International Nuclear Information System (INIS)

Short communication.

Science.gov (United States)

ABS>A horizontal steam generator design for the SM-2 was lysis to determine the per formance of such a steam generator under steady state operating conditions and during load transients, The configuration for this design is a two- drum unit consisting of a heat exchanger unit and separator drum interconnected by integral riser and downcomer. An analog computer was used to analyze the steam generator behavior Wring load transients. The effect of various design changes on the response of the steam generator to step chages in load was determined. The horizontal steam generator design was compared to the existing vertical steam generator design for weight, size, price, and performance. (auth)

Science.gov (United States)

... Title : RESULTS ON INCOLOY 800 AND ALLIED STEAM GENERATOR MATERIALS IN FLORIDA FIELD CORROSION TESTS,. ...

Science.gov (United States)

Analytical relations are obtained for estimating how the distributions of temperature and heat flux vary along a steam-generating tube and how the steam-generator power output reduces due to formation and accumulation of deposits.

British Library Electronic Table of Contents (United Kingdom)

Statistical data on experience gained from operation of steam generators around the world are presented, problems arising in vertical and horizontal steam generators are described, and the conditions of heattransfer tubes used in them are compared.

British Library Electronic Table of Contents (United Kingdom)

Results are given of numerical simulation of three-dimensional pattern of flow of a two-phase steam-water mixture in the house of a PGV-1000 horizontal steam generator obtained using the BAGIRA best-estimate thermohydrodynamic computer codes. The space distributions of velocities and local void fractions in the steam generator housing for different modes of operation of power-generating unit are calculated and compared with available experimental data.

International Nuclear Information System (INIS)

Russian 2004 p. 50-51 Russian Federation Kalyakin, SG Kapinos,

Energy Technology Data Exchange (ETDEWEB)

The thermal-hydraulic tests for a horizontal steam generator of a next-generation PWR (New PWR-21) were performed. The purpose of these tests is to understand the thermal-hydraulic behavior in the secondary side of horizontal steam generator during the plant normal operation. A test was carried out with cross section slice model simulated the straight tube region. In this paper, the results of the test is reported, and the effect of the horizontal steam generator internals on the thermal-hydraulic behavior of the secondary side and the circulation characteristics of the secondary side are discussed. (author)

International Nuclear Information System (INIS)

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)

International Nuclear Information System (INIS)

... p. SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS algorithms

Energy Technology Data Exchange (ETDEWEB)

The presentation deals with the flows on the primary side of a horizontal steam generator under conditions typical to natural circulation cooling of the reactor. The main goal is to analyse the effect of primary flow patterns on the heat transfer capability of the steam generator. Conclusions pertinent to steam generator modelling with system codes are also drawn. (10 refs., 9 figs., 4 tabs.).

International Nuclear Information System (INIS)

Peculiarities of design of the PGV-1000 horizontal steam generator affecting soluble impurity distribution in its water volume are considered in brief. The results of estimating sodium distribution in different zones of the steam generator are presented. The conclusion is made on the necessity of arrangement of representative measurements of sodium and chloride content in water volume of the steam generator, particularly, in the hot bottom zone for optimization of blow-through flowsheet and its regulations.

International Nuclear Information System (INIS)

The Steam Generator (SG) tubing degradation caused by corrosion and other age-related mechanisms continues to be a significant safety and cost concern for many Nuclear Power Plants (NPP). The understanding of the steam generator ageing mechanisms is the key to effective management of steam generator ageing and consists of the knowledge of steam generator materials and these one properties, stressors and operating conditions, like degradation sites and wear mechanisms. The principal types of corrosion are presented which can occur in CANDU steam generator. There are also presented the operation conditions, the specifications referring to the water chemistry and the construction materials of Steam Generator, the factors that have a great ...

International Nuclear Information System (INIS)

The invention concerns a procedure for operating reactors in nuclear power plants. It aims at utilizing power reserves in nuclear power plants. This can be achieved by a steam-side connection of the steam generators of two reactors. The amount of steam exchanged between the units is chosen in such a way that power changes at the steam turbines feedback mainly to the corresponding reactor. In order to realize a high power transfer it is necessary to return the amount of condensate produced in the steam receiving unit and corresponding to the power transferred to the feedwater system of the steam donating unit.

International Nuclear Information System (INIS)

A calculation model was developed for the analysis of thermal and hydraulic processes in the PGV-1000 horizontal steam generator. The model makes it possible to examine the hydraulics of the primary medium, i.e., the distribution of flow velocities and mass flows in the exchanger tube system and hydraulic losses of the primary medium in the steam generator during its flow between the two connecting sites to the main circulation pipeline, spatial distribution of heat fluxes between the primary and secondary sides of the steam generator and the total transmitted thermal power, pressure on the secondary side of the generator, natural circulation of the working medium on the secondary side, and the mean circulation number, spatial distribution of the volume fraction of the steam component in the intertubular space, effect of ...

International Nuclear Information System (INIS)

The water and steam circuits of steam generators in pressurized-water nuclear power plants are described together with the mechanism of denting, and the corrosion of spacer plates that leads to cracks in tubes by constriction. The different chemical specifications applicable to the water of the secondary circuit of the generators in normal operation and on first commissioning are listed. The results obtained and the measurements of chemical values taken in operation on the water in the secondary circuits of steam generators at Fessenheim and Bugey are presented.

International Nuclear Information System (INIS)

This is a manuscript for a lecture contents: 1) Steam power and fundamentals of the steam power process, 3) conventional, nuclear and other steam generation processes, 4) cooling systems for steam power plants, 5) gas turbine power plants and combined-cycle power plants, 6) cogeneration, 7) development of thermal power plants and environmental effects. (GL).

International Nuclear Information System (INIS)

Current light water reactor (LWR) steam generators have been affected by a variety of corrosion and mechanical damage degradation mechanisms. Included are wear caused by tube vibration, intergranular corrosion, pitting, and thinning or wastage of the steam generator tubing and accelerated corrosion of carbon steel supports (denting). The Electric Power Research Institute (EPRI) and the Steam Generator Owners Groups (I, II) have sponsored laboratory and field studies to provide ameliorative actions for the majority of the damage forms experienced to date. Some of the current corrosion mechanisms are aggravated or caused by unique materials choices or materials interactions. New materials have been proposed and at least partially qualified for use in replacement model steam generators, including an advanced LWR design. In ...

International Nuclear Information System (INIS)

The manual presents the fundamentals of thermodynamics and fluid mechanics, the main components of steam power plants, and the power generation process. The following concepts and subjects are discussed at length: steam generator; steam turbines; turbogenerators; condensers; cooling technology; water/steam cycle and water treatment; design data of fossil-fuelled power plants; design and optimisation of nuclear power plant thermodynamics; pipelines and fittings; control systems in steam power plants; connection to the electricity grid and self-supply of thermal power plants; power plant transformer concepts and definitions. (HAG).

International Nuclear Information System (INIS)

The 3D modeling of the thermal hydraulic processes and soluble impurity distribution in the horizontal steam generator PGV-1000 was fulfilled with the thermal hydraulic code STEG. Steady-state operation of horizontal seam generator PGV-1000 was analyzed at nominal power. The modeling of the soluble impurity distribution was fulfilled on the basis of the previous thermal hydraulic modeling results. The processes of the soluble impurity deposition on the steam generator tube bundles and deposits outwash were considered in the mathematical model of the code. The modeling was fulfilled for horizontal steam generators with different peculiarities in construction. Calculation results were compared with experimental results obtained at NPP. The agreement between calculated and experimental results is quite reasonable. Results of modeling are ...

International Nuclear Information System (INIS)

Large scale steam storage power generation, one of the new energy storage systems for the future of inflexible electric power sources consisting of nuclear and coal power plants has been studied on the subjects of the systems to be attached to coal and nuclear power units, of the definition of storage efficiency and of the vertical steam storage vessel technology. Steam storage power generation may be hopeful for its higher efficiency similarly defined as of pumped storage plants while high temperature heat storage and the internal structure of large vertical steam storage vessel (accumulator) need to be developed. (author).

Science.gov (United States)

The Second International Seminar on Horizontal Steam Generator Modelling was arranged to continue the international cooperation that was started during the first seminar in March 1991. The main topics of the seminar were: (1) further experimental results ...

Energy Technology Data Exchange (ETDEWEB)

A model was developed to compute concentration of impurities in the WWER 440 steam generator (SG) secondary water along the tube bundle. Calculated values were verified by concentration values obtained from secondary water sample chemical analysis. (orig.). 2 refs.

Energy Technology Data Exchange (ETDEWEB)

A submerged nuclear power station has been proposed as an alternative power station with a relatively low environmental impact for use by both industrialized and developing countries. The station would be placed 10 m above the seabed at a depth of 30--100 m and a distance of 10--30 km from shore. The submerged nuclear power station would be manufactured and refueled in a central facility, thus gaining the economies of factoryfabrication and the flexibility of short-lead-time deployment. To minimize the size of the submerged hull, horizontal steam generators are proposed for the primary-to-secondary heat transfer, instead of the more traditional vertical steam generators. The horizontal steam generators for SNPS would be similar in design to the horizontal steam generators used in the N-Reactors except the tube ...

International Nuclear Information System (INIS)

In this paper the authors show the means putting in place for examination of steam generators tubes. These means (eddy current probes, ultrasonic testing) associated with a knowledge on degradation phenomena allow mapping controlled tubes and limiting undesirable obturations.

Science.gov (United States)

In the study three loss-of-feedwater type experiments which were preformed with the PACTEL facility has been calculated with two computer codes. The purpose of the experiments was to gain information about the behaviour of horizontal steam generator in a ...

Energy Technology Data Exchange (ETDEWEB)

Steam generator tube plugging rates have increased markedly at Bruce since the end of the 1980s. A new programme of refurbishment aims to keep the steam generators operating successfully until around 2015. (Author).

International Nuclear Information System (INIS)

In this paper, we report on the analysis of reverse flow in inverted U-tubes of a steam generator under natural circulation condition. The mechanism of reverse flow in inverted U-tubes of the steam generator with natural circulation is graphically analyzed by using the full-range characteristic curve of parallel U-tubes. The mathematical model and numerical calculation method for analyzing the reverse flow in inverted U-tubes of the steam generator with natural circulation have been developed. The reverse flow in an inverted U-tube steam generator of a simulated pressurized water reactor with natural circulation in analyzed. Through the calculation, the mass flow rates of normal and reverse flows in individual U-tubes are obtained. The predicted sharp drop of the fluid temperature in the inlet plenum of the ...

International Nuclear Information System (INIS)

The lateral bearing device is made of 7 lateral supports, each positioned to allow the displacement of the steam generator due to thermal or seismic effects. Each support includes a buffer plate that can be positioned on the steam generator using a position control assembly. This control assembly consists of a screw jack arrangement where the nut is fastened via an energy absorbing layer to a footplate that is fixed to the concrete wall of the steam generator enclosure. 4 figs.

Energy Technology Data Exchange (ETDEWEB)

A method for cleaning timing optimization in a parallel steam generator system is described. The method is based on the minimization of a suitable objective function, and takes into account the load allocation on steam generators. In order to establish appropriate fouling growth models the mechanism of the particle deposition and removal on heat transfer surfaces is analyzed. The objective function is related to the short time management costs which are based on depreciation of steam generators, fuel costs and the costs of cleaning interventions. The optimization problem is described; a direct one level method is compared with a two level method. Some applications and their results are reported and discussed. (author)

International Nuclear Information System (INIS)

The main features are presented of a computer programme for 3-D thermohydraulic and thermodynamic analysis of the PGV-1000 horizontal steam generator used at the Temelin NPP. The programme provides analyses of primary side hydraulics, heat exchange behavior and the steam generator secondary side thermohydraulics and thermodynamics. Given are calculated data on the circulation flow rate, void fraction, heat transfer dynamics and the swelled level. (Z.S.) 9 figs.

