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.

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

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

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.

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

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.

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

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)

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....

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

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.

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)

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.

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)

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.)

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.

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

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

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

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.

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

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)

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

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.

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

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

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.

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.

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.

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

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.

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

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.

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

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...

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.

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

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

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.)

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.)

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)

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

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

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

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

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.)

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)

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

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)

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)

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.)

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)

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)

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)

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

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

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...

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

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.

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.)

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.

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

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

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

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.

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

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

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.)

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

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)

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)

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

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.)

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

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

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...

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.

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

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%

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.

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

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)

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

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.

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

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.

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

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

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.

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

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

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

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...

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

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)

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)

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

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.

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

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

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.

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.

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.

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.

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.

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.

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.)

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

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

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)

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

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

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)

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.

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.

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.

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

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.

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...

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.

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

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....

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

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.

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

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.

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.)

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)

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)

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

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

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

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)

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

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.

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.

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.

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)

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

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.

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

Thermodynamic analysis of heat recovery steam generator in combined cycle power plant

Directory of Open Access Journals (Sweden)

Ravi Kumar Naradasu

2007-01-01

Full Text Available Combined cycle power plants play an important role in the present energy sector. The main challenge in designing a combined cycle power plant is proper utilization of gas turbine exhaust heat in the steam cycle in order to achieve optimum steam turbine output. Most of the combined cycle developers focused on the gas turbine output and neglected the role of the heat recovery steam generator which strongly affects the overall performance of the combined cycle power plant. The present paper is aimed at optimal utilization of the flue gas recovery heat with different heat recovery steam generator configurations of single pressure and dual pressure. The combined cycle efficiency with different heat recovery steam generator configurations have been analyzed parametrically by using first law and second law of thermodynamics. It is observed that in the dual cycle high pressure steam turbine pressure must be high and low pressure steam turbine pressure must be low for better heat recovery from heat recovery steam generator.

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.

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

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)

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.

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.

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.

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.)

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)

Operating results of 220 MW SKODA saturated steam turbines

International Nuclear Information System (INIS)

Drahy, J.

1992-01-01

One of the steam turbines produced by the SKODA Works, the 220 MW steam turbine for saturated admission steam of a speed of 3000 r.p.m. is described; it is used in nuclear power plants with 400 MW PWR type reactors. 16 units of 8 turbines each have been in operation in the Jaslovske Bohunice and Dukovany power plants with the total period of operation of all machines exceeding 750,000 hours. The 220 MW steam turbine consists of a two-flow high-pressure section and of two identical two-flow low-pressure sections. The pressure of saturated steam at the inlet of the high-pressure section is 4.32 MPa (the corresponding temperature of the saturation limit being 255 degC) and during the expansion in the high-pressure section it drops to 0.6 MPa; steam moisture reaches 12%. In a separator and two-stage reheater using blend steam, the steam is freed of the moisture and is reheated to a temperature of 217 degC. Some operational problems are discussed, as are the loss of the material of the stator parts of the high-pressure section due to corrosion-erosion wear and corrosion-erosion wear of the guide wheels of the high-pressure section, and measures are presented carried out for the reduction of the corrosion-erosion effects of wet steam. One of the serious problems were the fatigue fractures of the blades of the 4th high-pressure stage, which appeared after 20 000 to 24 000 hours of operation in the dented tee-root. The guide wheels of the 4th stage were substituted by new guide wheels with uniform pitch of the channels and with increased number of guide blades. Also discussed are the dynamic behavior of the low-pressure section of the bridge structure, the operating reliability and the heat off-take for water heating of long-distance heating systems. (Z.S.) 9 figs

A multiscale approach to blast neurotrauma modeling:Part II: Methodology for inducing blast injury to in vitro models

Directory of Open Access Journals (Sweden)

Gwen B. Effgen

2012-02-01

Full Text Available Due to the prominent role of improvised explosive devices (IEDs in wounding patterns of U.S. war-fighters in Iraq and Afghanistan, blast injury has risen to a new level of importance and is recognized to be a major cause of injuries to the brain. However, an injury risk-function for microscopic, macroscopic, behavioral, and neurological deficits has yet to be defined. While operational blast injuries can be very complex and thus difficult to analyze, a simplified blast injury model would facilitate studies correlating biological outcomes with blast biomechanics to define tolerance criteria. Blast-induced traumatic brain injury (bTBI results from the translation of a shock wave in air, such as that produced by an IED, into a pressure wave within the skull-brain complex. Our blast injury methodology recapitulates this phenomenon in vitro, allowing for control of the injury biomechanics via a compressed-gas shock tube used in conjunction with a custom-designed, fluid-filled receiver that contains the living culture. The receiver converts the air shock wave into a fast-rising pressure transient with minimal reflections, mimicking the intracranial pressure history in blast. We have developed an organotypic hippocampal slice culture model that exhibits cell death when exposed to a 530  17.7 kPa peak overpressure with a 1.026 ± 0.017 ms duration and 190 ± 10.7 kPa-ms impulse in-air. We have also injured a simplified in vitro model of the blood-brain barrier, which exhibits disrupted integrity immediately following exposure to 581  10.0 kPa peak overpressure with a 1.067 ms ± 0.006 ms duration and 222 ± 6.9 kPa-ms impulse in-air. To better prevent and treat bTBI, both the initiating biomechanics and the ensuing pathobiology must be understood in greater detail. A well-characterized, in vitro model of bTBI, in conjunction with animal models, will be a powerful tool for developing strategies to mitigate the risks of bTBI.

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.

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

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

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

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.

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)

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.

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

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

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)

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

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.

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

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.

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.

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

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)

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.)

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

Methods of increasing thermal efficiency of steam and gas turbine plants

Science.gov (United States)

Vasserman, A. A.; Shutenko, M. A.

2017-11-01

Three new methods of increasing efficiency of turbine power plants are described. Increasing average temperature of heat supply in steam turbine plant by mixing steam after overheaters with products of combustion of natural gas in the oxygen. Development of this idea consists in maintaining steam temperature on the major part of expansion in the turbine at level, close to initial temperature. Increasing efficiency of gas turbine plant by way of regenerative heating of the air by gas after its expansion in high pressure turbine and before expansion in the low pressure turbine. Due to this temperature of air, entering combustion chamber, is increased and average temperature of heat supply is consequently increased. At the same time average temperature of heat removal is decreased. Increasing efficiency of combined cycle power plant by avoiding of heat transfer from gas to wet steam and transferring heat from gas to water and superheated steam only. Steam will be generated by multi stage throttling of the water from supercritical pressure and temperature close to critical, to the pressure slightly higher than condensation pressure. Throttling of the water and separation of the wet steam on saturated water and steam does not require complicated technical devices.

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

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...

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.

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)

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

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)

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

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

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.

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

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.

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.

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.

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

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)

DIRECT AIR BLAST EXPOSURE EFFECTS IN ANIMALS, OPERATION UPSHOT-KNOTHOLE, PROJECT 4.2

Energy Technology Data Exchange (ETDEWEB)

DRAEGER, R.H. (UNITED STATES NAVY - DEPARTMENT OF); LEE, R.H. (UNITED STATES NAVY - DEPARTMENT OF)

1953-12-31

Project 4.2 was designed to study direct (primary) air blast injury, in animals, from an atomic weapon in the range of 20 to 50 psi under circumstances affording protection against missiles, thermal and ionizing radiation and to estimate the probable direct air blast hazard in man. The pressure levels at which atomic weapons direct air blast injuries occur will determine, to a large extent, the number of blast casualties likely to be encountered. It is probable that fatal overpressures are not reached until well within the range at which indirect (secondary) blast, thermal and ionizing radiation are practically certain to prove fatal. Only in special situations affording partial protection from other injuries are blast injuries likely to be of practical importance. Two animal species of widely different body weights (700 rats and 56 dogs) were exposed, together with air pressure recorders, in aluminum cylinders, covered by sandbags and dirt but open at both ends, at seven stations distributed within the intended overpressure range of 20 to 50 psi of Shot 10« About 200 rats were likewise exposed in Shot 9. Unfortunately, the destructive effect of the air blast of Shot 10 was much greater than anticipated. Many of the exposure cylinders were displaced and their contents destroyed. Only a partial recovery of the animals was possible due to the excessive radioactive contamination which greatly limited the time in the area. Most of the animals were dead upon recovery. Those living were in a state of severe shock. Autopsy findings showed remarkably few traumatic lesions and lung hemorrhages in spite of the rough treatment and high overpressure to which they were subjected. The rats recovered from Shot 9 were exposed to a recorded pressure of 18 to 2k psi. The autopsy findings showed moderate lung hemorrhage in most of the animals undoubtedly due to direct air blast injury. The findings were typical of those seen following exposure to air blast from HE or in the shock

Flow oscillations on the steam control valve in the middle opening condition. Clarification of the phenomena by steam flow experiment and CFD calculation

International Nuclear Information System (INIS)

Morita, Ryo; Inada, Fumio

2006-01-01

A steam control valve might cause vibrations of piping when the valve opening is in a middle condition. For rationalization of maintenance and management of the plant, the valve should be improved, but it is difficult to understand flow characteristics in detail by experiment because flow around the valve is complex 3D structure and becomes supersonic (M>1). Therefore, it is necessary to clarify the cause of the vibrations and to develop improvements by Computational Fluid Dynamics (CFD) technology. In previous researches, we clarified a mechanism of the pressure fluctuations in the middle opening condition and suggested the way to prevent the pressure fluctuations by experiments and CFD calculations. But, as we used air as a working fluid in our previous research instead of steam that is used in the power plant, we couldn't consider effects of condensation and difference of change of the quantity of state between air and steam. In this report, we have conducted steam flow experiments by multi-purpose steam experiment apparatus 'WISSH' and CFD calculations by steam flow code 'MATIS-SC' to clarify those effects. As a result, in the middle opening condition, we have observed rotating pressure fluctuations in the experiment and valve-attached flow and local high-pressure region in the CFD result. These results show the pressure fluctuations in steam experiments and CFD is same kind of the fluctuations found in air experiment and CFD. (author)

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

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.)

