nicholson finite difference scheme. Theoretical predictions from our model were compared to previously recorded real world data of muzzle blast wave signatures obtained by shooting a set different sniper weapons of varying calibers. Recordings containing gunfire acoustical signatures were taken at distances between 100 and 600 meters from the gun muzzle. Results shows that predicted blast wave slope and exponential decay agrees well with measured data. Analysis also reveals the persistency of an oscillatory phenomenon after blast overpressure in the recorded wave forms.

  18. Phenomenological modelling of steam explosions

    International Nuclear Information System (INIS)

    Corradini, M.L.; Drumheller, D.S.

    1980-01-01

    During a hypothetical core meltdown accident, an important safety issue to be addressed is the potential for steam explosions. This paper presents analysis and modelling of experimental results. There are four observations that can be drawn from the analysis: (1) vapor explosions are suppressed by noncondensible gases generated by fuel oxidation, by high ambient pressure, and by high water temperatures; (2) these effects appear to be trigger-related in that an explosion can again be induced in some cases by increasing the trigger magnitude; (3) direct fuel liquid-coolant liquid contact can explain small scale fuel fragmentation; (4) heat transfer during the expansion phase of the explosion can reduce the work potential

  19. CO2 pellet blasting studies

    International Nuclear Information System (INIS)

    Archibald, K.E.

    1997-01-01

    Initial tests with CO 2 pellet blasting as a decontamination technique were completed in 1993 at the Idaho Chemical Processing Plant (ICPP) at the Idaho National Engineering Laboratory (INEL). During 1996, a number of additional CO 2 pellet blasting studies with Alpheus Cleaning Technologies, Oak Ridge National Laboratory, and Pennsylvania State University were conducted. After the testing with Alpheus was complete, an SDI-5 shaved CO 2 blasting unit was purchased by the ICPP to test and determine its capabilities before using in ICPP decontamination efforts. Results of the 1996 testing will be presented in this report

  20. Steam Generator Tube Integrity Program: Surry Steam Generator Project, Hanford site, Richland, Benton County, Washington: Environmental assessment

    International Nuclear Information System (INIS)

    1980-03-01

    The US Nuclear Regulatory Commission (NRC) has placed a Nuclear Regulatory Research Order with the Richland Operations Office of the US Department of Energy (DOE) for expanded investigations at the DOE Pacific Northwest Laboratory (PNL) related to defective pressurized water reactor (PWR) steam generator tubing. This program, the Steam Generator Tube Integrity (SGTI) program, is sponsored by the Metallurgy and Materials Research Branch of the NRC Division of Reactor Safety Research. This research and testing program includes an additional task requiring extensive investigation of a degraded, out-of-service steam generator from a commercial nuclear power plant. This comprehensive testing program on an out-of-service generator will provide NRC with timely and valuable information related to pressurized water reactor primary system integrity and degradation with time. This report presents the environmental assessment of the removal, transport, and testing of the steam generator along with decontamination/decommissioning plans