International Nuclear Information System (INIS)

This paper dealt with the identification of possible damaging mechanism of the collector of the WWER 440 steam generator, cracking of primary collectors, corrosion damage of the protective coat of the primary collector circumferential weld, cracking of breathing space in the region of blinding effect by corrosion and strain, leaking of disassembling joint of the primary collector lid and with the integrity of heat exchanging tubes.

International Nuclear Information System (INIS)

Design layout of horizontal steam generator (SJ) with a special feedwater heating surface (by a surface water economizer), designated for NPPs with WWER-1000 reactors, is suggested. The design enables to decrease sharply the difference between the temperatures of saturation and feedwater. Blowdown outlet is organized against PG face, which increases the efficiency of flowing. The suggested layout enables to decrease thermal stresses in structural units and PG metal content, as compared to the PGV-1000 steam generator.

Energy Technology Data Exchange (ETDEWEB)

Results of weight measurement of PGV-213 steam generator during filling in, heating-up and power increase are described. Special measurement system based on stress gauges has been developed. Method of derivation of secondary water inventory is described. Comparison of the data for two steam generators prove accuracy of the measurements. (orig.). 1 refs.

Energy Technology Data Exchange (ETDEWEB)

The presentation gives a brief insight on possibility of using the VEERA facility in studying the stratification phenomenon. The idea for such experiments is to use the facility upper plenum part to simulate the conditions in upper part of horizontal steam generator hot collector. The upper part of steam generator hot collector is one of the locations where the stratification can take part during natural circulation mode. 4 refs.

International Nuclear Information System (INIS)

A nonlinear mathematical model is presented of a WWER-440 nuclear power plant horizontal steam generator. On the proposed model is based a computer program for investigating transients in steam generators during loss of coolant accidents. Processes taking place at the primary side of the steam generator are described by a set of partial differential equations while those at the secondary side of the steam generator are described by plain differential equations with the variables being complex time functions. The model takes account of the coolant as both a single- and two-phase medium, of changes in the direction of the primary coolant flow and of changes in the direction of heat transfer. Heat transfer through the wall is based on a simple model of heat transfer through a thin-walled tube and includes a correction for ...

Energy Technology Data Exchange (ETDEWEB)

Since the first operation of nuclear steam generator early 1960s, its performance requirements have been met but the steam generator problems have been met but the steam generator problems have been major cause of reducing the operational reliability, plant safety and availability. U-tube assembly of steam generator forms the primary system pressure boundary of the plant and have experienced several types of tube degradation problems. Tube failure and leakage resulting from the degradation will cause radioactive contamination of secondary system by the primary coolant, and this may lead to unplanned plant outages and costly repair operations such as tube plugging or steam generator replacement. For the case of steam generators for heavy ...

International Nuclear Information System (INIS)

Vertical U-tube steam generators in Pressurized Water Reactors (PWRs) operating an All Volatile Treatment (AVT) secondary chemistry have experienced corrosion problems, particularly denting and sludges. The studies reported evaluate the feasibility of using a low-concentration (0.5 wt%) chemical cleaning process to remove corrosion product deposits from steam generator surfaces and magnetite from tube-to-support plate crevices of PWR steam generators. The process potentially may be applied at schedule intervals, such as during normal refueling outages, to maintain a steam generator in clean operating condition. This report describes the results of testing to evaluate the effectiveness of several chelant acids for dissolving steam generator sludges and crevice magnetite. Corrosion ...

International Nuclear Information System (INIS)

The interrelations between the factors causing the main effects on the primary circuit coolant flow rate distribution in the horizontal steam generator pipes in reactor facilities with the WWER type reactors under the modes with natural circulation are discussed. The criterion showing the presence or absence of coolant circulation reversal in bottom rows of the steam generator pipes is obtained. It is shown that large hydraulic non-uniformity in steam generator pipes operating in parallel under coolant natural circulation leads to decreasing the heat transfer surface efficiency under reactor facility emergency cooling, restricts its servicing capabilities. The circulation reverse in steam generator pipes under coolant natural circulation mode can give unfavourable effect on separate structural elements of the ...

International Nuclear Information System (INIS)

Nuclear Engineering Laboratory of the Technical Research Centre of Finland has widely used RELAP5/MOD2 and -MOD3 codes to simulate horizontal steam generators. Several models have been developed and successfully used in the VVER-safety analysis. Nevertheless, the models developed have included only rather few nodes in the steam generator secondary side. The secondary side has normally been divided into about 10 to 15 nodes. Since the secondary side at the steam generators of VVER-440 type reactors consists of a rather large water pool, these models were only roughly capable to predict secondary side flows. The paper describes an attempt to use RELAP5/MOD3 code to predict secondary side flows in a steam generator of a VVER-440 reactor. A 2D/3D model has been developed using RELAP5/MOD3 codes cross-flow junctions. The ...

International Nuclear Information System (INIS)

The steam generator low water level trip setpoint of KORI NPP units 3 and 4(KNU 3 and 4), three-loop Westinghouse pressurized water reactor, is higher than that of OPR1000. In addition, steam generator downcomer water level in KNU 3 and 4 could fluctuate easily during a transient because of smaller downcomer water inventory, compared to the total water inventory in the steam generator. Due to these reasons, there is a higher possibility of unnecessary reactor trips caused by the steam generator low-low water level in KNU 3 and 4. Its operating history shows that most of reactor trips were caused by steam generator low-low level reactor trip signal. Such reactor trips, especially unnecessary ones, result in time and economic losses. In this paper, a feasibility study was performed ...

International Nuclear Information System (INIS)

Results of calculation of steady thermal-hydraulic characteristics of PVG-440 horizontal steam generator are presented. Steam flows in selected sections are compared to data provided by OKB Gidropress Calculated vapor void fractions are compared to measured ones. (orig.) (3 refs., 3 figs., 8 tabs.).

Energy Technology Data Exchange (ETDEWEB)

The results of the development work to improve the horizontal steam generator modelling using the CATHARE code as well as the results of the steady-state and the steam-line break calculations are presented. Also the results of the steady-state calculations are compared to the measurements performed in operating VVER power plants. (9 refs., 6 figs., 2 tabs.).

British Library Electronic Table of Contents (United Kingdom)

This paper describes the modeling of horizontal steam generator with the TRACE code and calculation results of a loss-of-feedwater (LOF-10) experiment at the PACTEL facility. Parallel Channel Test Loop (PACTEL) is an integral test facility for a VVER-440 type nuclear reactor. The main objectives were to prepare a simulation model for its horizontal steam generator with the TRACE thermal hydraulic code and assess different modeling options of the code. PACTEL experiment LOF-10 was chosen for this assessment. The calculation results showed that TRACE is capable in simulating horizontal steam generator behavior both in steady state and during loss-of-feedwater transient. The phenomenon of heat transfer from primary to secondary side, steam superheating and flow reversal in the lowest heat exc...

Energy Technology Data Exchange (ETDEWEB)

Three experiments were conducted to study the behaviour of the new horizontal steam generator construction of the PACTEL test facility. In the experiments the secondary side coolant level was reduced stepwise. The experiments were calculated with two computer codes RELAP5/MOD3.1 and APROS version 2.11. A similar nodalization scheme was used for both codes so that the results may be compared. Only the steam generator was modeled and the rest of the facility was given as a boundary condition. The results show that both codes calculate well the behaviour of the primary side of the steam generator. On the secondary side both codes calculate lower steam temperatures in the upper part of the heat exchange tube bundle than was measured in the experiments. (orig.) 4 refs.

International Nuclear Information System (INIS)

Steam generators for nuclear plants require access to the primary and secondary side for various inspection, cleaning and repair procedures and upon achieving access various inspection and cleaning tasks must be carried out. Steam generators currently at manufacture are therefore provided with; primary side access via the primary manways, steam drum access via the steam drum manway, U-bend region access via the secondary manway and drum-internals access passages, tubesheet secondary side access via tubesheet level inspection ports and tube support plate access via tube support level access ports. Many CANDU steam generators were built with only primary and drum manways and either no tubesheet ports or ports which were located ineffectively. The steam generators at Wolsong 1 were ...

Energy Technology Data Exchange (ETDEWEB)

The general objective of the International Seminars of Horizontal Steam Generator Modelling has been the improvement in understanding of realistic thermal hydraulic behaviour of the generators when performing safety analyses for VVER reactors. The main topics presented in the fourth seminar were: thermal hydraulic experiments and analyses, primary collector integrity, feedwater distributor replacement, management of primary-to-secondary leakage accidents and new developments in the VVER safety technology. The number of participants, representing designers and manufacturers of the horizontal steam generators, plant operators, engineering companies, research organizations, universities and regulatory authorities, was 70 from 10 countries.

Energy Technology Data Exchange (ETDEWEB)

In order to ensure a power balance in an electrical network a spinning reserve has to be capable of activation within seconds. However, because of the need to vary the fuel flow in the steam generator, the run-up of output takes minutes. The required capacity response is therefore achieved by storage of steam from the steam generator. In this process the turbine control valves come into action in a specific manner. The paper describes the importance of the initial reserve for unit capacity control. (orig.).

International Nuclear Information System (INIS)

The use of heat transfer augmentation devices has been proposed as a means for reducing the cost of steam generators for advanced LMR designs by approximately 25%. Experimental results are presented for two types of enhancers in the form of a twisted-tape and a core-tube insert in the steam tube of a 13.1-m-long, sodium-heated single-tube test section. Test parameters were prototypical of liquid metal reactor (LMR) steam generator design for a once-through high-pressure steam cycle. Heat transfer results are compared with the no-insert case on an overall and local basis. Although both augmentations increased overall performance, the mechanisms were different.

International Nuclear Information System (INIS)

The nuclear power plants with WWER type reactors are characterized by horizontally situated steam generators (SG). During seismic event the horizontal and vertical ground accelerations induce fluid motion in directions of longitudinal and transversal axis. Resulting dynamic forces act on the SG attachment and could cause the failure of screws. In obvious PSA scenarios, these phenomena are classified as a indirect induced LOCA. In this paper the effects of transversal sloshing of fluid are analyzed.

Science.gov (United States)

Eight years ago, corrosion and tube denting seriously threatened the reliability and design life of steam generators, especially for closed loop arrangements in pressurized water reactors (PWRs). Concentrated research by the Steam Generator Owners Group (SGOG) diagnosed the causes and produced effective solutions, notably guidelines for water chemistry control in the secondary loop. The guidelines recommend specific levels of water impurities and remedial actions to prevent cooling-water leaks in the condenser, prevent air leaks, limit corrosion product buildup, and remove some impurities while neutralizing others. Continued research in SGOB-II is investigating intergranular corrosion and stress corrosion cracking. 3 figures.

International Nuclear Information System (INIS)

In this paper the thermal design of a horizontal steam generator for the Indian PHWR nuclear power plant is described. The main attraction is absence of tube sheet and use of stainless steel 'U' tubes. It is emphasised that with appropriate water chemistry it is possible to use stainless steel tubes, which is many times cheaper than the Incoloy tubes used elsewhere. The design approach, applicable equation for the design and the results of computation in the form of heat transfer area and some important dimensions of the steam generator are presented.