Flow oscillations on the steam control valve in the middle opening condition. Clarification of the effects of valve body and valve seat by steam experiments

International Nuclear Information System (INIS)

Morita, Ryo; Inada, Fumio

2007-01-01

A steam control valve might cause vibrations of piping when the valve opening is in a middle condition. For rationalization of maintenance and management of the plant, the valve should be improved, but it is difficult to understand flow characteristics in detail by experiment because flow around the valve is complex 3D structure and becomes supersonic (M>1). Therefore, it is necessary to clarify the cause of the vibrations and to develop the countermeasures by CFD (Computational Fluid Dynamics) technology. In previous researches, we clarified a mechanism of the pressure fluctuations in the middle opening condition and suggested the new valve shape (named 'Extended Valve') that can suppress the pressure fluctuations by air experiments and CFD calculations. Then, we also conducted steam experiments and CFD calculations to understand the differences between air and the steam, and found that the pressure fluctuations in the middle opening condition also occurred in the steam tests and the differences between the air and steam were not remarkable. In this report, to clarify the effects of valve and valve seat shape in steam flow condition, we conduct the steam experiments with various valve and seat shape. As a result, we find the change of the valve seat can decrease the amplitude of pressure fluctuations, but can not quite suppress the pressure fluctuations in the middle opening condition. Then, we apply the 'Extended Valve' to clarify the valve shape effect, and find that the extended valve suppresses the pressure fluctuations in the middle opening condition completely and decreases the pressure amplitude drastically. (author)

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

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

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)

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

Steam process supply optimization for Arcelormittal Tubarao consumers; Otimizacao do sistema de fornecimento de vapor de processo para a usina (AMT)

Energy Technology Data Exchange (ETDEWEB)

Loss, Gecimar; Oliveira, Heron Domingues de; Silva, Jose Geraldo Lessa; Beccalli, Marcelo; Calente, Paulo Sergio Boni; Monteiro, Sergio Anderson [Companhia Siderurgica de Tubarao ArcelorMittal, Serra, ES (Brazil)

2010-07-01

The ArcelorMittal Tubarao Energy Production area is compounded by three units: Air Separation Units, Thermal Power Plants and Thermal Recovery Power Plants. The Thermo Power Plants are co-generated units responsible to generate electrical, mechanical (Blast Furnace blower) energy and also provide Steam to complement the facility internal consumption mainly provided by CDQ plant (CDQ - Coke Dry Quenching). Since RH2 (steel treatment process) start up, the steam consumption increased and the Thermal Power Plant contribution raised to attend this new demand. Solutions were needed to guarantee the steam supply by the Power Plant even in low steam header stoppages for maintenance, since the lack of steam caused by shortage in Power Plant steam supply resulting in steel production diminution in this new scenario. (author)

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.

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

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.

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

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

Viscoelastic Materials Study for the Mitigation of Blast-Related Brain Injury

Science.gov (United States)

Bartyczak, Susan; Mock, Willis, Jr.

2011-06-01

Recent preliminary research into the causes of blast-related brain injury indicates that exposure to blast pressures, such as from IED detonation or multiple firings of a weapon, causes damage to brain tissue resulting in Traumatic Brain Injury (TBI) and Post Traumatic Stress Disorder (PTSD). Current combat helmets are not sufficient to protect the warfighter from this danger and the effects are debilitating, costly, and long-lasting. Commercially available viscoelastic materials, designed to dampen vibration caused by shock waves, might be useful as helmet liners to dampen blast waves. The objective of this research is to develop an experimental technique to test these commercially available materials when subject to blast waves and evaluate their blast mitigating behavior. A 40-mm-bore gas gun is being used as a shock tube to generate blast waves (ranging from 1 to 500 psi) in a test fixture at the gun muzzle. A fast opening valve is used to release nitrogen gas from the breech to impact instrumented targets. The targets consist of aluminum/ viscoelastic polymer/ aluminum materials. Blast attenuation is determined through the measurement of pressure and accelerometer data in front of and behind the target. The experimental technique, calibration and checkout procedures, and results will be presented.

CATHENA analysis of CANDU 6 steam generators for steam main break at a remote location

Energy Technology Data Exchange (ETDEWEB)

Liu, T. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

2009-07-01

CATHENA (Canadian Algorithm for THErmalhydraulic Network Analysis) is a nonequilibrium, two-phase, two fluid network analysis code that has been in use for over two decades by various groups in Canada and around the world. It is the primary Thermalhydraulics network analysis tool used by Atomic Energy Canada Ltd. (AECL) in the design, safety and licensing analysis of power and research reactors as well as test facilities. In the thermalhydraulic model, the liquid and vapor phases may have different pressures, velocities, and temperatures. The objective of the present paper is to present the detailed modeling of a CANDU 6 Steam Generator (SG) using the transient, thermalhydraulics network code CATHENA. The model represents the secondary side, primary side and the main steam system including the main steam line up to the assumed break location. The present model is designed such that the transient pressure drops across Tube Support Plates (TSP) could be extracted. The resistances of degraded/fouled TSPs were modeled by using the increased/reduced flow area of the TSPs. CATHENA then calculates the flow resistance in two-phase flow based on the area contraction/expansion at the TSPs. Three sets of simulations were performed; one with the degraded steam generator data provided by the utility users, and the other two with waterlanced (cleaned with high pressure water jet) TSPs. One run assumed the flow area increased by 25 percent, the other run assumed the flow area increased by 50 percent. on the hot side of the SG. No significant changes in the break discharge rates were observed between the runs. However, the steam generator downcomer flow for the waterlanced case did not reverse during the blowdown as was calculated for the degraded case. As expected, the pressure drop across the TSPs was decreased in the waterlanced cases comparing with degraded cases. The CATHENA simulation provides estimates of the velocity, density, and quality in the tube bundle as well as

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.)

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

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.

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

Evaluation of acoustic resonance at branch section in main steam line. Part 2. Proposal of method for predicting resonance frequency in steam flow

International Nuclear Information System (INIS)

Uchiyama, Yuta; Morita, Ryo

2012-01-01

Flow-induced acoustic resonances of piping system containing closed side-branches are sometimes encountered in power plants. Acoustic standing waves with large amplitude pressure fluctuation in closed side-branches are excited by the unstable shear layer which separates the mean flow in the main piping from the stagnant fluid in the branch. In U.S. NPP, the steam dryer had been damaged by high cycle fatigue due to acoustic-induced vibration under a power uprating condition. Our previous research developed the method for evaluating the acoustic resonance at the branch sections in actual power plants by using CFD. In the method, sound speed in wet steam is evaluated by its theory on the assumption of homogeneous flow, although it may be different from practical sound speed in wet steam. So, it is necessary to consider and introduce the most suitable model of practical sound speed in wet steam. In addition, we tried to develop simplified prediction method of the amplitude and frequency of pressure fluctuation in wet steam flow. Our previous experimental research clarified that resonance amplitude of fluctuating pressure at the top of the branch in wet steam. However, the resonance frequency in steam condition could not be estimated by using theoretical equation as the end correction in steam condition and sound speed in wet steam is not clarified as same reason as CFD. Therefore, in this study, we tried to evaluate the end correction in each dry and wet steam and sound speed of wet steam from experimental results. As a result, method for predicting resonance frequency by using theoretical equation in each wet and dry steam condition was proposed. (author)

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.

Cavity pressure/residual stress measurements from the Non-Proliferation Experiment

Energy Technology Data Exchange (ETDEWEB)

Heinle, R.A.; Hudson, B.C. [Lawrence Livermore National Lab., CA (United States); Hatch, M.A. Jr.

1994-12-31

The Lawrence Livermore National Laboratory planned and conducted experiments on the Non-Proliferation Experiment to determine post-detonation gas pressure inside the explosive cavity and the residual rock stress in the region immediately outside the cavity. Before detonation there was significant concern that steam and detonation products would create very high temperatures and pressure in the blast cavity that would exist for weeks and months after firing. This could constitute a safety hazard to personnel re-entering the tunnel. Consequently the Lawrence Livermore National Laboratory was asked to field its Cavity Pressure/Residual stress monitor system on the Non-Proliferation Experiment. We obtained experimental data for the first 600 ms after the explosion and again several weeks after detonation upon tunnel re-entry. We recorded early-time cavity pressure of about 8.3 MPa. In addition we believe that the ends of our sensor hoses were subjected to an ambient driving pressure of about 0.5 MPa (absolute) that persisted until at least three weeks after zero time.

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

Draining down of a nuclear steam generating system

International Nuclear Information System (INIS)

Jawor, J.C.

1987-01-01

The method is described of draining down contained reactor-coolant water from the inverted vertical U-tubes of a 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 is 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. The method comprises continuously 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 tube sheet 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

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

Dual turbine power plant and a reheat steam bypass flow control system for use therein

International Nuclear Information System (INIS)

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

1977-01-01

An electric power plant having dual turbine-generators connected to a steam source that includes a high temperature gas cooled nuclear reactor is described. Each turbine comprises a high pressure portion operated by superheat steam and an intermediate-low pressure portion operated by reheat steam; a bypass line is connected across each turbine portion to permit a desired minimum flow of steam from the source at times when the combined flow of steam through the turbine is less than the minimum. Coolant gas is propelled through the reactor by a circulator which is driven by an auxiliary turbine which uses steam exhausted from the high pressure portions and their bypass lines. The pressure of the reheat steam is controlled by a single proportional-plus-integral controller which governs the steam flow through the bypass lines associated with the intermediate-low pressure portions. At times when the controller is not in use its output signal is limited to a value that permits an unbiased response when pressure control is resumed, as in event of a turbine trip. 25 claims, 2 figures

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

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.

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.

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)

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...

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

ORTURB, HTGR Steam Turbine Dynamic for FSV Reactor

International Nuclear Information System (INIS)

Conklin, J.C.

2001-01-01

1 - Description of program or function: ORTURB was written specifically to calculate the dynamic behavior of the Fort St. Vrain (FSV) High- Temperature Gas-Cooled Reactor (HTGR) steam turbines. The program is divided into three main parts: the driver subroutine; turbine subroutines to calculate the pressure-flow balance of the high-, intermediate-, and low-pressure turbines; and feedwater heater subroutines. 2 - Method of solution: The program uses a relationship derived for ideal gas flow in an iterative fashion that minimizes computational time to determine the pressure and flow in the FSV steam turbines as a function of plant transient operating conditions. An important computer modeling characteristic, unique to FSV, is that the high-pressure turbine exhaust steam is used to drive the reactor core coolant circulators prior to entering the reheater. A feedwater heater dynamic simulation model utilizing seven state variables for each of the five heaters is included in the ORTURB computer simulation of the regenerative Rankine cycle steam turbines. The seven temperature differential equations are solved at each time- step using a matrix exponential method. 3 - Restrictions on the complexity of the problem: The turbine shaft is assumed to rotate at a constant (rated) speed of 3600 rpm. Energy and mass storage of steam in the high-, intermediate-, and low-pressure turbines is assumed to be negligible. These limitations exclude the use of ORTURB during a turbine transient such as startup from zero power or very low turbine flows

Ex-Vessel Steam Explosion Analysis of Central Melt Pour Scenario

International Nuclear Information System (INIS)

Ursic, M.; Leskovar, M.

2008-01-01

An ex-vessel steam explosion may develop during a severe reactor accident when the reactor vessel fails and the molten core interacts with the coolant in the reactor cavity. At this process part of the corium energy is intensively transferred to water in a very short time scale. The water vaporizes at high pressure and expands, doing work on its surrounding. Although the steam explosion has probably a low probability of occurrence, it is an important nuclear safety issue in case of a severe reactor accident. Namely, the formed very high pressure region induces dynamic loadings on the surrounding structures that may potentially lead to an early release of the radioactive material into the environment. Although the steam explosion events have being studied for several years, the level of the process and consequences understanding is still not adequate. To increase the level of confidence the OECD programme SERENA (Steam Explosion REsolution for Nuclear Applications) was established in 2002. The objectives of the program were to evaluate capabilities of the current generation of the FCI (Fuel-Coolant Interaction) computer codes in predicting the steam explosion induced loads, identifying key FCI phenomena and associated uncertainties impacting the predictability of the steam explosion energetics in the reactor situations and proposing confirmatory research to reduce the uncertainties to acceptable levels for the steam explosion risk assessment. To get a better insight into the most challenging ex-vessel steam explosions, analyses for different locations of the melt release, the cavity water sub-cooling, the primary system pressure overpressure and the triggering time were preformed for a typical pressurized water reactor cavity. The results of some scenarios revealed that significantly higher pressure loads are predicted than obtained in the OECD programme SERENA Phase 1. Among the performed analyses for the central melt pour scenarios, the maximum pressure loads were

Nuclear steam generator tube to tubesheet joint optimization

International Nuclear Information System (INIS)

McGregor, Rod

1999-01-01

Industry-wide problems with Stress Corrosion Cracking in the Nuclear Steam Generator tube-to-tubesheet joint have led to costly repairs, plugging, and replacement of entire vessels. To improve corrosion resistance, new and replacement Steam Generator developments typically employ the hydraulic tube expansion process (full depth) to minimize tensile residual stresses and cold work at the critical transition zone between the expanded and unexpanded tube. These variables have undergone detailed study using specialized X-ray diffraction and analytical techniques. Responding to increased demands from Nuclear Steam Generator operators and manufacturers to credit the leak-tightness and strength contributions of the hydraulic expansion, various experimental tasks with complimentary analytical modelling were applied to improve understanding and control of tube to hole contact pressure. With careful consideration to residual stress impact, design for strength/leak tightness optimization addresses: Experimentally determined minimum contact pressure levels necessary to preclude incipient leakage into the tube/hole interface. The degradation of contact pressure at surrounding expansions caused by the sequential expansion process. The transient and permanent contact pressure variation associated with tubesheet hole dilation during Steam Generator operation. An experimental/analytical simulation has been developed to reproduce cyclic Steam Generator operating strains on the tubesheet and expanded joint. Leak tightness and pullout tests were performed during and following simulated Steam Generator operating transients. The overall development has provided a comprehensive understanding of the fabrication and in-service mechanics of hydraulically expanded joints. Based on this, the hydraulic expansion process can be optimized with respect to critical residual stress/cold work and the strength/leakage barrier criteria. (author)

A balanced strategy in managing steam generator thermal performance

International Nuclear Information System (INIS)

Hu, M. H.; Nelson, P. R.

2009-01-01

This paper presents a balanced strategy in managing thermal performance of steam generator designed to deliver rated megawatt thermal (MWt) and megawatt electric (MWe) power without loss with some amount of thermal margin. A steam generator (SG) is a boiling heat exchanger whose thermal performance may degrade because of steam pressure loss. In other words, steam pressure loss is an indicator of thermal performance degradation. Steam pressure loss is mainly a result of either 1) tube scale induced poor boiling or 2) tube plugging historically resulting from tubing corrosion, wear due to flow induced tube vibration or loose parts impact. Thermal performance degradation was historically due to tube plugging but more recently it is due to poor boiling caused by more bad than good constituents of feedwater impurities. The whole SG industry still concentrates solely on maintenance programs towards preventing causes for tube plugging and yet almost no programs on maintaining adequate boiling of fouled tubes. There can be an acceptable amount of tube scale that provides excellent boiling capacity without tubing corrosion, as operational experience has repeatedly demonstrated. Therefore, future maintenance has to come up balanced programs for allocating limited resources in both maintaining good boiling capacity and preventing tube plugging. This paper discusses also thermal performance degradation due to feedwater impurity induced blockage of tube support plate and thus subsequent water level oscillations, and how to mitigate them. This paper provides a predictive management of tube scale for maintaining adequate steam pressure and stable water level without loss in MWt/MWe or recovering from steam pressure loss or water level oscillations. This paper offers a balanced strategy in managing SG thermal performance to fulfill its mission. Such a strategy is even more important in view of the industry trend in pursuing extended power uprate as high as 20 percent

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...

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

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

Changes of the more relevant PHTS parameters after the cleaning of the steam generators primary side at Embalse nuclear power plant

International Nuclear Information System (INIS)

Moreno, Carlos A.; Coutsiers, Ernesto; Acevedo, Paul; Pomerantz, Marcelo E.

2003-01-01

During the operation of the plant magnetite deposition occurs at the inner walls of Primary Heat Transport System (PHTS). This deposition is particularly significant at the U-tubes of steam generators. The consequence of this is the deterioration of heat transfer to the Secondary System. In order to minimize this impact, during the annual outage of 2000, the steam generators primary side cleaning by the SIVABLAST technique was carried out. This technique consists in blasting the inner walls with tiny stainless steel balls propelled by air at high pressure. This paper presents the change of the more relevant parameters of PHTS after that cleaning. The parameters analyzed and the main results are the following: 1) Inlet header temperature dropped 4.7 C degrees at full power; 2) Exit quality at the outlet headers decreased from 3,5% to 1,5%; 3) Global PHTS flow in single phase evaluated from: a) In-site instrumentation increased 4,6%; b) Thermalhydraulic code NUCIRC 1.0 increased 3,2%; c) measured flows at the instrumented fuel channels increased 4.4%. (author)

Thermodynamics of Hydrogen Production from Dimethyl Ether Steam Reforming and Hydrolysis

Energy Technology Data Exchange (ETDEWEB)

T.A. Semelsberger

2004-10-01

The thermodynamic analyses of producing a hydrogen-rich fuel-cell feed from the process of dimethyl ether (DME) steam reforming were investigated as a function of steam-to-carbon ratio (0-4), temperature (100 C-600 C), pressure (1-5 atm), and product species: acetylene, ethanol, methanol, ethylene, methyl-ethyl ether, formaldehyde, formic acid, acetone, n-propanol, ethane and isopropyl alcohol. Results of the thermodynamic processing of dimethyl ether with steam indicate the complete conversion of dimethyl ether to hydrogen, carbon monoxide and carbon dioxide for temperatures greater than 200 C and steam-to-carbon ratios greater than 1.25 at atmospheric pressure (P = 1 atm). Increasing the operating pressure was observed to shift the equilibrium toward the reactants; increasing the pressure from 1 atm to 5 atm decreased the conversion of dimethyl ether from 99.5% to 76.2%. The order of thermodynamically stable products in decreasing mole fraction was methane, ethane, isopropyl alcohol, acetone, n-propanol, ethylene, ethanol, methyl-ethyl ether and methanol--formaldehyde, formic acid, and acetylene were not observed. The optimal processing conditions for dimethyl ether steam reforming occurred at a steam-to-carbon ratio of 1.5, a pressure of 1 atm, and a temperature of 200 C. Modeling the thermodynamics of dimethyl ether hydrolysis (with methanol as the only product considered), the equilibrium conversion of dimethyl ether is limited. The equilibrium conversion was observed to increase with temperature and steam-to-carbon ratio, resulting in a maximum dimethyl ether conversion of approximately 68% at a steam-to-carbon ratio of 4.5 and a processing temperature of 600 C. Thermodynamically, dimethyl ether processed with steam can produce hydrogen-rich fuel-cell feeds--with hydrogen concentrations exceeding 70%. This substantiates dimethyl ether as a viable source of hydrogen for PEM fuel cells.

Analysis of ex-vessel steam explosion with MC3D

International Nuclear Information System (INIS)

Leskovar, M.; Mavko, B.

2007-01-01

An ex-vessel steam explosion may occur when, during a severe reactor accident, the reactor vessel fails and the molten core pours into the water in the reactor cavity. A steam explosion is a fuel coolant interaction process where the heat transfer from the melt to water is so intense and rapid that the timescale for heat transfer is shorter than the timescale for pressure relief. This can lead to the formation of shock waves and production of missiles that may endanger surrounding structures. A strong enough steam explosion in a nuclear power plant could jeopardize the containment integrity and so lead to a direct release of radioactive material to the environment. In the paper, different scenarios of ex-vessel steam explosions in a typical pressurized water reactor cavity are analyzed with the code MC3D, which was developed for the simulation of fuel-coolant interactions. A comprehensive parametric study was performed varying the location of the melt release (central, left and right side melt pour), the cavity water subcooling, the primary system overpressure at vessel failure and the triggering time for explosion calculations. The main purpose of the study was to determine the most challenging ex-vessel steam explosion cases in a typical pressurized water reactor and to estimate the expected pressure loadings on the cavity walls. The performed analysis shows that for some ex-vessel steam explosion scenarios significantly higher pressure loads are predicted than obtained in the OECD programme SERENA Phase 1. (author)

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

Genetic optimization of steam multi-turbines system

International Nuclear Information System (INIS)

Olszewski, Pawel

2014-01-01

Optimization analysis of partially loaded cogeneration, multiple-stages steam turbines system was numerically investigated by using own-developed code (C++). The system can be controlled by following variables: fresh steam temperature, pressure, and flow rates through all stages in steam turbines. Five various strategies, four thermodynamics and one economical, which quantify system operation, were defined and discussed as an optimization functions. Mathematical model of steam turbines calculates steam properties according to the formulation proposed by the International Association for the Properties of Water and Steam. Genetic algorithm GENOCOP was implemented as a solving engine for non–linear problem with handling constrains. Using formulated methodology, example solution for partially loaded system, composed of five steam turbines (30 input variables) with different characteristics, was obtained for five strategies. The genetic algorithm found multiple solutions (various input parameters sets) giving similar overall results. In real application it allows for appropriate scheduling of machine operation that would affect equable time load of every system compounds. Also based on these results three strategies where chosen as the most complex: the first thermodynamic law energy and exergy efficiency maximization and total equivalent energy minimization. These strategies can be successfully used in optimization of real cogeneration applications. - Highlights: • Genetic optimization model for a set of five various steam turbines was presented. • Four various thermodynamic optimization strategies were proposed and discussed. • Operational parameters (steam pressure, temperature, flow) influence was examined. • Genetic algorithm generated optimal solutions giving the best estimators values. • It has been found that similar energy effect can be obtained for various inputs

Investigations to the potential of the high temperature reactor for steam power processes with highest steam conditions and comparison with according conventional power plants

International Nuclear Information System (INIS)

Mondry, M.