Science.gov (United States)

... the effect of MBH on steam generator crevice corrosion; model boiler test results show that MBH is effective against denting corrosion and Alloy 600 ...

International Nuclear Information System (INIS)

The steam generator availability is one of the important problems encountered during the pressurized water nuclear plant operation. Various kinds of corrosion phenomena were observed in the past. These phenomena result from the concentration of impurities mainly in three locations in the steam generators: the tubesheet crevices, the tube support plate crevices, and the sludge pile. Corrections were made in the design and the materials used but a number of steam generators suffer or will suffer from corrosion processes inducing in many cases forcing their replacement. In order to prevent or to retard the corrosions several laboratories have performed experiments to reproduce and to study the corrosion processes. The first step of the degradation is the concentration of chemical species. A method using /sup 24/Na as a radioactive tracer was used to establish the ...

International Nuclear Information System (INIS)

In order to improve the reliability of PWR steam generators, we have performed research to improve the tubing material and tube-support-plate configuration, based on our wide operating experience, and have developed and verified Alloy TT690 as the optimum tubing material and the BEC (Broached Egg Crate) type tube-support design. In the research, we have studied the metallurgical mechanism of the alloy to improve its corrosion resistance, evaluated corrosion susceptible region quantitatively, estimated the actual environment in a steam generator and confirmed the reliability by a model boiler test over a long period. It has been verified that the steam generator with the latest design has higher reliability with respect to the corrosion resistance of tubes. (author).

Energy Technology Data Exchange (ETDEWEB)

This paper deals with the heat transfer characteristics of horizontal steam generators, particularly under natural circulation (decay heat removal) conditions on the primary side. Special emphasis is on the inherent features of horizontal steam generator behaviour. A mathematical model of the horizontal steam generator primary side is developed and qualitative results are obtained analytically. A computer code, called HSG, is developed to solve the model numerically, and its predictions are compared with experimental data. The code is employed to obtain for VVER 440 steam generators quantitative results concerning the dependence of primary-to-secondary heat transfer efficiency on the primary side flow rate, temperature and secondary level. It turns out that the depletion of the secondary inventory leads to an inherent ...

Energy Technology Data Exchange (ETDEWEB)

This project evaluated the feasibility of using a low-concentration (approx. 0.5 wt %) chemical cleaning process to remove corrosion product deposits from steam generator surfaces and magnetite from tube-to-support plate crevices of PWR steam generators. The primary objective was to develop a dilute process that could be safely applied at scheduled intervals, such as during normal refueling outages, to maintain a clean operating condition in the steam generator. The dilute chemical cleaning process developed in this project was demonstrated successfully on two model generators which were operated on faulted chemistry by DOE/CRC at Commonwealth's State Line Facility. Unit 5 was cleaned after 48 days of operation with 1% seawater fouling, and Unit 6 was cleaned after 112 days of operations with Lake Michigan water. This report ...

International Nuclear Information System (INIS)

The investigations carried out on Alloy 800 in aqueous high-temperature environments in France as well as in other countries are reviewed. These studies are mainly concerned with nuclear industry where Alloy 800 can be used as structural material for steam generators of PWR, breeders or HTR. As results referred to in the literature on cracking in caustic environmens do not always agree, a discussion is presented on the matter. The behaviour of Alloy 800 in superheated steam is examined. (Auth.).

International Nuclear Information System (INIS)

For the conversion of thermal energy into electric energy in modern condenser power plants, according to the way of steam generation, two different types of power stations are built: power stations for fossile fuels and nuclear power stations. Also two classes of steam turbines were developed, corresponding to the two power station types, whose steam conditions, by experience and extensive calculations of economy, were determined so that a minimum of power generating cost will result. The two classes, the HMN and the SN series, are composed according to the modular system and designed in such a manner that with a small number of standard components, steam turbines for the power range between 100 and 2,500 MW can be built. (orig.).

International Nuclear Information System (INIS)

The mechanism of the salt concentration increase in pits and crevices formed in a steam generator due to its imperfect manufacture or to its design features is described. The probability of corrosion can be reduced by choosing a suitable steel and by securing low concentrations of salts (chlorides in particular) and corrosion products in the feedwater. Attention is paid to the distribution of salts in the water-steam circuit and to the conditions of erosion corrosion as the principal source of corrosion products in feedwater. Experience with the suppression of erosion corrosion at nuclear power plants abroad is described. (E.J.).

International Nuclear Information System (INIS)

The approach estimating the leak probability of flanged joint due to the destruction of fastening studs is described. The mentioned approach consists of two stages. The probability of destroying one stud is calculated at the first stage, and the probability of different combination interpositions of intact and destroyed studs is calculated at the second one. The probability calculation of leak in the area of collector cover of steam generator PGV-1000 is used as an example of developed approach

International Nuclear Information System (INIS)

Some of the origins of corrosion encountered in the secondary side of pressurized water reactor steam generators are:-sludge accumulation (a mixture of metal oxides, primarily magnetite and copper) on tube sheet and attack of tube support plates by aggressive impurities leading to denting. Although Electricite de France has not suffered from these problems, it has developed a chemical cleaning process to dissolve corrosion products at both locations. (author).

International Nuclear Information System (INIS)

In this paper the possibility of using the test facility PACTEL concerning the investigations of thermal hydraulic special features of the primary coolant circuit acting under natural circulation is under consideration. It is suggested to study a stratification phenomenon of a coolant in upper plenum of a reactor and also a horizontal steam generator (HSG) hot collector temperature regime. For such investigations the facility must be modified. It is shown that this work is not large and expensive, as the facility is a lightly suitable unit for different researches. (orig.)

Energy Technology Data Exchange (ETDEWEB)

The Finnish simulation code APROS and especially its 5-equation model is applied to modelling the horizontal steam generator. Different nodalizations are used in the secondary side of different models. Simulation results of the stationary state run are compared with results of RELAP5/MOD2 calculations and with an experimental plant data. (2 refs., 3 figs., 4 tabs.).

International Nuclear Information System (INIS)

This paper studied the mathematical model in the steady state for the horizontal steam generator, and based on this study, the thermal-hydraulics analysis code HSG-S for the HSG had been developed, and the steady state calculation had been preformed. The results were correct and fit well with RELAP5 results

International Nuclear Information System (INIS)

It is desirable to avoid vibration problems by following appropriate guidelines and specifications at the design stage. Accordingly, design specifications were developed to prevent tube failures due to vibration in nuclear steam generators and liquid heat exchangers. These specifications are outlined in this report. (author). 14 refs., 2 figs.

Energy Technology Data Exchange (ETDEWEB)

The experiments and the test facilities for the study of the stratification phenomenon in the hot plenum of reactor and the upper parts of the steam generator collectors in a nuclear power plant are described. The aim of the experiments was to define the conditions of the stratification initiation, to study the temperature field in the upper part, the definition of the characteristics in the stratification layer, and also to study the factors which cause the intensity of the stagnant volume cooling.

International Nuclear Information System (INIS)

In in-service inspection of steam-generator tubes, the need for reproducibility of the measurement results and for comparison of these results over two successive inspections has led EDF to specify the main characteristics of the Eddy-Current probes and to design equipment to measure them. This equipment is presented here and the electrical and magnetic results are correlated with those obtained by testing tubes with reference defects. 5 refs.

Energy Technology Data Exchange (ETDEWEB)

Corrosion failures in circuits of foreign NPPs are considered. According to American statistics there are more corrosion failures in two-circuit NPPs than in NPPs with one circuit. Steam generators mostly suffer from ''corrosion denting''. Lately pitting corrosion becomes a potentially serious problem. Steam generator vertical tubes are mainly subjected to this corrosion type. Attention is drawn to intercrystalline corrosion. The causes of corrosion are described. The problem of optimization of structural materials is discussed to reduce corrosion failures as well as other methods of decreasing corrosion failures. Organization of nondestructive testing, increased requirements to water and steam purity are of great importance.

Energy Technology Data Exchange (ETDEWEB)

The marine reactor MRX, an integral typed PWR with 100 MWt adopts one-through steam generators with coiling tubes. The cold feed water enters the steam generator and the super heated steam flows out. To avoid occurrence of flow instability in the steam generator due to a density wave oscillation, it is necessary to increase of flow resistance at the feed water inlet. The magnitude of flow resistance to stabilize the flow is determined by a simple linear analysis using a D-division method, of which accuracy is clarified by comparison with SRI's experiment. The external force due to heaving, one of ship motions will affect the flow behavior. Analysis by a modified RELAP5 capable of simulating the ship motions reveals that the effect of heaving becomes especially greater when the state of flow approaches both the conditions of density ...

International Nuclear Information System (INIS)

Youngkwang Unit 3,4 and Ulchin Unit 3 and 4 have had problem of the KSNP Steam Generator due to a severe fretting wear on the tube. In particular, the wears were localized and concentrated in the upper part of U-bend of the Central Cavity region. At the upper tube bundle Central Cavity, the fluid flow velocities and void fraction are very high, because the steam is made by high heat transfer at secondary region. Also, this region is affected easily by fretting wear due to it's unsupported span is longer than another regions. The fretting wear is assumed to be result of Flow-Induced Vibration (F. I. V), which can occur by many mechanisms. EFDP was added to UCN 5,6 for prevent fretting wear by the SEC LCC and DSHIC, a company of design and manufacture of the steam generator, respectively. In order to evaluate the efficacy of EFDP, ANSYS and ATHOS-3 Code were used. From sensitivity ...

Energy Technology Data Exchange (ETDEWEB)

When materials such as metal are into the secondary side of steam generator, they, so called foreign objects, may have influences on the integrity of the steam generator tubes. They cause the tube wear due to the relative motion between the tubes and foreign objects and the tube impact due to flow. The best way to avoid the effects is to remove all the foreign objects. However, it is not easy to remove the foreign materials thoroughly due to their condition such as the location. If the locations of the foreign materials are in the middle of tube bundle and the tube arrangement of the steam generator is the triangle type, the equipment such as FOSAR(Foreign Object Search and Retrieval) can not reach their locations. If the foreign materials stick together with the tubes or tube sheet, they can not be removed. In the case of operating the ...

International Nuclear Information System (INIS)

In the present paper we estimate the effect of sodium in the in-service inspection of non-magnetic steam generators tubes using eddy current technique and eddy current probes based on a differential double bobbin coil configuration. Experimental measurements of defects signals in steam generator tubes of fast breeder reactor are compared with simulations results of a two-dimensional axisymmetric finite element code to validate a reliable electromagnetic model of the system (eddy current coils, steam generator tube, defect) when there is no sodium on the outer steam generator tube surface. The electromagnetic code is used to evaluate the sodium band and sodium layer signals when a defect is located under steam generator support plate. Using a multi-frequency algorithm, its ...

International Nuclear Information System (INIS)

The understanding of fission product deposition in realistic steam generator conditions is needed for release estimates in PSA studies, and for the design of efficient accident management procedures. This is considered very important because primary-to-secondary leakages risk dominant sequences in many plants. Furthermore, the decay heat of the fission product deposits adds to the thermal load to the steam generator tubes also in other sequences, especially in case of cold leg leakages. This brings out the concern of induced steam generator tube ruptures in cases, where the steam generators are initially intact. The experimental data showed that the highest deposited fraction within the tubes were found in cases with lowest flow velocities. The minimum value of the deposited fraction was observed at intermediate flow ...