1988-04-01

Already in the fifties conventional power plants with high parameters of the live steam were built to improve the total efficiency. The power plant with the highest steam conditions in the Federal Republic of Germany has 300 bar pressure and 600deg C temperature. Because of high material costs and other problems power plants with such high conditions were not continued to be built. Standard conditions of today's power plants are in the order of 180-250 bar pressure and 535deg C temperature. As the high temperature reactor is partly built up in another way than a conventional power plant, the results regarding the high steam parameters are not transferable. Possibilities for the technical realization of determined HTR-specific components are introduced and discussed. Then different HTR-power plants with steam conditions up to 350 bar pressure and 650deg C temperature are projected. Economical considerations show that an HTR with higher steam parameters brings financial profits. Further efficiency increase, which is possible by the high steam conditions, is shortly presented. The work ends with a technical and economical comparison of corresponding conventional power plants. (orig./UA) [de

Simulation and scaling analysis of a spherical particle-laden blast wave

Science.gov (United States)

Ling, Y.; Balachandar, S.

2018-05-01

A spherical particle-laden blast wave, generated by a sudden release of a sphere of compressed gas-particle mixture, is investigated by numerical simulation. The present problem is a multiphase extension of the classic finite-source spherical blast-wave problem. The gas-particle flow can be fully determined by the initial radius of the spherical mixture and the properties of gas and particles. In many applications, the key dimensionless parameters, such as the initial pressure and density ratios between the compressed gas and the ambient air, can vary over a wide range. Parametric studies are thus performed to investigate the effects of these parameters on the characteristic time and spatial scales of the particle-laden blast wave, such as the maximum radius the contact discontinuity can reach and the time when the particle front crosses the contact discontinuity. A scaling analysis is conducted to establish a scaling relation between the characteristic scales and the controlling parameters. A length scale that incorporates the initial pressure ratio is proposed, which is able to approximately collapse the simulation results for the gas flow for a wide range of initial pressure ratios. This indicates that an approximate similarity solution for a spherical blast wave exists, which is independent of the initial pressure ratio. The approximate scaling is also valid for the particle front if the particles are small and closely follow the surrounding gas.

Simulation and scaling analysis of a spherical particle-laden blast wave

Science.gov (United States)

Ling, Y.; Balachandar, S.

2018-02-01

A spherical particle-laden blast wave, generated by a sudden release of a sphere of compressed gas-particle mixture, is investigated by numerical simulation. The present problem is a multiphase extension of the classic finite-source spherical blast-wave problem. The gas-particle flow can be fully determined by the initial radius of the spherical mixture and the properties of gas and particles. In many applications, the key dimensionless parameters, such as the initial pressure and density ratios between the compressed gas and the ambient air, can vary over a wide range. Parametric studies are thus performed to investigate the effects of these parameters on the characteristic time and spatial scales of the particle-laden blast wave, such as the maximum radius the contact discontinuity can reach and the time when the particle front crosses the contact discontinuity. A scaling analysis is conducted to establish a scaling relation between the characteristic scales and the controlling parameters. A length scale that incorporates the initial pressure ratio is proposed, which is able to approximately collapse the simulation results for the gas flow for a wide range of initial pressure ratios. This indicates that an approximate similarity solution for a spherical blast wave exists, which is independent of the initial pressure ratio. The approximate scaling is also valid for the particle front if the particles are small and closely follow the surrounding gas.

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

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)

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...

A mathematical model of steam-drum dynamics

International Nuclear Information System (INIS)

Moeck, E.O.; Hinds, H.W.

1976-12-01

Mathematical equations describing the dynamic behaviour of pressure, water mass, etc. in a steam drum are derived from basic principles. The resultant model includes such effects as steam superheating and water subcooling as well as spontaneous flashing of liquid and condensation of vapour. Experimental data from a pressurizer are adequately predicted by the model. The pressure rise following a turbine trip can be predicted by the isentropic-compression model but not by the thermodynamic-equilibrium model. The equations are individually linearized and implemented on an analog computer in such a way that their non-linear behaviour is retained for small-perturbation studies. (author)

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

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

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.)

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)

Structural Changes of Lignin from Wheat Straw by Steam Explosion and Ethanol Pretreatments

Directory of Open Access Journals (Sweden)

Cheng Pan

2016-06-01

Full Text Available Effects of the pretreatment of wheat straw by steam explosion and ethanol were evaluated relative to the structural changes of lignin from the pretreated pulp. The lignin from steam explosion pulp (LS, lignin from steam blasting residual liquid (LL, lignin from ethanol pretreatment pulp (LE, lignin from black liquor (LB, and lignin from wheat straw (LW were separated, and the structural characteristics of the lignin fractions were compared based on analyses of Fourier transform-infrared, ultraviolet, thermogravimetric, and 1H and 13C nuclear magnetic resonance spectra. The proportions of the three structural units in all lignin fractions clearly changed during the pretreatment process because of inter-conversion reactions. The conjugated structure of lignin was destroyed in the pretreatment process and was also affected by the alkali extraction process. The alcoholic hydroxyl links on the aliphatic side chain were partly transformed into carbonyl groups during ethanol pretreatment. Demethoxylation occurred in all lignin fractions during the ethanol pretreatment and steam explosion process. The thermal stability of the LB fraction was relatively high because of the condensation reaction.

400-MWe consolidated nuclear steam system (CNSS): 1200-MWt/conceptual design

International Nuclear Information System (INIS)

1977-06-01

A 1200-MWt consolidated nuclear steam system (CNSS) conceptual design is described. The concept, derived from nuclear merchant ship propulsion steam systems but distinctly different from those systems in detail, incorporates the steam generators within the reactor pressure vessel. This configuration eliminates primary coolant circulating piping external to the reactor pressure vessel since the primary coolant circulating pumps are mounted in the pressure vessel head. So arranged, the maximum piping break that must be assumed is that of the pressurizer surge line, which is substantially smaller than a primary coolant circulating line. A fracture of the pressurizer surge line would result in substantially lower mass and energy release rates of the primary coolant during the assumed loss-of-coolant accident. This in turn makes practical a pressure-suppression containment rather than the ''dry'' containment commonly used for pressurized water reactors

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.

Development of Technologies on Innovative-Simplified Nuclear Power Plant Using High-Efficiency Steam Injectors (12) Evaluations of Spatial Distributions of Flow and Heat Transfer in Steam Injector

International Nuclear Information System (INIS)

Yutaka Abe; Yujiro Kawamoto; Chikako Iwaki; Tadashi Narabayashi; Michitsugu Mori; Shuichi Ohmori

2006-01-01

Next-generation nuclear reactor systems have been under development aiming at simplified system and improvement of safety and credibility. One of the innovative technologies is the supersonic steam injector, which has been investigated as one of the most important component of the next-generation nuclear reactor. The steam injector has functions of a passive pump without large motor or turbo-machinery and a high efficiency heat exchanger. The performances of the supersonic steam injector as a pump and a heat exchanger are dependent on direct contact condensation phenomena between a supersonic steam and a sub-cooled water jet. In previous studies of the steam injector, there are studies about the operating characteristics of steam injector and about the direct contact condensation between static water pool and steam in atmosphere. However, there is a little study about the turbulent heat transfer and flow behavior under the great shear stress. In order to examine the heat transfer and flow behavior in supersonic steam injector, it is necessary to measure the spatial temperature distribution and velocity in detail. The present study, visible transparent supersonic steam injector is used to obtain the axial pressure distributions in the supersonic steam injector, as well as high speed visual observation of water jet and steam interface. The experiments are conducted with and without non-condensable gas. The experimental results of the interfacial flow behavior between steam and water jet are obtained. It is experimentally clarified that an entrainment exists on the water jet surface. It is also clarified that discharge pressure is depended on the steam supply pressure, the inlet water flow rate, the throat diameter and non-condensable flow rate. Finally a heat flux is estimated about 19 MW/m 2 without non-condensable gas condition in steam. (authors)

A three-dimensional laboratory steam injection model allowing in situ saturation measurements

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.

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

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.

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.

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

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

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

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

Feasibility and application on steam injector for next-generation reactor

International Nuclear Information System (INIS)

Narabayashi, Tadashi; Ishiyama, Takenori; Miyano, Hiroshi; Nei, Hiromichi; Shioiri, Akio

1991-01-01

A feasibility study has been conducted on steam injector for a next generation reactor. The steam injector is a simple, compact passive device for water injection, such as Passive Core Injection System (PCIS) of Passive Containment Cooling System (PCCS), because of easy start-up without an AC power. An analysis model for a steam injector characteristics has been developed, and investigated with a visualized fundamental test for a two-stage Steam Injector System (SIS) for PCIS and a one-stage low pressure SIS for PCCS. The test results showed good agreement with the analysis results. The analysis and the test results showed the SIS could work over a very wide range of the steam pressure, and is applicable for PCIS or PCCS in the next generation reactors. (author)

Steam turbine chemistry in light water reactor plants

International Nuclear Information System (INIS)

Svoboda, Robert; Haertel, Klaus

2008-01-01

Steam turbines in boiling water reactor (BWR) and pressurized water reactor (PWR) power plants of various manufacturers have been affected by corrosion fatigue and stress corrosion cracking. Steam chemistry has not been a prime focus for related research because the water in nuclear steam generating systems is considered to be of high purity. Steam turbine chemistry however addresses more the problems encountered in fossil fired power plants on all volatile treatment, where corrosive environments can be formed in zones where wet steam is re-evaporated and dries out, or in the phase transition zone, where superheated steam starts to condense in the low-pressure (LP) turbine. In BWR plants the situation is aggravated by the fact that no alkalizing agents are used in the cycle, thus making any anionic impurity immediately acidic. This is illustrated by case studies of pitting corrosion of a 12 % Cr steel gland seal and of flow-oriented corrosion attack on LP turbine blades in the phase transition zone. In PWR plants, volatile alkalizing agents are used that provide some buffering of acidic impurities, but they also produce anionic decomposition products. (orig.)

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...

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

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

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

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

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)

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.

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.

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)

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

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

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)

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

Steam generator tube rupture simulation using extended finite element method

Energy Technology Data Exchange (ETDEWEB)

Mohanty, Subhasish, E-mail: smohanty@anl.gov; Majumdar, Saurin; Natesan, Ken

2016-08-15

Highlights: • Extended finite element method used for modeling the steam generator tube rupture. • Crack propagation is modeled in an arbitrary solution dependent path. • The FE model is used for estimating the rupture pressure of steam generator tubes. • Crack coalescence modeling is also demonstrated. • The method can be used for crack modeling of tubes under severe accident condition. - Abstract: A steam generator (SG) is an important component of any pressurized water reactor. Steam generator tubes represent a primary pressure boundary whose integrity is vital to the safe operation of the reactor. SG tubes may rupture due to propagation of a crack created by mechanisms such as stress corrosion cracking, fatigue, etc. It is thus important to estimate the rupture pressures of cracked tubes for structural integrity evaluation of SGs. The objective of the present paper is to demonstrate the use of extended finite element method capability of commercially available ABAQUS software, to model SG tubes with preexisting flaws and to estimate their rupture pressures. For the purpose, elastic–plastic finite element models were developed for different SG tubes made from Alloy 600 material. The simulation results were compared with experimental results available from the steam generator tube integrity program (SGTIP) sponsored by the United States Nuclear Regulatory Commission (NRC) and conducted at Argonne National Laboratory (ANL). A reasonable correlation was found between extended finite element model results and experimental results.

Steam generator tube rupture simulation using extended finite element method

International Nuclear Information System (INIS)

Mohanty, Subhasish; Majumdar, Saurin; Natesan, Ken

2016-01-01

Highlights: • Extended finite element method used for modeling the steam generator tube rupture. • Crack propagation is modeled in an arbitrary solution dependent path. • The FE model is used for estimating the rupture pressure of steam generator tubes. • Crack coalescence modeling is also demonstrated. • The method can be used for crack modeling of tubes under severe accident condition. - Abstract: A steam generator (SG) is an important component of any pressurized water reactor. Steam generator tubes represent a primary pressure boundary whose integrity is vital to the safe operation of the reactor. SG tubes may rupture due to propagation of a crack created by mechanisms such as stress corrosion cracking, fatigue, etc. It is thus important to estimate the rupture pressures of cracked tubes for structural integrity evaluation of SGs. The objective of the present paper is to demonstrate the use of extended finite element method capability of commercially available ABAQUS software, to model SG tubes with preexisting flaws and to estimate their rupture pressures. For the purpose, elastic–plastic finite element models were developed for different SG tubes made from Alloy 600 material. The simulation results were compared with experimental results available from the steam generator tube integrity program (SGTIP) sponsored by the United States Nuclear Regulatory Commission (NRC) and conducted at Argonne National Laboratory (ANL). A reasonable correlation was found between extended finite element model results and experimental results.

Modeling and simulation of blast-induced, early-time intracranial wave physics leading to traumatic brain injury.

Energy Technology Data Exchange (ETDEWEB)

Ford, Corey C. (University of New Mexico, Albuquerque, NM); Taylor, Paul Allen

2008-02-01

The objective of this modeling and simulation study was to establish the role of stress wave interactions in the genesis of traumatic brain injury (TBI) from exposure to explosive blast. A high resolution (1 mm{sup 3} voxels), 5 material model of the human head was created by segmentation of color cryosections from the Visible Human Female dataset. Tissue material properties were assigned from literature values. The model was inserted into the shock physics wave code, CTH, and subjected to a simulated blast wave of 1.3 MPa (13 bars) peak pressure from anterior, posterior and lateral directions. Three dimensional plots of maximum pressure, volumetric tension, and deviatoric (shear) stress demonstrated significant differences related to the incident blast geometry. In particular, the calculations revealed focal brain regions of elevated pressure and deviatoric (shear) stress within the first 2 milliseconds of blast exposure. Calculated maximum levels of 15 KPa deviatoric, 3.3 MPa pressure, and 0.8 MPa volumetric tension were observed before the onset of significant head accelerations. Over a 2 msec time course, the head model moved only 1 mm in response to the blast loading. Doubling the blast strength changed the resulting intracranial stress magnitudes but not their distribution. We conclude that stress localization, due to early time wave interactions, may contribute to the development of multifocal axonal injury underlying TBI. We propose that a contribution to traumatic brain injury from blast exposure, and most likely blunt impact, can occur on a time scale shorter than previous model predictions and before the onset of linear or rotational accelerations traditionally associated with the development of TBI.

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

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.

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)

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.)

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

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

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...

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.

1979-01-01

In accordance with the present invention, a power plant includes a steam source to generate superheat and reheat steam which flows through a turbine-generator and an associated bypass system. A high-pressure and an intermediate-pressure turbine portion drive a first electrical generating means, and a low-pressure turbine portion drives a second electrical generating means. A first flow of superheat steam flows through the high-pressure portion, while a second flow of reheat steam flows through the intermediate and low-pressure portions in succession. Provision is made for bypassing steam around the turbine portions; in particular, one bypass means permits a flow of superheat steam from the steam source to the exhaust of the high-pressure portion, and another bypass means allows reheated steam to pass from the source to the exhaust of the low-pressure portion. The first and second steam flows are governed independently. While one of such flows is varied for purposes of controlling the rotational speed of the first generating means according to a desired speed, the other flow is varied to regulate a power plant variable at its desired level. (author)

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)

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

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

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.)

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

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.

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.)

Large-leak sodium-water reaction analysis for steam generators

International Nuclear Information System (INIS)

Sakano, K.; Shindo, Y.; Hori, M.

1975-01-01

The guillotine rupture of 4 tubes is assumed as a design basis regarding the large-leak sodium-water reaction in the system of the MONJU steam generator. Three kinds of analyses were performed with the view to showing the integrity of the steam generator system on the reaction. The first one is the analysis of the initial pressure spike, assuming the initial guillotine rupture of 1 tube. The analysis was performed by utilizing one-dimensional sphere-cylinder model code SWAC-7 and two-dimensional axisymmetric code PISCES 2DL. The second one is the analysis of the secondary peak pressure and its propagation in the system, assuming the instantaneous guillotine rupture of 4 tubes. The third one is the analysis of the dynamic deformation of the steam generator shell. The integrity of the steam generator system was shown by the analyses. (author)

Large-leak sodium-water reaction analysis for steam generators

Energy Technology Data Exchange (ETDEWEB)

Sakano, K; Shindo, Y; Hori, M

1975-07-01

The guillotine rupture of 4 tubes is assumed as a design basis regarding the large-leak sodium-water reaction in the system of the MONJU steam generator. Three kinds of analyses were performed with the view to showing the integrity of the steam generator system on the reaction. The first one is the analysis of the initial pressure spike, assuming the initial guillotine rupture of 1 tube. The analysis was performed by utilizing one-dimensional sphere-cylinder model code SWAC-7 and two-dimensional axisymmetric code PISCES 2DL. The second one is the analysis of the secondary peak pressure and its propagation in the system, assuming the instantaneous guillotine rupture of 4 tubes. The third one is the analysis of the dynamic deformation of the steam generator shell. The integrity of the steam generator system was shown by the analyses. (author)

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.

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.

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

Two-dimensional modeling of water spray cooling in superheated steam

Directory of Open Access Journals (Sweden)

Ebrahimian Vahid

2008-01-01

Full Text Available Spray cooling of the superheated steam occurs with the interaction of many complex physical processes, such as initial droplet formation, collision, coalescence, secondary break up, evaporation, turbulence generation, and modulation, as well as turbulent mixing, heat, mass and momentum transfer in a highly non-uniform two-phase environment. While it is extremely difficult to systematically study particular effects in this complex interaction in a well defined physical experiment, the interaction is well suited for numerical studies based on advanced detailed models of all the processes involved. This paper presents results of such a numerical experiment. Cooling of the superheated steam can be applied in order to decrease the temperature of superheated steam in power plants. By spraying the cooling water into the superheated steam, the temperature of the superheated steam can be controlled. In this work, water spray cooling was modeled to investigate the influences of the droplet size, injected velocity, the pressure and velocity of the superheated steam on the evaporation of the cooling water. The results show that by increasing the diameter of the droplets, the pressure and velocity of the superheated steam, the amount of evaporation of cooling water increases. .

Small-scale tunnel test for blast performance

International Nuclear Information System (INIS)

Felts, J E; Lee, R J

2014-01-01

The data reported here provide a validation of a small-scale tunnel test as a tool to guide the optimization of new explosives for blast performance in tunnels. The small-scale arrangement consisted of a 2-g booster and 10-g sample mounted at the closed end of a 127 mm diameter by 4.6-m long steel tube with pressure transducers along its length. The three performance characteristics considered were peak pressure, initial energy release, and impulse. The relative performance from five explosives was compared to that from a 1.16-m diameter by 30-m long tunnel that used 2.27-kg samples. The peak pressure values didn't correlate between the tunnels. Partial impulse for the explosives did rank similarly. The initial energy release was determined from a one-dimensional point-source analysis, which nearly tracked with impulse suggesting additional energy released further down the tunnel for some explosives. This test is a viable tool for optimizing compositional variations for blast performance in target scenarios of similar geometry.

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

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.

The fate of injectant coal in blast furnaces: The origin of extractable materials of high molecular mass in blast furnace carryover dusts

Energy Technology Data Exchange (ETDEWEB)

Dong, S.N.; Wu, L.; Paterson, N.; Herod, A.A.; Dugwell, D.R.; Kandiyoti, R. [University of London Imperial College of Science & Technology, London (United Kingdom). Dept. of Chemical Engineering

2005-07-01

The aim of the work was to investigate the fate of injectant coal in blast furnaces and the origin of extractable materials in blast furnace carryover dusts. Two sets of samples including injectant coal and the corresponding carryover dusts from a full sized blast furnace and a pilot scale rig have been examined. The samples were extracted using 1-methyl-2-pyrrolidinone (NMP) solvent and the extracts studied by size exclusion chromatography (SEC). The blast furnace carryover dust extracts contained high molecular weight carbonaceous material, of apparent mass corresponding to 10{sup 7}-10{sup 8} u, by polystyrene calibration. In contrast, the feed coke and char prepared in a wire mesh reactor under high temperature conditions did not give any extractable material. Meanwhile, controlled combustion experiments in a high-pressure wire mesh reactor suggest that the extent of combustion of injectant coal in the blast furnace tuyeres and raceways is limited by time of exposure and very low oxygen concentration. It is thus likely that the extractable, soot-like material in the blast furnace dust originated in tars is released by the injectant coal. Our results suggest that the unburned tars were thermally altered during the upward path within the furnace, giving rise to the formation of heavy molecular weight (soot-like) materials.

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

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

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

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)

Proceedings of the twenty-fourth annual conference on explosives and blasting technique

Energy Technology Data Exchange (ETDEWEB)

1998-01-01

Papers of interest to the coal industry include: death of a coal shovel; deep hole blasting with SMS (site-mix slurry system); trend of bulk explosives in India; bottomhole annular pressure - a theoretical problem with real effects; maximizing rotary blast hole drills; explosive energy concept for drill productivity and higher overall productivity at reduced excavation costs; large diameter presplitting improved through two novel techniques; avoiding tragedy - lessons to be learned from a flyrock fatality; and an economic analysis of cast blasting compared to other stripping alternatives.

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)

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.)

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

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)

CrocoBLAST: Running BLAST efficiently in the age of next-generation sequencing.