Energy Technology Data Exchange (ETDEWEB)

As part of a coordinated program, AECL is developing a set of tools to aid with the prediction and management of steam generator performance. Although stress corrosion cracking (of Alloy 800) has not been detected in any operating steam generator, for life management it is necessary to develop mechanistic models to predict the conditions under which stress corrosion cracking is plausible. Therefore, constant extension rate tests were carried out for Alloy 800 under various steam generator crevice chemistry conditions at applied potentials. These tests were designed to evaluate the stress corrosion cracking susceptibility of Alloy 800 under CANDU( steam generator operating conditions. Based on the experimental results, the recommended electrochemical corrosion potential/pH zone for Alloy 800 determined by electrochemical ...

International Nuclear Information System (INIS)

Three-dimensional computer simulation and analyses of the horizontal steam generator thermal-hydraulics of the WWER 1000 nuclear power plant have been performed for 50% and 75% partial loads, 100% nominal load and 110% over-load. Presented results show water and steam mass flow rate vectors, steam void fraction spatial distribution, recirculation zones, swell level position, water mass inventory on the shell side, and other important thermal-hydraulic parameters. The simulations have been performed with the computer code 3D ANA, based on the 'two-fluid' model approach. Steam-water interface transport processes, as well as tube bundle flow resistance, energy transfer, and steam generation within tube bundles are modelled with "closure laws". Applied approach implies non-equilibrium thermal and flow conditions. The model is solved by the ...

International Nuclear Information System (INIS)

Corrosion of steam generator heat transfer tubes (SGHTT) is one of the important problems which affect safety operation of nuclear power plants (NPP), and the hazard of pitting corrosion of heat transfer tubes is the most serious. With an acoustic emission device, the signals during a corrosion test on SGHTT were collected and analyzed, and the corrosion points in the tubes were located accurately. The results show that pitting corrosion of heat transfer tubes has passed through three periods in its development: expansion phase, stationary phase and rapid developing phase. The corrosion damage of HTT can be found earlier with acoustic emission than any other non-destructive testing methods. Acoustic emission can be used for on-line and real-time monitoring of the safety and operation of the steam generator and has therefore a great significance. (orig.)

International Nuclear Information System (INIS)

With steam generator check testing, the numerous types of defect not only have to be detected reliably but also to be identified correctly and classified according to depth. A priori, no signals are provided by the physical eddy-current mechanism, i.e. the disturbance of an excited eddy-current field by a defect, which will allow unambiguous identification and classification of the defects. The task, therefore, consists in designing the eddy-current method in terms of measurement technology in such a way as to obtain useful detection, identification and classification by a translation of the signals as simple and as save as possible. Internal through-flow wils and internal rotating probes were examined for their suitability to fulfill this three-pronged task with steam generator inspection. (orig./RW).

International Nuclear Information System (INIS)

Problems are discussed of heat exchange tubes of Westinghouse type vertical steam generators exhibiting corrosion damage such as point corrosion, planar corrosion, tube denting, corrosion stress cracking, crevice corrosion, fretting corrosion and intergranular corrosion. Attention is also paid to problems of WWER-440 type horizontal steam generators, where the level fluctuation area is critical; noncompact porous deposits of the corrosion products give rise to crevice effects and cause significant concentration of chloride ions and other additions. This problem can be partly resolved by a modification of the collector design at the level variation area. An additional measure is the production of steel 08Kh18N10T with a very low level of harmful elements and inclusions. (Z.M.). 3 figs., 11 refs.

International Nuclear Information System (INIS)

This report discusses the work done on EPRI program S149-1. In this program the feasibility of cleaning steam generators off line with organic chelants as a means of arresting denting corrosion was investigated. The rationale behind this program is to make use of those periods during which nuclear steam generators are in cold shutdown or wet layup to carry out a low-temperature soak with a combined chelant-inhibitor solution in order to dissolve some of the magnetite which has built up in crevices and to concomitantly remove entrained corrodents such as chloride ion. It was hoped that these soaks would be effective in reducing carbon steel support plate corrosion which produces tube denting.

Energy Technology Data Exchange (ETDEWEB)

Currently it can be observed that in the case of generator drives as well as 'mechanical drives' smaller units are demanded with a steam turbine capacity of up to 150 MW and clearly higher efficiencies. MAN TURBO is meeting the challenge through realisation of a comprehensive development project aiming at the extension of the application range of the current steam turbine series.

Energy Technology Data Exchange (ETDEWEB)

This paper presents the Cheng cycle system featured by variable heating/generation ratio for effective operation of cogeneration systems (CGS). In this system, a superheater and reheating burner are added to an exhaust heat recovery (EHR) boiler for conventional gas turbine CGSs, while additional injection steam piping is attached to a gas turbine. Steam is injected through manifolds mounted on the periphery of a combustion chamber, and hot gas mixture of steam and air in a combustion chamber is expanded in a turbine and converted to motive power. This technology thus can improve efficiency and output power, and can operate variably CGSs corresponding to heating and generation demands. This technology has been promoted by introducing the technology of middle class Cheng cycle CGS with 4MW class gas turbine from IPT Co., U.S.A. The first system of 6400 kW is now under production for ...

Energy Technology Data Exchange (ETDEWEB)

The externally fired combined cycle is emerging as an economically viable repowering options for old non-competitive coal fired steam plants. This paper describes the initial operation of a pilot plant located at Kennebunk, Maine and the initial work on the repowering of a 48MW coal fired steam plant located in Warren, PA.

Energy Technology Data Exchange (ETDEWEB)

Electricity is a universal energy and one of the driving powers of civilisation. World-wide, future developments like growing economies and structural changes in industry, a higher standard of living and a bigger world population - are impossible without appropriate supply of power and electricity. An expert of Siemens Power Generation presents his views. (orig.)

Energy Technology Data Exchange (ETDEWEB)

The PGV-1000 secondary side natural circulation calculation results for steady-state operational conditions and their analysis using the data of actual NPP circulation investigation are presented. (3 refs., 3 figs., 3 tabs.).

Science.gov (United States)

This report is the final report on the seismic testing of reactor components conducted since 1977 with opening of the vibration laboratory at KAERI. In 1979, forced vibration testing of Wolsung-1 steam generator model using sine dwell and white nosie rand...

Energy Technology Data Exchange (ETDEWEB)

CANDU steam generators and heat exchangers have been afflicted with a significant variety of degradation mechanisms. Until recently, detection and characterization of most of the varieties of flaws required usage of several different eddy current techniques. Tubes needed to be scanned multiple times to ensure they were inspected for all possible degradation mechanisms. The time required to deploy so many different types of techniques to inspect the steam generator tubes has had a significant impact on the inspection schedules. The X-Probe was designed with an array of eddy current coils capable of detecting, characterizing and sizing axial, circumferential and volumetric flaws. The X-Probe is a fast-scanning probe with scanning speed equivalent to speeds typically used for simple bobbin coil probes. With its capability of fulfilling most if not all of the functions of the other tube-testing eddy current ...

Energy Technology Data Exchange (ETDEWEB)

Published in summary form only.

International Nuclear Information System (INIS)

The thermal performance monitoring and optimization system TEMPO is developed at the OECD Halden Reactor Project. The system supports staff of nuclear power plants in identification and correction of problems, which cause small decreases in plant efficiency but which may lead to significant economical losses. The system-wide physical model consists of mathematical description of individual components, such as the reactor, the pumps, the heat exchangers, or the turbines, etc. TEMPO code has recently been extended with new steam generator (SG) models. The present paper summarizes the thermal-hydraulic modelling aspects of the vertical and the horizontal SG. The heat balance equations and their solution are shown with the appropriate initial and boundary conditions. The method of the calculation of the pressure losses are also introduced. The vertical SG model is based on a U-tube structure and treated as a 1D flow channel. The horizontal approach ...

Energy Technology Data Exchange (ETDEWEB)

The authors report how the Midland cogeneration venture (MCV) is repowering the incomplete Midland nuclear plant to operate as a gas-fired combined cycle cogeneration facility. They discuss how their company is responsible for performing engineering, procurement, licensing, construction, start-up, training, and operational assistance for this facility. As shown, twelve gas turbine generator sets supply heat to twelve heat recovery steam generators (HRSGs), which are headered together on the steam side to provide energy to either of two existing steam turbine generators. This combination of equipment enables approximately 1,380 MWe of electrical generation capability, while supplying an average steam flow of 629,000 lbs/hr to an adjacent Dow Chemical Company pant. The MCV facility will also provide 60 MW of electric power ...

International Nuclear Information System (INIS)

Object: To control the average liquid level of each steam generator at a constant level irrespective of the flow rate of sodium thereby to decrease change in the retained amount of sodium and at the same time to improve the load response characteristic. Construction: A method for decreasing to as large an extent as possible a change in the amount of sodium retained in a steam generator due to change in the flow rate, which comprises the steps of detecting the main recirculating flow rate of liquid sodium by the use of a sodium flow rate detector, amplifying the detected flow rate signal depending upon the ratio between the flow rates respectively in a super-heater and a re-heater (distribution ratio), delivering the amplified signal to a function generator which generates a liquid level setting signal for maintaining the respective average liquid levels of the ...

Energy Technology Data Exchange (ETDEWEB)

Mitsubishi is now developing a new Pressurized water reactor (PWR), the Mitsubishi simplified PWR (MS-PWR), which has the innovative features of hybrid safety systems (an optimum combination of passive and active systems) and cooling by horizontal steam generators. In order to confirm the feasibility of the Mitsubishi hybrid safety system, various kinds of safety analyses are performed for loss-of-coolant accident events. In parallel to these safety analysis efforts, the following thermal-hydraulic tests are to be performed: (1) thermal-hydraulic test of a horizontal steam generator; (2) integrated thermal-hydraulic test using a simulation loop for the innovative MS-PWR (SLIM).

Energy Technology Data Exchange (ETDEWEB)

The effect of heat transfer is described from heat exchange tubes of a horizontal steam generator on the distribution of primary water to the individual tubes of the tube bundle. It is shown that in a broad interval of mass flow rates and lengths of heat exchange tubes, the simplified method of calcualtion, i.e., calculation of the distribution of primary water into heat exchange tubes neqlecting the changes of physical properties of water along the heat exchange tubes, will yield sufficiently accurate results.

International Nuclear Information System (INIS)

The present process allows to eliminate the corrosion products formed on the tube plates and in the interstices of plate-tube crosspieces of a PWR steam generator in order to avoid a corrosion phenomenon which may cause denting by presence of oxides. The process consists in applying on these oxides at about 50-100 degrees, an aqueous solution containing 6-8% of gluconic acid, 3-5% of citric acid, about 0.5% of a corrosion inhibitor and ammonia until a pH of 3-9.5 is obtained.

International Nuclear Information System (INIS)

The objectives of the project were to develop operational techniques for promoting contaminant hideout return from tube support crevices and to identify the effect of chemical inhibitor application on corrodent transport. The implementation of routine procedures for promoting the return of sequestered corrodents could retard the progression of denting or other corrosion processes and improve steam generator availability. Tests also quantified the effect of inhibitor application on crevice hideout and hideout return processes, with the intention of developing a better understanding of the inhibition mechanism. By carefully monitoring the hideout and hideout return inventories, the program also has provided the opportunity to study steam generator concentration processes in general.

International Nuclear Information System (INIS)

The paper is concerned with the simulation of practical testing situations which are too difficult and/or expensive to replicate in a laboratory environment. Numerical experiments are described which simulate the differential eddy current probe response to the build-up and chemical flushing of magnetite in the crevice gap of a PWR steam generator unit. The simulation results agree well with the only experimental data available to the authors and lead to the conclusion that conventional differential eddy current probes should be capable of characterizing crevice gap conditions with respect to the presence of magnetite. (author).