Science.gov (United States)

Tristão Ramos, Ravi José; de Azevedo Martins, Allan Cézar; da Silva Delgado, Gabrielle; Ionescu, Crina-Maria; Ürményi, Turán Peter; Silva, Rosane; Koca, Jaroslav

2017-11-15

CrocoBLAST is a tool for dramatically speeding up BLAST+ execution on any computer. Alignments that would take days or weeks with NCBI BLAST+ can be run overnight with CrocoBLAST. Additionally, CrocoBLAST provides features critical for NGS data analysis, including: results identical to those of BLAST+; compatibility with any BLAST+ version; real-time information regarding calculation progress and remaining run time; access to partial alignment results; queueing, pausing, and resuming BLAST+ calculations without information loss. CrocoBLAST is freely available online, with ample documentation (webchem.ncbr.muni.cz/Platform/App/CrocoBLAST). No installation or user registration is required. CrocoBLAST is implemented in C, while the graphical user interface is implemented in Java. CrocoBLAST is supported under Linux and Windows, and can be run under Mac OS X in a Linux virtual machine. jkoca@ceitec.cz. Supplementary data are available at Bioinformatics online. © The Author 2017. Published by Oxford University Press. All rights reserved. For Permissions, please e-mail: journals.permissions@oup.com

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)

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)

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

Three-dimensional modeling of nuclear steam generator

International Nuclear Information System (INIS)

Bogdan, Z.; Afgan, N.

1985-01-01

In this paper mathematical model for steady-state simulation of thermodynamic and hydraulic behaviour of U-tube nuclear steam generator is described. The model predicts three-dimensional distribution of temperatures, pressures, steam qualities and velocities in the steam generator secondary loop. In this analysis homogeneous two phase flow model is utilized. Foe purpose of the computer implementation of the mathematical model, a special flow distribution code NUGEN was developed. Calculations are performed with the input data and geometrical characteristics related to the D-4 (westinghouse) model of U-tube nuclear steam generator built in Krsko, operating under 100% load conditions. Results are shown in diagrams giving spatial distribution of pertinent variables in the secondary loop. (author)

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

Monitoring of Cyclic Steam Stimulation by Inversion of Surface Tilt Measurements

Science.gov (United States)

Maharramov, M.; Zoback, M. D.

2014-12-01

Temperature and pressure changes associated with the cyclic steam simulation (CSS) used in heavy oil production from sands are accompanied by significant deformation. Inversion of geomechanical data may provide a potentially powerful reservoir monitoring tool where geomechanical effects are significant. Induced pore pressure changes can be inverted from measurable surface deformations by solving an inverse problem of poroelasticity. In this work, we apply this approach to estimating pore pressure changes from surface tilt measurements at a heavy oil reservoir undergoing cyclic steam simulation. Steam was injected from November 2007 through January 2008. Surface tilt measurements were collected from 25 surface tilt stations during this period. The injection ran in two overlapping phases: Phase 1 ran from the beginning of the injection though mid-December, and Phase 2 overlapped with Phase 1 and ran through the beginning of January. During Phase 1 steam was injected in the western part of the reservoir, followed by injection in the eastern part in Phase 2. The pore pressure evolution was inverted from daily tilt measurements using regularized constrained least squares fitting, the results are shown on the plot. Estimated induced pore pressure change (color scale), observed daily incremental tilts (green arrows) and modeled daily incremental tilts (red arrows) are shown in three panels corresponding to two and five weeks of injection, and the end of injection period. DGPS measurements available for a single location were used as an additional inversion constraint. The results indicate that the pore pressure increase in the reservoir follows the same pattern as the steam injection, from west to east. This qualitative behaviour is independent of the amount of regularization, indirectly validating our inversion approach. Patches of lower pressure appear to be stable with regard to regularization and may provide valuable insight into the efficiency of steam injection

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

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)

Thermal circuit and supercritical steam generator of the BGR-300 nuclear power plant

International Nuclear Information System (INIS)

Afanas'ev, B.P.; Godik, I.B.; Komarov, N.F.; Kurochnkin, Yu.P.

1979-01-01

Secondary coolant circuit and a steam generator for supercritical steam parameters of the BGR-300 reactor plant are described. The BGR-300 plant with a 300 MW(e) high-temperature gas-cooled fast reactor is developed as a pilot commercial plant. It is shown that the use of a supercritical pressure steam increases the thermal efficiency of the plant and descreases thermal releases to the environment, permits to use home-made commercial turbine plants of large unit power. The proposed supercritical pressure steam generator has considerable advantages from the viewpoint of heat transfer and hydrodynamical processes

A fast response miniature probe for wet steam flow field measurements

International Nuclear Information System (INIS)

Bosdas, Ilias; Mansour, Michel; Abhari, Reza S; Kalfas, Anestis I

2016-01-01

Modern steam turbines require operational flexibility due to renewable energies’ increasing share of the electrical grid. Additionally, the continuous increase in energy demand necessitates efficient design of the steam turbines as well as power output augmentation. The long turbine rotor blades at the machines’ last stages are prone to mechanical vibrations and as a consequence time-resolved experimental data under wet steam conditions are essential for the development of large-scale low-pressure steam turbines. This paper presents a novel fast response miniature heated probe for unsteady wet steam flow field measurements. The probe has a tip diameter of 2.5 mm, and a miniature heater cartridge ensures uncontaminated pressure taps from condensed water. The probe is capable of providing the unsteady flow angles, total and static pressure as well as the flow Mach number. The operating principle and calibration procedure are described in the current work and a detailed uncertainty analysis demonstrates the capability of the new probe to perform accurate flow field measurements under wet steam conditions. In order to exclude any data possibly corrupted by droplets’ impact or evaporation from the heating process, a filtering algorithm was developed and implemented in the post-processing phase of the measured data. In the last part of this paper the probe is used in an experimental steam turbine test facility and measurements are conducted at the inlet and exit of the last stage with an average wetness mass fraction of 8.0%. (paper)

Application and Development of an Environmentally Friendly Blast Hole Plug for Underground Coal Mines

Directory of Open Access Journals (Sweden)

Donghui Yang

2018-01-01

Full Text Available Drilling and blasting technology is one of the main methods for pressure relief in deep mining. The traditional method for blasting hole blockage with clay stemming has many problems, which include a large volume of transportation, excess loading time, and high labor intensity. An environmentally friendly blast hole plug was designed and developed. This method is cheap, closely blocks the hole, is quickly loaded, and is convenient for transportation. The impact test on the plug was carried out using an improved split Hopkinson pressure bar test system, and the industrial test was carried out in underground tunnel of coal mine. The tests results showed that, compared with clay stemming, the new method proposed in this paper could prolong the action time of the detonation gas, prevent premature detonation gas emissions, reduce the unit consumption of explosives, improve the utilization ratio, reduce the labor intensity of workers, and improve the effect of rock blasting with low cost of rock breaking.

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)

Study on the characteristics of the supersonic steam injector

International Nuclear Information System (INIS)

Abe, Yutaka; Shibayama, Shunsuke

2014-01-01

Steam injector is a passive jet pump which operates without power source or rotating machinery and it has high heat transfer performance due to the direct-contact condensation of supersonic steam flow onto subcooled water jet. It has been considered to be applied to the passive safety system for the next-generation nuclear power plants. The objective of the present study is to clarify operating mechanisms of the steam injector and to determine the operating ranges. In this study, temperature and velocity distribution in the mixing nozzle as well as flow directional pressure distribution were measured. In addition, flow structure in whole of the injector was observed with high-speed video camera. It was confirmed that there were unsteady interfacial behavior in mixing nozzle which enhanced heat transfer between steam flow and water jet with calculation of heat transfer coefficient. Discharge pressure at diffuser was also estimated with a one-dimensional model proposed previously. Furthermore, it was clarified that steam flow did not condense completely in mixing nozzle and it was two-phase flow in throat and diffuser, which seemed to induce shock wave. From those results, several discussions and suggestions to develop a physical model which predicts the steam injectors operating characteristics are described in this paper

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)

Estimation of Temperature Influence on Creep Rate of High-Temperature Elements in Steam Turbines and Steam Pipelines

Directory of Open Access Journals (Sweden)

A. G. Gerasimova

2011-01-01

Full Text Available The paper considers a high temperature influence on strength characteristics of steam pipelines and steam turbine parts of high and medium pressure. The charts showing a decisive temperature importance in diffuse creep have been presented in the paper. The paper contains a calculation of steel self-diffusion coefficient. Dependence Dsd = f(t for more accurate assessment of  resource characteristics of the applied steel has been proposed in the paper.

The temperature control and water quality regulation for steam generator secondary side hydrostatic test

International Nuclear Information System (INIS)

Xiao Bo; Liu Dongyong

2014-01-01

The secondary side hydrostatic test for the steam generator of M310 unit is to verify the pressure tightness of steam generator secondary side tube sheet and related systems. As for the importance of the steam generator, the water temperature and water quality of hydrostatic test has strict requirements. The discussion on the water temperature control and water quality regulation for the secondary loop hydrostatic test of Fuqing Unit 1 contribute greatly to the guiding work for the preparation of the steam generator pressure test for M310 unit. (authors)

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

Thermal-Hydraulic Analysis of a Once-Through Steam Generator Considering Performance Degradation

International Nuclear Information System (INIS)

Han, Hun Sik; Kang, Han Ok; Yoon, Ju Hyeon; Kim, Young In; Song, Jae Seung; Kim, Keung Koo

2016-01-01

Several countries have entered into a global race for the commercialization of SMRs, and considerable research and development have been implemented. Among the various reactor designs, many SMRs have adopted an integral type pressurized water reactor (PWR) to enhance the nuclear safety and system reliability. In the integral reactor design, a single reactor pressure vessel contains primary system components such as fuel and core, steam generators, pumps, and a pressurizer. For the component integration into a reactor vessel, it is important to design each component as small as possible. Thus, it is a common practice to employ a once-through steam generator in the integral reactor design due to its advantages in compactness. In general, gradual degradation in thermal-hydraulic performance of the steam generator occurs with time, and it changes slowly the operating point of the steam generator during plant lifetime. Numerical solutions are acquired to evaluate the thermal-hydraulic performance of the steam generator at various AUFs. The design results obtained show that the average tube length of the steam generator is augmented with the increase of design margin to compensate for the design uncertainties and heat transfer area reduction by plugging, fouling, etc. A helically coiled tube once-through steam generator with 30% design margin is considered for comparison of thermal-hydraulic performances according to the degradation rate

Thermal-Hydraulic Analysis of a Once-Through Steam Generator Considering Performance Degradation

Energy Technology Data Exchange (ETDEWEB)

Han, Hun Sik; Kang, Han Ok; Yoon, Ju Hyeon; Kim, Young In; Song, Jae Seung; Kim, Keung Koo [KAERI, Daejeon (Korea, Republic of)

2016-05-15

Several countries have entered into a global race for the commercialization of SMRs, and considerable research and development have been implemented. Among the various reactor designs, many SMRs have adopted an integral type pressurized water reactor (PWR) to enhance the nuclear safety and system reliability. In the integral reactor design, a single reactor pressure vessel contains primary system components such as fuel and core, steam generators, pumps, and a pressurizer. For the component integration into a reactor vessel, it is important to design each component as small as possible. Thus, it is a common practice to employ a once-through steam generator in the integral reactor design due to its advantages in compactness. In general, gradual degradation in thermal-hydraulic performance of the steam generator occurs with time, and it changes slowly the operating point of the steam generator during plant lifetime. Numerical solutions are acquired to evaluate the thermal-hydraulic performance of the steam generator at various AUFs. The design results obtained show that the average tube length of the steam generator is augmented with the increase of design margin to compensate for the design uncertainties and heat transfer area reduction by plugging, fouling, etc. A helically coiled tube once-through steam generator with 30% design margin is considered for comparison of thermal-hydraulic performances according to the degradation rate.

Dynamics and stability of relativistic gamma-ray-bursts blast waves

Science.gov (United States)

Meliani, Z.; Keppens, R.

2010-09-01

Aims: In gamma-ray-bursts (GRBs), ultra-relativistic blast waves are ejected into the circumburst medium. We analyse in unprecedented detail the deceleration of a self-similar Blandford-McKee blast wave from a Lorentz factor 25 to the nonrelativistic Sedov phase. Our goal is to determine the stability properties of its frontal shock. Methods: We carried out a grid-adaptive relativistic 2D hydro-simulation at extreme resolving power, following the GRB jet during the entire afterglow phase. We investigate the effect of the finite initial jet opening angle on the deceleration of the blast wave, and identify the growth of various instabilities throughout the coasting shock front. Results: We find that during the relativistic phase, the blast wave is subject to pressure-ram pressure instabilities that ripple and fragment the frontal shock. These instabilities manifest themselves in the ultra-relativistic phase alone, remain in full agreement with causality arguments, and decay slowly to finally disappear in the near-Newtonian phase as the shell Lorentz factor drops below 3. From then on, the compression rate decreases to levels predicted to be stable by a linear analysis of the Sedov phase. Our simulations confirm previous findings that the shell also spreads laterally because a rarefaction wave slowly propagates to the jet axis, inducing a clear shell deformation from its initial spherical shape. The blast front becomes meridionally stratified, with decreasing speed from axis to jet edge. In the wings of the jetted flow, Kelvin-Helmholtz instabilities occur, which are of negligible importance from the energetic viewpoint. Conclusions: Relativistic blast waves are subject to hydrodynamical instabilities that can significantly affect their deceleration properties. Future work will quantify their effect on the afterglow light curves.

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

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

Local properties of countercurrent stratified steam-water flow

International Nuclear Information System (INIS)

Kim, H.J.

1985-10-01

A study of steam condensation in countercurrent stratified flow of steam and subcooled water has been carried out in a rectangular channel/flat plate geometry over a wide range of inclination angles (4 0 -87 0 ) at several aspect ratios. Variables were inlet water and steam flow rates, and inlet water temperature. Local condensation rates and pressure gradients were measured, and local condensation heat transfer coefficients and interfacial shear stress were calculated. Contact probe traverses of the surface waves were made, which allowed a statistical analysis of the wave properties. The local condensation Nusselt number was correlated in terms of local water and steam Reynolds or Froude numbers, as well as the liquid Prandtl number. A turbulence-centered model developed by Theofanous, et al. principally for gas absorption in several geometries, was modified. A correlation for the interfacial shear stress and the pressure gradient agreed with measured values. Mean water layer thicknesses were calculated. Interfacial wave parameters, such as the mean water layer thickness, liquid fraction probability distribution, wave amplitude and wave frequency, are analyzed

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

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)

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)

Indian Point 2 steam generator tube rupture analyses

International Nuclear Information System (INIS)

Dayan, A.

1985-01-01

Analyses were conducted with RETRAN-02 to study consequences of steam generator tube rupture (SGTR) events. The Indian Point, Unit 2, power plant (IP2, PWR) was modeled as a two asymmetric loops, consisting of 27 volumes and 37 junctions. The break section was modeled once, conservatively, as a 150% flow area opening at the wall of the steam generator cold leg plenum, and once as a 200% double-ended tube break. Results revealed 60% overprediction of breakflow rates by the traditional conservative model. Two SGTR transients were studied, one with low-pressure reactor trip and one with an earlier reactor trip via over temperature ΔT. The former is more typical to a plant with low reactor average temperature such as IP2. Transient analyses for a single tube break event over 500 seconds indicated continued primary subcooling and no need for steam line pressure relief. In addition, SGTR transients with reactor trip while the pressurizer still contains water were found to favorably reduce depressurization rates. Comparison of the conservative results with independent LOFTRAN predictions showed good agreement

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

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)

An experimental investigation of the isochoric heat capacity of superheated steam and mixtures of superheated steam and hydrogen gas

International Nuclear Information System (INIS)

Nowak, E.S.; Chan, J.S.

1975-01-01

Measurements on the specific heat at constant volume of superheated steam and hydrogen gas mixtures at concentrations varying from 1.6 to 0.8 moles of water vapor per mole of hydrogen gas were made for temperatures ranging from 240 to 400 deg C. It was found that the experimental specific heat values of the mixtures are in good agreement with the ideal mixture values only near the saturation temperature of steam. The difference between the measured and the calculated ideal mixture values is a function of temperature, pressure and composition varying from about 11 to 24% at conditions far removed from the saturation temperature of steam. This indicates the heat of mixing is of significance in the steam-hydrogen system

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).

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.)

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

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

A GASFLOW analysis of a steam explosion accident in a typical light-water reactor confinement building

International Nuclear Information System (INIS)

Travis, J.R.; Wilson, T.L.; Spore, J.W.; Lam, K.L.; Rao, D.V.

1994-01-01

Steam over-pressurization resulting from ex-vessel steam explosion (fuel-coolant interaction) may pose a serious challenge to the integrity of a typical light-water reactor confinement building. If the steam generation rate exceeds the removal capacity of the Airborne Activity Confinement System, confinement over pressurization occurs. Thus, there is a large potential for an uncontrolled and unfiltered release of fission products from the confinement atmosphere to the environment at the time of the steam explosion. The GASFLOW computer code was used to analyze the effects of a hypothetical steam explosion and the transport of steam and hydrogen throughout a typical light-water reactor confinement building. The effects of rapid pressurization and the resulting forces on the internal structures and the heat exchanger service bay hatch covers were calculated. Pressurization of the ventilation system and the potential damage to the ventilation fans and high-efficiency particulate air filters were assessed. Because of buoyancy forces and the calculated confinement velocity field, the hydrogen diffuses and mixes in the confinement atmosphere but tends to be transported to its upper region

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.

Utilization of steam treated agricultural by -product as ruminant feed

International Nuclear Information System (INIS)

Naeem, M.; Rajput, N.; Lili, Z.; Su, Z.; Rui, Y.; Tian, W.

2014-01-01

Shortage of animal food is a burning issue of the recent time, whereas the agricultural by -products are readily available to be used as ruminants feed. However, the low protein and digestibility are the hindrances in utilization of these low quality crop residues as a feed. In order to utilize rice straw as animal feed this study was conducted to investigate the influence of steam explosion treatment on its composition and in vitro degradability. The samples, I (untreated rice straw), II (rice straw exposed to 15.5 kgf/cm/ sub2/ steam pressure for 90 sec) and III (rice straw exposed to 15.5 kgf/cm/sup 2/steam pressure for 120 sec) were prepared. The results revealed that the crude protein (CP), ether extract (EE) and acid detergent lignin (ADL) contents of rice straw were improved after treatment with steam explosion in time dependent manner (P<0.05). The neutral detergent fiber (NDF), acid detergent fiber (ADF) and organic matter (OM) were higher, while dry matter (DM) and ash contents were lower (P<0.05) in II as compared to group I and III; however, after increasing the time at same pressure these parameters decreased. Furthermore, group III showed higher concentration of propionate, acetate, butyrate, and total VFA (P<0.05). While, group I exhibited higher concentration of iso-butyrate and iso-valerate (P<0.05). The concentration of valeric acid and acetate to propionate ratio were not affected by steam explosion treatment. Moreover, group III showed the higher in vitro DM degradability, OM degradability, DNDF and gas production (P<0.05); while, lower DADF and pH (P<0.05) compared with groups I and II. These findings suggest that the steam explosion treatment at 15.5 kgf/cm/sup 2/ pressure for 120 sec, may be used to enhance the nutritive value and digestibility of rice straw. (author)

Analysis of induced steam generator tube rupture using MAAP 4.0

International Nuclear Information System (INIS)

Kenton, M.; Epstein, M.; Henry, R.E.; Paik, C.; Fuller, E.

1996-01-01

The nuclear industry has initiated a program of Steam Generator Degradation Specific Management (SGDSM) to cope with the various types of corrosion that have been observed in pressurized water reactor (PWR) steam generators. In parallel, the U.S. Nuclear Regulatory Commission is promulgating revised rules on steam generator tube integrity. To support these efforts, the Electric Power Research Institute has sponsored calculations with the MAAP 4 code. The principal objective of these calculations is to estimate the peak temperatures experienced by the steam generator tubes during high-pressure severe accidents. These results are used to evaluate the potential for degraded tubes to leak or rupture. Attention was focused on station blackout (SBO) accidents with loss of turbine-driven auxiliary feedwater because these generally result in the greatest threat to the tubes

Pressure suppression device

International Nuclear Information System (INIS)

Mizumachi, Wataru; Fukuda, Akira; Kitaguchi, Hidemi; Shimizu, Toshiaki.

1976-01-01

Object: To relieve and absorb impact wave vibrations caused by steam and non-condensed gases releasing into the pressure suppression chamber at the time of an accident. Structure: The reactor container is filled with inert gases. A safety valve attached main steam pipe is provided to permit the excessive steam to escape, the valve being communicated with the pressure suppression chamber through an exhaust pipe. In the pressure suppression chamber, a doughnut-like cylindrical outer wall is filled at its bottom with pool water to condense the high temperature vapor released through the exhaust pipe. A head portion of a vent tube which leads the exhaust pipe is positioned at the top, and a down comer and an exhaust vent tube are locked by means of steady rests. At the bottom is mounted a pressure adsorber device which adsorbs a pressure from the pool water. (Kamimura, M.)