International Nuclear Information System (INIS)

As part of a coordinated program, AECL is developing a set of tools to aid with the prediction and management of steam generator performance. Although stress corrosion cracking (of Alloy 800) has not been detected in any operating steam generator, for life management it is necessary to develop mechanistic models to predict the conditions under which stress corrosion cracking is plausible. Therefore, constant extension rate tests were carried out for Alloy 800 under various steam generator crevice chemistry conditions at applied potentials. These tests were designed to evaluate the stress corrosion cracking susceptibility of Alloy 800 under CANDU( steam generator operating conditions. Based on the experimental results, the recommended electrochemical corrosion potential/pH zone for Alloy 800 determined by electrochemical ...

International Nuclear Information System (INIS)

We propose a dissolution-passivation model to account for the electrochemical behavior of Alloy 800 and 316L stainless steel under steam generators operating conditions. This model is similar to the one defined for nickel, considering the known importance of this element on the studied alloys. The comparison of the rates of the elementary steps points out the favorable influence of dissolved oxygen and phosphate ions on the dissolution of Alloy 800, even in the presence of chloride ions. The sensitivity of 316L SS is greater, especially because of the lowest stability of oxide-like intermediate.

International Nuclear Information System (INIS)

Although there is service experience of Alloy 800 as tubing for superheaters in conventional and nuclear (HTR) power stations and in PWR heat exchangers, there is no corresponding service experience in sodium-cooled fast reactor steam generators. However, some limited experimental studies have been made of corrosion behaviour, and of possible structure modifications and effects on mechanical properties which occur during exposure of this material to a high temperature sodium environment, and these are summarised in the paper. It is concluded that further work needs to be done before Alloy 800 can be confidently endorsed for use as tubing in fast reactor steam generators. (author).

Energy Technology Data Exchange (ETDEWEB)

The SGTR project has been carried out within the fifth EURATOM Framework Programme (Contract No FIKS-CT-1999-0007). Its main objective was to provide an experimental database and to develop and/or verify models to support definition of accident management measures in the hypothetical case of a Steam Generator tube Rupture (SGTR) sequence. The project addressed both vertical and horizontal steam generator designs. This report summarises the main results obtained in the intermediate scale experimentation that addressed Western type steam generators. The specific goal of this test programme was to investigate aerosol retention in the break stage of the secondary side of a water-empty steam generator. The test matrix consisted of 12 tests that explored the influence of variables such as break type and orientation and inlet ...

Energy Technology Data Exchange (ETDEWEB)

A consequent use of computers and standardized components enables high efficiency for steam turbines while produced at lower manufacturing cost as shown here in the Muehlheim company of Siemens Power Generation, where steam turbines are produced for new stations and retrofitted elder power plants. [German] Auch eine ueber hundert Jahre alte Technik bietet Raum fuer Innovationen: Konsequenter Computereinsatz und standardisierte Baureihen ermoeglichen bei Dampfturbinen hohe Wirkungsgrade bei relativ niedrigen Fertigungskosten - umgesetzt im Werk Muelheim der Siemens Power Generation, wo Dampfturbinen fuer Neuanlagen und zum Modernisieren bestehender Kraftwerke entstehen. (orig.)

Energy Technology Data Exchange (ETDEWEB)

This paper describes the modeling of horizontal steam generator with the TRACE code and calculation results of a loss-of-feedwater (LOF-10) experiment at the PACTEL facility. Parallel Channel Test Loop (PACTEL) is an integral test facility for a VVER-440 type nuclear reactor. The main objectives were to prepare a simulation model for its horizontal steam generator with the TRACE thermal hydraulic code and assess different modeling options of the code. PACTEL experiment LOF-10 was chosen for this assessment. The calculation results showed that TRACE is capable in simulating horizontal steam generator behavior both in steady state and during loss-of-feedwater transient. The phenomenon of heat transfer from primary to secondary side, steam superheating and flow reversal in the lowest heat exchange tubes were studied in detail. Different ...

Energy Technology Data Exchange (ETDEWEB)

The Coal Fired Combined Cycle (CFCC) power system thermodynamic cycle is illustrated schematically. Pressurized air supplied at the discharge of gas turbine compressors is ducted to the pressure vessel of pressurized, fluidized-bed, combustor-steam generator modules. The air is introduced in parallel to the beds, entering through distribution grids beneath each bed. Steam generation tubes are buried within the beds and are also arranged as membrane tube walls enclosing the four sides. Crushed coal (1/4 inch x 0) is pneumatically fed at locations just above the air inlet grids at the bottom of each bed. Dolomite is similarly fed to the individual beds. The coal is burned at a temperature below the ash fusion point. Sulfur is removed in the fluid beds through reaction of the SO/sub 2/ with CaCO/sub 3/ and O/sub 2/ to form solid CaSO/sub 4/ and CO/sub 2/ gas. The combustion gases leave the beds at a ...

Energy Technology Data Exchange (ETDEWEB)

In this study, the adoptable essential technologies and reference design concept of the advanced reactor were developed and related basic experiments were performed. (1) Once-through Helical Steam Generator: a performance analysis computer code for heli-coiled steam generator was developed for thermal sizing of steam generator and determination of thermal-hydraulic parameters. (2) Self-pressurizing pressurizer : a performance analysis computer code for cold pressurizer was developed. (3) Control rod drive mechanism for fine control : type and function were surveyed. (4) CHF in passive PWR condition : development of the prediction model bundle CHF by introducing the correction factor from the data base. (5) Passive cooling concepts for concrete containment systems: development of the PCCS heat transfer coefficient. (6) Steam injector ...

Energy Technology Data Exchange (ETDEWEB)

Early in 1988 RWE decided to overhaul the power station at Goldenberg. This was necessary due to the age of the plant, the obligation to supply heat and process steam and recent amendments of environmental legislation. After analysing demands and needs one developed a plant concept with a 290 t/h and a 400 t/h steam generator. Due to lack of space the decision was made in favour of a circulating atmospheric fluidised bed combustion instead of steam generators with pulverised lignite combustion as these steam generators take up much more space due to the legally required desulphurisation plants for gaseous emissions. Apart from generating steam RWE wishes to gain experience with this combustion technology and test it for suitability and use in large-scale plant. (orig./KO) [Deutsch] Aufgrund des ...

Energy Technology Data Exchange (ETDEWEB)

Short communication

Energy Technology Data Exchange (ETDEWEB)

Short communication

Science.gov (United States)

Described herein is a system for the co-generation of hydrogen gas and electricity, wherein the proportion of hydrogen to electricity can be adjusted from 0% to 100%. The system integrates fuel cell technology for power generation with fuel-assisted steam-electrolysis. A hydrocarbon fuel, a reformed hydrocarbon fuel, or a partially reformed hydrocarbon fuel can be fed into the system.

International Nuclear Information System (INIS)

For shut-down safety of the turbine generator (securing of auxiliary power operation after load shut-down and preventing the reaching of overspeed after load shut-down with disturbed turbine governing system) additional measures compared to those for superheated steam turbines are required for turbine generators in plants with pressurized water reactor (PWR) as well as those with boiling water reactor (BWR) . Equipment is described (e.g. overspeed govern or selecting connection, vacuum breaker, bypass valves, intercepting valves) which, depending on the own conditions of the individual turbine generator (e.g. run-up time, vacuum, enclosed energy), may be applied alone or in jointly. (orig.).

Energy Technology Data Exchange (ETDEWEB)

Heat transfer processes use fluids which are generally not pure and can react with transfer surfaces. These surfaces are subject to deposits which can be sediments harmful to heat transfer and to integrity of materials. For nuclear plant steam generators, sludge build-up accelerates secondary side corrosion by concentrating chemical species. A major safety problem involved with such a corrosion is the growing of circumferential cracks which are very difficult to detect and size with eddy current probes. With a view to understand and control this problem, it is necessary to develop a mathematical model for the prediction of sludge behavior in PWR steam generators. Based on fundamental principles, this work intends to use different models available in literature for the prediction of the phenomenon leading to the accumulation of sludge particles at the bottom (the tubesheet) of a PWR. For that, a ...

Energy Technology Data Exchange (ETDEWEB)

Fully-fired combined cycle plants utilizing the high temperature gas turbine discharge as gas intake for a coal-fired steam generator were developed in the 1960's. Long-term experience and the excellent reliability statistics of the gas turbines of these plants have proven that the thermodynamic, economical and ecological advantages of combined cycle units over conventional units strongly suggest such plants for fossil-fuel applications. This paper describes the design and operation of these plants. Topics are Gersteinwerk Unit K combined-cycle power plant, plant design concept, design criteria and performance, gas turbines, steam generators, steam turbines, various modes of operation, typical plant start-up, environmental considerations, prototype plants and topping old steam turbine plants. It concludes that combined-cycle plants provide efficiencies ...

Energy Technology Data Exchange (ETDEWEB)

Nature of tube degradation was studied in support of horizontal steam generator lifetime calculations. Experiments were carried out to evaluate the resistance to environmentally assisted cracking (EAC) of steam generator tubes during operation. Estimations of the incubation period for crack initiation and the threshold K value, K(ISCC), and the crack growth rate were made to predict evolution of damage in tube walls. Experimental results were compared with in-service inspections experience. The paper summarizes results of experiments of C ring specimen for the initiation testing and results of SENT (single edge notch tensile) specimen for the crack growth rate (CGR) testing. The specimens were exposed to concentrated environments at elevated temperatures simulating crevice environments in secondary side crevices in horizontal steam generators. The results show ...

Energy Technology Data Exchange (ETDEWEB)

In the last two years, the VDMA work group 'Functional security' compiled a qualitative and quantitative procedure for the determination and evaluation of turbo-machine specific risks. With the calibrated risk graph for turbo-machines the requirements (Safety Integrity Level SIL) to the protective functions for turbo-machines (steam turbines, gas turbines, generators and compressors) were determined. With consideration of the legislation, the work group compiled a recommendation for the renewal and/or reconstruction of protective functions with old facilities.

Energy Technology Data Exchange (ETDEWEB)

This paper presents the objective, scope, and status of the US Department of Energy's Materials and Structures Program to develop a data base on mechanical properties of structural materials for out-of-core structures and components for LMFBRs. Information on the development of a reference data base on materials for the reactor system, reactor enclosure system, primary heat transport system, intermediate heat transport system, and steam generator system is included. In addition, the development of the data and analyses to account for the effects of temperature and stress, as well as water/steam, sodium, and radiation environments, is described. Plans for the development of alternative materials for future out-of-core applications are presented.

Energy Technology Data Exchange (ETDEWEB)

Kawasaki Heavy Industries, Ltd. has developed a gas turbine combined cycle and cogeneration plant computer aided engineering (CAE) system, which creates cycle heat balance automatically on given configurations of gas turbines, heat recovery boilers, steam turbines and auxiliaries. The system can predict the performance of the plant as a function of control set points, loads, process steam requirements and ambient conditions, which enables accurate and fast design of gas turbine combined cycle and cogeneration plants. (author)

International Nuclear Information System (INIS)

Existing applications of Alloy 800 are summarized, with particular reference to its use in various types of reactor. The need for a co-ordinated research and development programme is stressed, and the variables to be explored are outlined. The papers relating to the problem of corrosion and cracking in water and steam are considered. the strength and ductility of Alloy 800 is considered. Finally, sections of the summary deal with the use of Alloy 800 for (a) sodium cooled fast reactor boiler tubes; (b) the high temperature gas cooled reactor; and (c) PWR steam generator tubes. (U.K.).

International Nuclear Information System (INIS)

Appraisals of the isothermal general corrosion of Incoloy 800 in steam at 595 and 650 deg.C indicated good long-term resistance (<7 mg/cm"2 weight gained in 20 years) whether surfaces were ground or electropolished. Electropolished surfaces corroded much faster than ground surfaces in earlier stages, (39 and 15 times as fast in 3000 hr at 595 and 650 deg.C, respectively), but the corrosion rate decreased with time and gave comparable estimates over the long term for both surface preparations. Both nonwelded and welded Incoloy 800 were definately susceptible in chloride stress-corrosion cracking tests of U-bend specimens in a cyclic, wet-dry steam environment. However, C-configuration specimens of Incoloy 800, which were less severly strained, did not crack in this environment. (author).

International Nuclear Information System (INIS)

The sixth AER benchmark is the second AER benchmark for couplings of the thermal hydraulic codes and three dimensional neutron kinetic core models. It concerns a double end break of one main steam line in a WWER-440 plant. The core is at the end of its first cycle in full power conditions. In VTT HEXTRAN2.9 is used for the core kinetics and dynamics and SMABRE4.8 as a thermal hydraulic model for the primary and secondary loop. The plant model for SMABRE consists mainly of two input models, Loviisa model and a standard WWER-440/213 plant model. The primary loop includes six separate loops, the pressure vessel is divided into six parallel channels in SMABRE and the whole core calculation is performed in the core with HEXTRAN. The horizontal steam generators are modelled with heat transfer tubes in five levels and vertically with two parts, riser and downcomer. With this kind of detailed modelling of steam ...

International Nuclear Information System (INIS)

In 1972 a consortium consisting of Gebruder Sulzer A.G., Winterthur and EVT, Energie- und Verfahrenstechnik G.m.b.H., Stuttgart, was awarded the contract by HBR, Hochtemperatur Reaktorbau G.m.b.H., Koln/Mannheim, for the design, manufacture and installation of the 6 steam generator units for the prototype 300 MWe Thorium High Temperature Reactor (THTR) at Schmehausen, Germany. The design and manufacture of the 6 units is done by Sulzer; installation of the units as well as the manufacture of the external pipework and headers is done by EVT. The units are built under semiclean conditions in a special workshop at CCM-Sulzer in mantes, France. This workshop had been used previously for the manufacture of the steam generators for the nuclear power plants St. Laurent I, II, Vandellos (500 Mwe each) and EL-4 (70 Mwe). The present paper defines the operating conditions and the chemical composition of the ...

International Nuclear Information System (INIS)

The Belgian Tihange 2 nuclear power plant went into commercial operation in 1983 producing a thermal power of 2785 MW. Since the commissioning of the plant the steam generators U-tubes have been affected by primary stress corrosion cracking. In order to avoid further degradation of the performance and an increase in repair costs, Electrabel, the owner of the plant, decided in 1997 to replace the 3 steam generators. This decision was supported by the feasibility study performed by Tractebel Energy Engineering which demonstrated that an increase of 10% of the initial power together with a fuel cycle length of 18 months was achieved. Tractebel Energy Engineering was entrusted by Electrabel as the owner's engineer to manage the project. This paper presents the role of Tractebel Energy Engineering in this project and the safety analysis program necessary to justify the new operation point and the fuel cycle ...

International Nuclear Information System (INIS)

'Full text:' In Germany, Alloy 800 (high nickel austenitic stainless steel) was modified and qualified for the use as steam generator tubing by Siemens/KWU (now AREVA NP GmbH). The service reliability of Alloy 800 has been demonstrated over a long period of time. 1968 Siemens/KWU decided to use this material for the NPP Stade, which started operation in 1972. The steam generators operating with Alloy 800 tubes have now been in service for more than 30 years. Up to now no PWSCC or secondary-side SCC has been observed in the more than 285,000 tubes installed in Siemens/KWU steam generators (including RSG) in 19 PWRs in Europe. The operating experience will be shown and discussed. During the past regular SG tubing inspections using eddy current testing, a few indications were detected within the tube sheet between upper and lower tube expansion. These indications ...

International Nuclear Information System (INIS)

PWR steam generators have suffered from a variety of degradation phenomena. This paper identifies the corrosion-related defects and their probable causes and suggests approaches to correct and prevent corrosive activity. In the attempt to solve the degradation problems, research programs have concentrated on modifying materials, stresses, and the chemical environment in both new and operating steam generators. The following corrosion-related defects have been studied: tube wastage, denting, primary side (ID) intergranular stress corrosion cracking (IGSCC), OD-initiated intergranular attack (IGA), pitting, and corrosion fatigue. Plants affected by wastage have greatly reduced their problem by adopting an all volatile treatment (AVT). In the case of denting, a less aggressive chemical environment is recommended. Primary side IGSCC responds to temperature reduction, stress relief, and material ...

Energy Technology Data Exchange (ETDEWEB)

In support of a project on lifetime calculation experiments were carried out to evaluate the resistance to environmentally assisted cracking (EAC) of steam generator tubes during operation. Estimations of the incubation period for crack initiation and the threshold K value, K{sup Iscc}, and the crack growth rate were made to predict evolution of damage in tube walls. The paper summarizes results of experiments of C ring specimen for the initiation testing and results of SENT (single edge notch tensile) specimen for the crack growth rate (CGR) testing. The specimens were exposed to concentrated environments at elevated temperatures simulating crevice environments in secondary side crevices in horizontal steam generators. The results show that the material of SG tubes is sensitive to transgranular environmentally assisted cracking in the three basic concentrated environments used, alkaline, neutral and ...

International Nuclear Information System (INIS)

Special T-shaped feedwater distribution pipes were installed in steam generators at the Loviisa (Finland) and Rovno (Russia) nuclear power plants. The new shape was tested in an extensive testing programme. Since the tubes frequently suffer from corrosion damage, large-scale water hammer experiments were performed on a model facility in 1996. The main objectives of the water hammer experiments were to find out the prevailing parameters leading to water hammers, as well as the sensitivity of hammering to boundary conditions. A water hammer may occur when the mass flow rate into the steam generator exceeds 6 kg/s and the temperature difference between steam generator and feedwater exceeds 100 degC. Visual experiments and stress analyses of the pipe were also carried out. The weakest part, the T-joint, may hold against such water hammers only for a limited time of ...

International Nuclear Information System (INIS)

The application of chemical cleaning for dissolving and removing scale and sludge is being planned in the Japanese pressurized water reactor (PWR) plant in order to maintain high heat transfer performance and to prevent steam generator (SG) tube degradation. In this paper, the effectiveness of the Electric Power Research Institute (EPRI) and German Kraftwerk Union (KWU) processes on the integrity of structural materials other than SG tubes and the comprehensive applicability of chemical cleaning are discussed. The integrity of structural materials such as carbon steel, low-alloy steel and stainless steel was maintained after the EPRI and KWU processes. KWU chemical cleaning tailored for crevice cleaning has been studied to improve its cleaning effectiveness in crevices and to control the corrosion depth of structural materials less than the criterion for corrosion depth. (author)

Energy Technology Data Exchange (ETDEWEB)

ALSTOM Power Inc.'s Power Plant Laboratories (ALSTOM) has teamed with the U.S. Department of Energy National Energy Technology Laboratory (DOE NETL), American Electric Company (AEP) and Parsons Energy and Chemical Group to conduct a comprehensive study evaluating coal fired steam power plants, known as Rankine Cycles, equipped with three different combustion systems: Pulverized Coal (PC), Circulating Fluidized Bed (CFB), and Circulating Moving Bed (CMB{trademark}). Five steam cycles utilizing a wide range of steam conditions were used with these combustion systems. The motivation for this study was to establish through engineering analysis, the most cost-effective performance potential available through improvement in the Rankine Cycle steam conditions and combustion systems while at the same time ensuring that the most stringent emission performance based on CURC (Coal Utilization Research ...

Energy Technology Data Exchange (ETDEWEB)

The project is focused on the thermal hydraulic analyses of nuclear power plants. Specific areas of research have been the modelling of heat transfer in horizontal steam generator in presence of non-condensable gas, and the development of tools for multidimensional two-phase flow simulations. The effect of non-condensable gas on the heat transfer in the horizontal steam generator (SG) has been studied by calculating with APROS the PACTEL experiments NCG-1 (air injection) and NCG-3 (helium injection). The work done for the two-phase flow model development consists of two parts; improving the solution algorithm of porous media code PORFLO, and adding a homogeneous two-phase model to the commercial CFD code Fluent. (orig.)

International Nuclear Information System (INIS)

Alloy 800 has been used for steam generator (SG) tubing for more than 30 years, primarily in CANDU reactors worldwide and in reactors in Germany. Extensive laboratory testing and in-service experience suggest that the Alloy 800 tubing has excellent resistance to corrosion-related degradation under appropriate operating conditions. In planning refurbishment of nuclear plants stations, a key concern is the longevity of existing SGs up to the 60-year lifetime of the refurbished plant. The paper reviews an existing methodology based on the concept of the improvement factor, and refines its estimation based on data specific to CANDU operating conditions. The paper presents a more advanced Bayesian probabilistic approach to estimate the degradation free lifetime distribution of Alloy 800 tubing, which is used to quantify the probability of degradation during the service life and to evaluate the impact of potential occurrences of degradation on ...

International Nuclear Information System (INIS)

The Seismic Proving Test of Heavy Component with Energy Absorbing Supports has been conducted to prove the reliability of advanced seismic technology, supporting heavy component such as PWR steam generator with large capacity energy absorbing supports under the sponsorship of Ministry of Economical Trade and Industry. If energy absorbing supports are adopted for NPP heavy components, support structure of facility will be much simplified due to their seismic energy absorbing effect. The paper describes the results of lead damper element test and seismic test at Tadotsu Laboratory, using 1/2.5 scale PWR Steam Generator model supported by Lead Extraction Damper (LED) and development of characteristics evaluation formula of energy absorbing support. (author)

International Nuclear Information System (INIS)

An overview of LOCA analysis in Japanese industry is presented. The BASH-M code, developed for large scale LOCA reflooding analysis, is given as an example of verification and improvement of US computer programs are given. The code's application to the operational safety analysis concerns the following main areas: 1D drift flux model base computer program CANAC; CANAC-based advanced training simulator; emergency operating procedures. The author considers also the code application to the following new PWR safety design concepts: use of steam generators for decay heat removal at LOCA conditions; use of horizontal type steam generator for maintaining two-phase natural circulation under the reactor coolant system submerged. 9 figs.

British Library Electronic Table of Contents (United Kingdom)

In this paper, a robust water level control system for the horizontal steam generator (SG) using the quantitative feedback theory (QFT) method is presented. To design a robust QFT controller for the nonlinear uncertain SG, control oriented linear models are identified. Then, the nonlinear system is modeled as an uncertain linear time invariant (LTI) system. The robust designed controller is applied to the nonlinear plant model. This nonlinear model is based on a locally linear neuro-fuzzy (LLNF) model. This model is trained using the locally linear model tree (LOLIMOT) algorithm. Finally, simulation results are employed to show the effectiveness of the designed QFT level controller. It is shown that it will ensure the entire designer's water level closed loop specifications.