Investigation of low pressure ES-SAGD

Energy Technology Data Exchange (ETDEWEB)

Ivory, J.; Zheng, R.; Nasr, T.; Deng, X.; Beaulieu, G.; Heck, G. [Alberta Research Council, Edmonton, AB (Canada)

2008-10-15

This paper described a scaled model experiment conducted to investigate the effectiveness of expanding solvent steam assisted gravity drainage (ES-SAGD) processes at low pressures. Lower SAGD pressures typically result in reduced oil production as a result of correspondingly lower steam temperatures. However, lower pressures may also result in a reduced steam to oil ratio (SOR) and a higher vaporization heat. Steam was injected into an injection well at 33 cm{sup 3} per minute and in a production well at 31 cm{sup 3} per minute. Steam and solvents were then co-injected into the injection well at a temperature of 206 degrees C. The experiment was history-matched and a parametric analysis was conducted using a simulation tool. The 2-D and 3-D field-scale simulations investigated the impact of operating pressures, injection rates; sub-cool; oil and gas phase diffusion and dispersion; live oil versus dead oil performance; and the use of drawdown when oil rates declined. Low pressure ES-SAGD was then compared with low-pressure SAGD. Results of the study suggested that production pressures, sub-cool and solvent concentrations are important parameters in ES-SAGD processes. At 1500 kPa production pressure and 10 degrees C sub-cool, the co-injection of solvent with steam increased average oil rates by 15 per cent more than the SAGD process. SOR was also reduced. 6 refs., 8 tabs., 20 figs.

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)

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

Pressure control device in a BWR type reactor

International Nuclear Information System (INIS)

Nagata, Yoshifumi.

1983-01-01

Purpose: To perform an adequate pressure control with no erroneous scram operation even when the balance of pressure is lost between main steam pipelines. Constitution: Pressure detectors are disposed respectively to a plurality of main steam pipelines and pressure detection values therefrom are inputted into a higher value preference circuit to select a higher value. The deviation between the higher pressure value signal and an aimed value is calculated in an addition circuit and the calculated deviation is inputted to a succeeding higher value preference circuit by way of a servo mechanism as an output from an electronic main steam pressure controller. The above output and the output from another mechanical main steam pressure controller are compared in this circuit to issue a higher value signal to a governer to control the degree of a steam control valve by way of the governor and the servo mechanism. The deviation hereinafter is converged through the same procedures into an aimed predetermined value. (Sekiya, K.)

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)

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.

Numerical analysis of water hammer induced by injection of subcooled water into steam flow in a horizontal pipe

International Nuclear Information System (INIS)

Minato, Akihiko; Nagoyoshi, Takuji; Nakamura, Akira; Fujii, Yuzo; Aya, Izuo; Yamane, Kenji

2004-01-01

Subcooled water injection into steam flow in piping systems may generate a water column containing a large steam slug. The steam slug collapses due to rapid condensation and interfaces on both sides collides with each other. Water hammer takes place and sharp pressure pulse propagates through the pipe. The purpose of this study is to show capability of the present numerical simulation method for predictions of pressure transient and loads on a piping system following steam slug collapse. A three-dimensional computer code for transient gas-liquid two-phase flow was applied to simulate an experiment of steam-condensation-induced water hammer with a horizontal polycarbonate pipe. The code was based on the extended two-fluid model, which treated interface motion using the VOF (Volume of Fluid) technique. The Godunov scheme of highly compressible single-phase flow was modified for application to the Riemann problem solution of gas-liquid mixture. Analysis of local steam slug collapse resulted in comparable peak pressure and pulse width of pressure transients with the observation. The calculation of pressure pulse propagation and impact load on piping system showed the quasi-steady pressure load was imposed especially on elbow at 1/10 of water hammer peak pressure. (author)

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.

Model studies of the vertical steam generator thermal-hydraulic characteristics

International Nuclear Information System (INIS)

Desyatun, V.F.; Moskvichev, V.F.; Ulasov, V.M.; Morozov, V.G.; Burkov, V.K.; Grebennikov, V.N.

1984-01-01

Results of investigations conducted to clarify the calculation technique and to test the workability of the main elements and units of the PGV-250 vertical steam generator of saturated steam are considered. The steam generating capacity of the plant is 1486 t/h, thermal power is 792 MW. Steam generation follows a multiple circulation scheme. The heat surface comprises 330-shields. The investigations are carried out with a model which reproduces all the main elements of the steam generator xcluding the economizer section. The flow rates of feed water, generated steam and coolant of the first circuit as well as temperature, pressure and humidity of the generated steam past the separator are determined. The average heat transfer factors of the heat surface are calculated on the base of the data obtained and a conclusion is drawn on the correctness of the thermohydraulic calculation technique used in development of the PGV-250 steam generator design. Temperature pulsations and heat surface steaming are not observed. The steam humidity at the outlet and steam capture into sinking tubes are within permissible values

Constructal design of a blast furnace iron-making process based on multi-objective optimization

International Nuclear Information System (INIS)

Liu, Xiong; Chen, Lingen; Feng, Huijun; Qin, Xiaoyong; Sun, Fengrui

2016-01-01

For the fixed total raw material cost and based on constructal theory and finite time thermodynamics, a BFIM (blast furnace iron-making) process is optimized by taking a complex function as optimization objective. The complex function is integrated with HM (hot metal) yield and useful energy of the BF (blast furnace). The optimal cost distribution of raw materials (namely “generalized optimal construct”) is obtained. The effects of some parameters, such as oxygen enrichment, blast temperature and pulverized coal dosage, on the optimization results are analyzed. The results show that the HM yield, useful energy and complex function are, respectively, increased by 3.13%, 2.66% and 2.90% after generalized constructal optimization. The utilization efficiencies of the BFG (blast furnace gas) and slag are 41.3% and 57.1%, respectively, which means that the utilization potentials of the BFG and slag can be further exploited. Increasing pulverized coal dosage and decreasing the agglomerate ratio can increase the complex function. The performance the BFIM process can be improved by adjusting the oxygen enrichment, blast temperature, blast dosage, pressure ratio of the Brayton cycle's air compressor and relative pressure drop of the air compressor inlet to their optimal values, respectively, which are new findings of this paper. - Highlights: • Constructal optimization of a blast furnace iron-making process is performed. • Finite time thermodynamic model of open Brayton cycle is adopted. • Weighting function is taken as optimization objective. • Optimal cost distribution of the raw materials is obtained.

Design of large steam turbines for PWR power stations

International Nuclear Information System (INIS)

Hobson, G.; Muscroft, J.

1983-01-01

The thermodynamic cycle requirements for use with pressurized water reactors are reviewed and the manner in which thermal efficiency is maximised is outlined. The special nature of the wet steam cycle associated with turbines for this type of reactor is discussed. Machine and cycle parameters are optimised to achieve high thermal efficiency, particular attention being given to arrangements for water separation and steam reheating and to provisions for feedwater heating. Principles and details of mechanical design are considered for a range of both full-speed turbines running at 3000 rpm on 50 Hz systems and half-speed turbines running at 1800 rpm on 60 Hz systems. The importance of service experience with nuclear wet steam turbines and its relevance to the design of modern turbines for pressurized water reactor applications is discussed. (author)

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.

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

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

Liquid abrasive grit blasting literature search and decontamination scoping tests report

International Nuclear Information System (INIS)

Ferguson, R.L.

1993-10-01

Past decontamination and solvent recovery activities at the Idaho Chemical Processing Plant (ICPP) have resulted in the accumulation of 1.5 million gallons of radioactively contaminated sodium-bearing liquid waste. Future decontamination activities at the ICPP could result in the production of 5 million gallons or more of sodium-bearing waste using the current decontamination techniques of chemical/water flushes and steam jet cleaning. With the curtailment of reprocessing at the ICPP, the focus of decontamination is shifting from maintenance for continued operation of the facilities to decommissioning. As decommissioning plans are developed, new decontamination methods must be used which result in higher decontamination factors and generate lower amounts of sodium-bearing secondary waste. The primary initiative of the WINCO Decontamination Development Program is the development of methods to eliminate/minimize the use of sodium-bearing decontamination chemicals. One method that was chosen for cold scoping studies during FY-93 was abrasive grit blasting. Abrasive grit blasting has been used in many industries and a vast amount of research and development has already been conducted. However, new grits, process improvements and ICPP applicability was investigated. This evaluation report is a summary of the research efforts and scoping tests using the liquid abrasive grit blasting decontamination technique. The purpose of these scoping tests was to determine the effectiveness of three different abrasive grits: plastic beads, glass beads and alumina oxide

ORCENT-2, Full Load Steam Turbine Cycle Thermodynamics for LWR Power Plant

International Nuclear Information System (INIS)

Fuller, L.C.

1979-01-01

1 - Description of problem or function: ORCENT-2 performs heat and mass balance calculations at valves-wide-open design conditions, maximum guaranteed rating conditions, and an approximation of part-load conditions for steam turbine cycles supplied with throttle steam, characteristic of contemporary light-water reactors. The program handles both condensing and back-pressure turbine exhaust arrangements. Turbine performance calculations are based on the General Electric Company method for 1800-rpm large steam turbine- generators operating with light-water-cooled nuclear reactors. Output includes all information normally shown on a turbine-cycle heat balance diagram. 2 - Method of solution: The turbine performance calculations follow the procedures outlined in General Electric report GET-6020. ORCENT-2 utilizes the 1967 American Society of Mechanical Engineers (ASME) formulations and procedures for calculating the properties of steam, adapted for ORNL use by D.W. Altom. 3 - Restrictions on the complexity of the problem: Maxima of: 12 feed-water heaters, 5 moisture removal stages in the low-pressure turbine section. ORCENT-2 is limited to 1800-rpm tandem-compound turbine-generators with single- or double-flow high pressure sections and one, two, or three double-flow low-pressure turbine sections. Steam supply for LWR cycles should be between 900 and 1100 psia and slightly wet to 100 degrees F of initial superheat. Generator rating should be greater than 100 MVA

Steam CFD simulation of injection in suppression pool

International Nuclear Information System (INIS)

Naveen Samad, A.M.; Ghosh, Sumana

2015-01-01

Boiling water reactor (BWR) is one of the common types of electricity generating nuclear reactor. Suppression pool system is a major component of the BWR which has to be designed efficiently for the safe operations. During some accidents like Loss of Coolant Accident (LOCA) large amount of steam are injected to the pressure suppression system resulting in increase in temperature of the pool and thereby increasing the pressure. The present work discuss about the Computational Fluid Dynamics (CFD) simulation of steam injected to the wet well of BWR through the blow down pipes and there by investigating the hydrodynamic and thermal characteristics of the system. The simulations were carried out for three different steam injection velocities. The numerical simulations were performed with ANSYS FLUENT using multiphase 3D Volume of Fluid (VOF) model and k-ε model was adopted for modelling turbulence flow. (author)

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

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

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

Investigation of the effect of pressure increasing in condensing heat-exchanger