International Nuclear Information System (INIS)

Steam generator tubes in nuclear power plants are periodically checked by means of eddy current probes. The output of a probe is composed of three types of signals: known events (rolling zone, support plates, U-bend part), noise (mainly metallurgical noise) and possible flaws. The latter are random transients, both in arrival time and in shape: they have to be detected and then estimated, before to be fed to the high level stages of a diagnostic system. The objective of the study presented is to develop a semi-automatic system, which could manage and process more than 1 M-bytes of data per tube and provide an operator with reliable diagnostics proposals within a few minutes. This can be achieved only by cooperation of several digital signal processing techniques: detection, segmentation, estimation, noise subtraction, adaptive filtering, modelization, pattern recognition. The paper describes some of these items.

International Nuclear Information System (INIS)

Briefly described is knowledge of crevice corrosion, corrosion cracking and denting. In evaluating the corrosion resistance of steam generator materials it is necessary to distinguish corrosion problems caused by the primary coolant eide and by the secondary circuit side. At present tubes are manufactured of 7 austenitic alloys of a different chemical composition, and available information shows that views on their corrosion resistance differ. Greatest attention has been devoted to corrosion cracking in the presence of HaOH. Findings related to I-600, I-800, I-690 and AISI 316 are given. Corrodibility by sulfur-containing products is now being studied, namely the intercrystalline corrosion cracking caused by the presence of H_2S_4O_6. Knowledge gained in this respect is summed up. (J.P.).

International Nuclear Information System (INIS)

Titanium stabilized austenitic steel is sensitive to SCC in the secondary water under the horizontal steam generator operating conditions. SCC was observed under crevice conditions both at the primary collector flanges and the heat exchange tubes. In the crevice environment sulfates and chlorides as aggressive species and silicates and alumino-silicates as ''non-aggressive'' species are present in significant amounts. Local water chemistry parameters were evaluated using the MULTEQ Code. SCC experiments were carried out by rising displacement tests ar 275 deg C in an environment simulating the crevice conditions. Crack growth rate and K_I_S8C_C were determined for the environment where contents of some species were from 10"2 to 10"4 times higher than in blowdowns. (authors)

Energy Technology Data Exchange (ETDEWEB)

A three-dimensional (3D) thermal-hydraulic model and a numerical procedure for the simulation and analysis of a steady-state, as well as transient operation of nuclear power plant components are presented. A two-fluid approach is applied to modelling of two-phase flow. Thermal-hydraulics of a horizontal steam generator in the WWER 1000 nuclear power plant has been simulated at the full load, steady-state operation. A comparison of the numerical results with data measured at the NPP Novovoronjezh shows good agreement. 3D numerical results can be used in plant design or retrofitting, in nuclear power plant operation and safety analysis and as improvement of existing one-dimensional thermal-hydraulics models of the horizontal steam generator which are assessed by system codes used for the nuclear power plant safety analyses. (author)

Energy Technology Data Exchange (ETDEWEB)

This project is to provide the regulatory direction of containment bypass during multiple steam generator tube failure issue for the Korean Next Generation Reactors, which is a part of major technical issues resulted from the safety regulation R and D on the KNGR. The outstanding results are as follows : the Multiple Steam Generator Tube Repture(MSGTR) event has never been occurred in the history of commercial nuclear reactor operation but single Steam Generator Tube Rupture(SGTR) event is reported to occur every two years. A probabilistic safety analysis study on MSGTR event, however, show its probability of occurrence is to be the same order as the design basis accidents such as LACA. In this regard, the ability of NPPs to cope with MSGTR event is required. Some requirements on initial and boundary conditions are suggested to be used in ...

Energy Technology Data Exchange (ETDEWEB)

The work reported is to evaluate two alternative resource recovery systems which would produce a fuel or an energy product (steam) for sale to an industrial plant. The first alternative is a refuse derived fuel (RDF) production facility which would process waste to produce fuel for sale to the plant. The RDF would be co-fired with coal in new boilers. The second alternative is a mass burn steam generation facility which would produce high pressure superheated steam for sale. The steam would be used by the plant to cogenerate electricity and for the process. Conceptual designs are developed for each alternative to form the basis of estimates of capital and operating costs and potential energy revenues. Operating requirements are developed for both alternatives. An assessment is made of the impact of both alternatives on the environment, and it is concluded that the impact of either ...

International Nuclear Information System (INIS)

The main reasons of tube failures in steam generators (SG) are considered. 1.Stress corrosion craining which has 28% of SG (most of them have stainless steel tubes). 2. Corrosion loss of metal, which accounts for 24% of tubes (phosphate corrosion due to addition of PO into water). 3.Denting-peripheral pressing of tubes in the openings of the foundation plates by the corrosive products, which are formed on internal surface of drillings in the foundation plates made of carbon steel. 4.Separation of a plating layer on tube panels. 5.Dratting-corrosion. 6.Metal fatigue. A series of experiments were conducted to study the influence of material selection on tube reliability (stainless steel 304, inconel-600, mone-400, incalloy-800). The problem of increase of SG elements reliability is a complex one and can be solved by direct selection of material, proper control of water-chemical conditions and other measures of corrosion prevention such as direct ...

British Library Electronic Table of Contents (United Kingdom)

The rational use of energy has become a priority for all industries in Brazil, mainly after the energy rationing in 2001. Methodologies to quantify and improve the performance of plants that consume and generate electricity and thermal energy are being used to reach this goal. Exergoeconomic analysis provides a complete diagnosis of a plant, both in exergetic and in monetary values. This study shows the methodology used to assess the power generation system of the Companhia Siderurgica Tubarao (CST). The current system is based on a regenerative Rankine cycle using two gases from steel production - blast furnace gas (BFG) and coke oven gas (COG) - to generate electric power and occasionally steam for the process. Use of the Theory of Exergetic Cost allowed determination of monetary and exe...

International Nuclear Information System (INIS)

Solid solution-strengthened nickel base alloys have been used for nuclear stream generator tubing in pressurized water reactors since the beginnings of commercial nuclear power. The purpose of this paper is to recap and update the authors understanding of the relationship between processing, resulting structure, and properties for Alloy 600 and to discuss the requirements for optimized performance in both primary and secondary environments. Potential replacement materials and their performance will be discussed. Also discussed is the role and importance of system chemistry, bulk and local, and control and its relationship to performance. A discussion of potential mechanisms of environmentally assisted failure is also discussed.

International Nuclear Information System (INIS)

Tube supports used in the tube bundles of PWR steam generators have consisted of mechanical devices located at intervals along the tube bundle. The presence of tube supports creates regions of restricted flow with altered flow patterns and increased pressure drop. An additional and very important effect is also the possibility of local complete vaporization or dryout occurring in the tube/support flow passage and crevices. The thermal/hydraulic conditions at which dryout occurs are of particular interest because of the possibility of the deposition of dissolved solids with the occurrence of dryout. As long term build-up of solid deposition could have a deleterious effect, knowledge of the conditions at which dryout occurs would possibly provide a means to avoid this build-up. A test program, sponsored by the Steam Generator Project Office of the Electric Power Research Institute, was conducted to ...

Energy Technology Data Exchange (ETDEWEB)

Restructured power markets require concepts not only for extending the lifetime of existing plants, but also concepts for reducing the life-cycle costs of existing older steam power plants in order to remain competitive with the most modern and advanced power generation technology such as combined cycle power plants. Out of the broad range of methods to extend service life and increase availability, and to increase power output and efficiency, this paper focuses on full repowering. Full repowering replaces the old fired boiler with gas turbine(s) and heat-recovery steam generators (HRSG) as heat input to the existing bottoming cycle and reuses the steam turbine (ST) condenser and further balance of plant (BOP) equipment. The potential offered by this concept is demonstrated. Repowering and modernization projects are executed if they result in economic benefits for the operator. ...

Energy Technology Data Exchange (ETDEWEB)

This volume on specifications for the Saguaro Power Plant includes the following: subsystem interface definition document; solar collector subsystem specification; receiver specification; thermal energy storage specification; solar steam generator specification; and master control system specification.

Energy Technology Data Exchange (ETDEWEB)

The special geodetic problems encountered during the construction of the Paks nuclear power plants are treated. The main building with its hermetically connected components including the reactor, the steam generators, the circulation pumps etc. impose special requirements on the control net of datum points. The geodesy tasks solved during the construction of the main building are presented in details.

Energy Technology Data Exchange (ETDEWEB)

Some 200 visitors attended the official opening of a new diesel engine driven cogeneration plant in the city of Hamburg at the end of March. The dual fuel engine built by Blohm+Voss will supply 7.7 MW of electricity to the grid and 8.26 MW of thermal power to generate steam and hot water. (author)

Energy Technology Data Exchange (ETDEWEB)

The operating histories and performance of a large (180 MWsub(e)) central condensing plant and five small (5 MWsub(e)) wellhead backpressure power plants are presented. The thermodynamic performance of the central plant is compared with an estimated possible performance based upon the work of Michaelides. Both types of plant have been shown to be reliable despite some problems due to the contaminants which are associated with geothermal steam.

Energy Technology Data Exchange (ETDEWEB)

The device, aimed at non destructive control of the inner side of tubes such as steam generator tubes, is composed of a control sensor mounted on a support; the sensor head may rotate in the tube and a measuring system and signal processing allow for the exact determination of the angular position of the sensor head (application to ultrasonic or eddy current probes).

Energy Technology Data Exchange (ETDEWEB)

The oxidation of boron carbide in steam or air was recently extensively studied especially in Forschungszentrum Karlsruhe, Institut fuer Materialforschung. An important data set is available for the interaction modelling. An oxygen diffusion model through the superficial liquid boron oxide formed on the boron carbide external surface associated to a superficial reaction between the liquid boron oxide and steam is proposed to simulate the experimental kinetics from BOX rig and thermogravimetric tests on the interaction between steam and boron carbide at a temperature range 800 C to 1400 C. The oxygen diffusion model will be also useful to simulate interaction between boron carbide and Ar+O2 (air simulation) atmosphere when the steam pressure becomes zero. From the analysis of BOX rig experimental kinetics of non-condensable (H2, CO2, CO and CH4) gases we propose an oxygen diffusion model through the ...

Energy Technology Data Exchange (ETDEWEB)

This article reports on the world's first gas turbine to top 40% thermal efficiency in simple cycle and is now operating commercially as part of a natural-gas-fired combined-cycle cogeneration (CC) plant serving the Ottawa Health Sciences Center, Ottawa, Ont, Canada. Owned and operated by TransAlta Energy Corp, Calgary, Alta, the CC plant supplies steam and hot and chilled water to the health center and another thermal user while the electricity is sold to Ontario Hydro. Part of the project included the conversion of 15-psig steam-carrying equipment to hot-water service. Commercial success of this installation heralds a new era in the application of efficient gas turbines for power generation.

Energy Technology Data Exchange (ETDEWEB)

Computer programs have been written to allow the analysis of different types of eddy-current probes and their performance under different steam generator test conditions. The probe types include the differential bobbin probe, the absolute bobbin probe, the pancake probe and the reflection probe. The generator test conditions include tube supports, copper deposits, magnetite deposits, denting, wastage, pitting, cracking and IGA. These studies are based mostly on computed values, with the limited number of test specimens available used to verify the computed results. The instrument readings were computed for a complete matrix of the different test conditions, and then the test conditions determined as a function of the readings by a least-squares technique. A comparison was made of the errors in fit and instrument drift for the different probe types. The computations of the change in instrument reading due to the defects have ...

International Nuclear Information System (INIS)

In a Pb-Bi cooled direct contact steam generation fast reactor water is injected directly above the core, the produced steam is separated at the top and is send to the turbine. Neither the direct contact phenomenon nor the two-phase flow simulations in CFD have been thoroughly described yet. A first attempt in simulating such two-phase flow in 2D using the CFD code Fluent is presented in this paper. The volume of fluid explicit model was used. Other important simulation parameters were: pressure velocity relation PISO, discretization scheme body force weighted for pressure, second order upwind for momentum and CISCAM for void fraction. Boundary conditions were mass flow inlet (Pb-Bi 0 kg/s and steam 0.07 kg/s) and pressure outlet. The effect of mesh size (0.5 mm and 0.2 mm cells) was investigated as well as the effect of the turbulent model. It was found that using a fine mesh is very important in order ...

Energy Technology Data Exchange (ETDEWEB)

Once again Brazil faces the possibility of an electricity shortage in the near future. The low level of rain and the consequent reduction of hydroelectric reservoirs (hydroelectric energy represents almost 80% the total Brazilian electricity generation) led to the dispatch of thermal plants based on natural gas and/or fuel oil. However, the recent crisis regarding the Bolivian natural gas supply to Brazil increased the risk of a shortage of this energy source. Hence, the uncertainty of the electricity supply scenario is pushing the electricity prices up in the near future. A possible alternative to increase the electricity supply is the maximization of the electricity surplus in sugarcane mills. Back in 1970's and 1980's, cogeneration plants in sugarcane mills were primarily designed to consume all bagasse, and produce steam and electricity to the process. The plants used medium pressure steam boilers (21 ...

Energy Technology Data Exchange (ETDEWEB)

The Externally-Fired Combined Cycle (EFCC) is an attractive emerging technology for powering high efficiency combined gas and steam turbine cycles with coal or other ash bearing fuels. In the EFCC, the heat input to a gas turbine is supplied indirectly through a ceramic air heater. The air heater, along with an atmospheric coal combustor and ancillary equipment, replaces the conventional gas turbine combustor. A steam generator located downstream from the ceramic air heater and steam turbine cycle, along with an exhaust cleanup system, completes the combined cycle. A key element of the EFCC Development Program, the 25 MMBtu/h heat-input Kennebunk Test Facility (KTF), has recently begun operation. The KTF has been operating with natural gas and will begin operating with coal in early 1995. The US Department of Energy selected an EFCC repowering of the Pennsylvania Electric Company`s Warren Station for ...

Energy Technology Data Exchange (ETDEWEB)

A CANDU nuclear reactor was shut down for over one year because steam generator (SG) tubes had failed with outer diameter stress-corrosion cracking (ODSCC) in the U-bend section. Novel, single-pass eddy current transmit-receive probes, denoted as C3, were successful in detecting all significant cracks so that the cracked tubes could be plugged and the unit restarted. Significant numbers of tubes with SCC were removed from a SG in order to validate the results of the new probe. Results from metallurgical examinations were used to obtain probability-of-detection (POD) and sizing accuracy plots to quantify the performance of this new inspection technique. Though effective, the above approach of relying on tubes removed from a reactor is expensive, in terms of both economic and radiation-exposure costs. This led to a search for more affordable methods to validate inspection techniques and procedures. Methods are presented for calculating POD curves ...

Energy Technology Data Exchange (ETDEWEB)

The effect of accelerated thermal cycling on a joint between modified 9Cr-1Mo steel (Grade 91) and Alloy 800 welded with Inconel 82 and 182 filler material is discussed. This is part of a trimetallic transition joint involving Grade 91-Alloy 800-316LN austenitic stainless steel for steam generator application. It has been shown that, during thermal cycling following the typical post-weld tempering treatment at 760 deg. C for 2 h, no carbon diffusion occurs from the ferritic steel towards the weld metal. There is, in fact, a hardness increase at the ferritic steel/weld metal interface which is probably a result of work hardening. Carbon migration sets in only after unusually long post-weld heat treatments for 20 and 50 h at 760 deg. C followed by thermal cycling. Significantly, even under the most severe thermal cycling test conditions imposed, no cracking or oxide notching could be detected, thus demonstrating the superior performance potential ...

International Nuclear Information System (INIS)

Methodology based on probabilistic fracture mechanics (PFM) is finding increasing acceptability in demonstrating safety of Nuclear Power Plant (NPP) piping. In PFM, the methods of fracture mechanics and reliability theory are combined for assessing the reliability of components, which contain cracks. In this work, reliability assessment of Tarapur Atomic Power Plant (TAPP) 3-4 Primary Heat Transport (PHT) piping is done using PFM. Monte Carlo simulation with stratified sampling is used as a variance reduction technique. PFM model assumes a pre-existing circumferential surface crack before the start of plant operation. The crack grows in size during the lifetime of the plant due to the fatigue loading. This part-through wall crack having escaped hydro-test and pre-service inspection, may result in either a through wall flaw (leak) or may lead to the rupture of the piping. R6 method is used as failure criteria. Steam generator inlet (SGI), ...

Science.gov (United States)

Metal cations of arsenic, antimony, tin, manganese, zinc, cadmium, indium, and thallium have been evaluated in a preliminary way as possible3 inhibitors for controlling denting corrision observed in steam generators used with pressurized water reactors (PWR). The rationale for this approach was based upon the well-known inhibition effects of metal cations on corrosion rates in electrolyte/metal systems. A review of corrosion inhibition by metal cations (H. Leidheiser, Jr., Corrosion 36, 339 (1982)) has identified eleven inhibition mechanisms. The major test methods used for this evaluation were: (1) Isothermal capsule tests of carbon/steel/Inconel 600 tube bulging rates at temperatures up to 288/sup 0/C in seawater/copper-nickel chloride bulge-accelerating solutions. (2) Immersion weight-loss tests of steel coupled to Inconel 600 in boiling (102/sup 0/C) 3% sodium chloride solutions. In addition, electrochemical measuremens and surface analyses ...

Energy Technology Data Exchange (ETDEWEB)

The purpose of this paper is to assess a new local air coolers model in ISAAC 2.0, as ISAAC 1.0 could model local air coolers only at two locations. In the new model, local air coolers up to twelve locations could be handled. Large LOCA and loss of feed water sequences were selected for the model comparison. Two cases were analyzed with ISAAC 2.0: one with 6 local air coolers in one of the fueling machine room and in the steam generator room, respectively, and the other with 3 local air coolers at both fueling machine room and 6 local air coolers in the steam generator room. The study assumes that the safety systems such as emergency core cooling system, shield cooling system and moderator cooling system are unavailable. According to the ISAAC 2.0 results, the new local air coolers model showed almost no difference between two cases. Also it was found that as the location of LACs increased, the new ...

International Nuclear Information System (INIS)

To raise the efficiency of the redox method and decrease the amount of radioactive wastes, a possibility of improving the decontamination process for NPP heat exchanger tubes made of stainless steel is studied. In the home practice the redox method of equipment decontamination is carried out as a multi-cycle process. In each cycle the surface is treated first with a permanganate alkaline solution ther with an oxalic acid solution, with a condensate washing-at between the treatments. Using samples cut out of the steam generator pipelines of the first and third power units of the Novovoronezh NPP the effect of the oXalic acid concentration, as well as washout time and conditions on the decontamination factor are studied. On the basis of analysis of the obtained data a conclusion is drawn that using oxalic acid of low concentrations and increasing its concentrations from cycle to cycle maximum decontamination factor values can be obtained at a ...

International Nuclear Information System (INIS)

Denting studies have been undertaken in order to assess the influence of the most important parameters which could initiate corrosion of the carbon-steel occurring in the tube-tube support plate crevices of some PWR steam generators. Tests have been carried out in model boilers, feedwater being polluted with sea or river water. Specific effect of chloride or sulfate and influence of oxygen content, magnetite addition and pH value were investigated. In magnetite prepacked crevices, denting is obtained within 1000 hours for sea-water pollution of 0.3 ppm chloride at the blowdown. In neutral chloride or in river water denting is observed only with oxygen addition. Denting prevention is effective in the case of an on-line addition of phosphate, boric acid or calcium hydroxide. For denting stopping, boric acid or calcium hydroxide is efficient even with a high sea-water pollution. Soaks cannot stop denting if they are not followed by an on-line ...

Energy Technology Data Exchange (ETDEWEB)

Steam generator tubes in nuclear power plants are periodically checked by means of eddy current probes. The output of a probe is composed of three types of signals: known events (rolling zone, support plates, U-bend part), noise (mainly metallurgical noise) and possible flaws. The latter are random transients, both in arrival time and in shape: they have to be detected and then estimated, before to be fed to the high level stages of a diagnostic system. The objective of the study presented is to develop a semi-automatic system, which could manage and process more than 1 M-bytes of data per tube and provide an operator with reliable diagnostics proposals within a few minutes. This can be achieved only by cooperation of several digital signal processing techniques: detection, segmentation, estimation, noise subtraction, adaptive filtering, modelization, pattern recognition. The paper describes some of these items.

Energy Technology Data Exchange (ETDEWEB)

Constant extension rate (CERT) tests were performed on Alloy 690 and Alloy 800 nuclear steam generator tubing specimens. For the Alloy 690 specimens, tests were performed in deaerated 10% sodium hydroxide solutions at 315 deg C with a +100 mV applied potential. For the Alloy 800 specimens, tests were performed in deaerated 5% sodium hydroxide solutions at 343 deg C with no applied potential. The test specimens were machined from tubing which was produced by different manufacturing processes and in different heat treated conditions. The Alloy 690 tubing was tested in six different thermomechanical conditions, while the Alloy 800 tubing was tested in four different thermomechanical conditions. The results from the test program include a complete microstructural examination using light-optical and scanning electron microscopy. The CERT test results (such as maximum stress achieved and crack morphology) are correlated to tubing microstructure, ...

Energy Technology Data Exchange (ETDEWEB)

In the frame of the use of the Relap5 thermal hydraulic code in the predictions of LOCA transient scenarios in PWRs and considering the recent development of a methodology to evaluate the related uncertainty, the response to a Small Break LOCA of Eastern and Western type PWRs has been analyzed. A four loop/horizontal Steam Generator WWER-1000 (KOZLODUY in Bulgaria) and a two loop/vertical U-tubes Steam Generator Westinghouse (KRSKO in Slovenia) nuclear power plants have been considered in the analysis. The reference transient is a 2% equivalent cold leg break accident, without High Pressure Injection System intervention, as specified in the frame of a ``counterpart test`` activity involving experimental tests on four Integral Test Facilities: LOBI (European Community), SPES (Italy), BETHSY (France) and LSTF (Japan). The code results in the two cases, also taking into account the related uncertainty as ...

International Nuclear Information System (INIS)

Rapid corrosion of PWR steam generator carbon steel support structures and consequential denting of steam generator tubes led to considerable investigation of the mechanism of corrosion and for palliatives. The basic mechanism is now fairly well understood. Chloride present in the boiler feedwater from condenser leakage is concentrated in the support plate crevice as a result of the superheat present. This leads to the formation of a low pH acid ferrous chloride environment either through the hydrolysis of a weak base chloride such as MgCl_2 or the combined action of a neutral chloride with an oxidizing agent. Rapid attack of the carbon steel ensues with the Fe_3O_4 corrosion product forming at the metal/oxide interface. This oxide has a volume approximately twice that of the steel consumed and thus eventually fills the crevice between the tube and the support plate. Oxide growth continues leading to ...