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 the heat resistance of ...

Energy Technology Data Exchange (ETDEWEB)

Heat transfer experiments were performed in the post-CHF two-phase flow regime in a vertical tube. The tube inside diameter was 7.75 mm, and the boiling fluid was R-113. The experiments were performed at steady state by means of liquid heating of the test tube. Wall superheats were maintained below 70 C for heat exchanger/steam generator application. The mass flux range of the data was 379-816 kg m{sup -2} s{sup -1}. The use of R-113 significantly extended the property range of the existing low wall-superheat data base. Experimental data are presented in tabular as well as graphical form, and the results were used with low wall-superheat data from other fluids to add generality to a predictive heat transfer correlation. (orig.)

Energy Technology Data Exchange (ETDEWEB)

Full text of publication follows: A porous media solution PORFLO has been developed for the 3-dimensional two-phase flow by describing the process facility in Cartesian or cylindrical coordinates. The local porosity fraction is applied for distinguishing the fluid filled volumes from the solid structures. The solid structure contribute the two-phase flow through the wall friction, flow area and heat transfer. Optionally the solid structure may contain primary liquid of steam generators, steam in the higher temperature and pressure to be condensed or electrical heating power. By using these optional boundary conditions three different process facilities have been analysed. The thermohydraulic solution based on 5-equation approach, where the conservation equations are solved for the liquid and gas (vapour) mass, mixture momentum (giving the velocity only for the mixture), liquid and gas energy, is described shortly. In ...

International Nuclear Information System (INIS)

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"I"s"c"c, 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 ...

International Nuclear Information System (INIS)

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 ...

British Library Electronic Table of Contents (United Kingdom)

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)

Solar cavity receiver plays a dominant role in the light-heat conversion. Its performance can directly affect the efficiency of the whole power generation system. A combined calculation method for evaluating the thermal performance of the solar cavity receiver is raised in this paper. This method couples the Monte-Carlo method, the correlations of the flow boiling heat transfer, and the calculation of air flow field. And this method can ultimately figure out the surface heat flux inside the cavity, the wall temperature of the boiling tubes, and the heat loss of the solar receiver with an iterative solution. With this method, the thermal performance of a solar cavity receiver, a saturated steam receiver, is simulated under different wind environments. The highest wall temperature of the boi...

Energy Technology Data Exchange (ETDEWEB)

Real time neutron radiography has been used to study the dynamic behavior of two phase flow and measure the time averaged vapor fraction in a heated metal tube containing boiling steam water operating at up to 15 bar pressure. The neutron radiographic technique is non-intrusive and requires no special transparent window region. This is the first time this technique has been used in an electrically heated pressurized flow loop. This unique experimental method offers the opportunity to observe and record on videotape, flow patterns and transient behavior of two phase flow inside opaque containers without disturbing the environment. In this study the test sections consisted of stainless steel tubes with a 1.27 cm outer diameter and wall thicknesses of 0.084 cm and 0.124 cm. The experiments were carried out at the Pennsylvania State University 1 megawatt TRIGA reactor facility utilizing a Precise Optics ...

International Nuclear Information System (INIS)

Real time neutron radiography has been used to study the dynamic behavior of two phase flow and measure the time averaged vapor fraction in a heated metal tube containing boiling steam water operating at up to 15 bar pressure. The neutron radiographic technique is non-intrusive and requires no special transparent window region. This is the first time this technique has been used in an electrically heated pressurized flow loop. This unique experimental method offers the opportunity to observe and record on videotape, flow patterns and transient behavior of two phase flow inside opaque containers without disturbing the environment. In this study the test sections consisted of stainless steel tubes with a 1.27 cm outer diameter and wall thicknesses of 0.084 cm and 0.124 cm. The experiments were carried out at the Pennsylvania State University 1 megawatt TRIGA reactor facility utilizing a Precise Optics ...

Energy Technology Data Exchange (ETDEWEB)

Conventional eddy current bobbin probes are known to be ineffectual in detecting circumferential cracks in tubing. Multi-pancake and/or rotating pancake probes are required to detect circumferential cracks. It has recently been demonstrated in CANDU steam generator tubes, with deformation, ferromagnetic deposits, and copper deposits, that multi-channel probes with transmit-receive (T/R) coils are superior to those using surface impedance coils. Unlike rotating probes, the design of a new probe denoted as C3 permits fast, single-scan inspection of a full-length tube at inspection speeds comparable to conventional bobbin probes. Since 1992, the probe has been used routinely for steam generator in-service inspection at 4 CANDU plants. Defective tubes have been plugged and the units returned to service, and they continue to operate without leaks. This paper ...

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 ...

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.

Science.gov (United States)

We present the results obtained from tests and studies carried out on the model of tube bundles for a PGV-1000 horizontal steam generator that were conducted for experimentally substantiating the design service life of a steam generator tube bundle intended for use at new nuclear power stations equipped with a PGV-1000MKP steam generator. Measures taken to minimize the incipience and development of local corrosion damage to the heat-transfer tubes and ensure their design service life are substantiated and confirmed.

Energy Technology Data Exchange (ETDEWEB)

A sodium-water reaction drove from the single tube break in steam generator of FBR might overheat labor tubes rapidly under internal pressure loadings. If the temperature of tube wall becomes too high, it has to be evaluated that the stress of tube does not exceed the material strength limit to prevent the propagation of tube rupture. This study clarified the tensile and creep properties of Mod.9Cr-1Mo steel at ultra-high temperature which will be used in evaluation of the tube burst by sodium-water reaction. The strain rates for tensile test are from 10%/min to 10%/sec, and creep-rupture time is maximum 277sec. The range of test temperature is 700degC to 1300degC. The main results obtained were as follows; (1) The evaluation data on the relationship between tensile strength and strain rate and creep-rupture strength in ...

International Nuclear Information System (INIS)

Zircaloy fabrication plant manufactures all the necessary Zr-2 components like fuel canning tubes, calandria tubes and other rod and sheet products. This plant is having a capacity of producing about 4 lakh nos. of PHWR fuel tubes per annum. These tubes are seamless, thin walled with close dimensional tolerances and stringent mechanical properties. The plant has established all the facilities required to produce these tubes with required quality.

Energy Technology Data Exchange (ETDEWEB)

Eddy current testing (ECT) method is widely used as the non-destructive evaluation (NDE) of the various kinds of material degradation occurring in nuclear power plants (NPPs) components including steam generator (SG) tubes. Detection of any defects or flaws in the steam generator tubes in the early stage is very important in maintenance of NPP for its primary role as the pressure boundary with thin wall thickness. Although the ECT technique provides lots of information for a SG management, it has a generic problem in its reliability due to a low ability in detect of small defects and some difficulty in a signal analysis. For the improvement of these shortcomings in conventional ECT, profile-MRPC (motorized rotating pancake coil) ECT technology was developed in KAERI. The key of this new technology is the development of new eddy current probes, designated as a diagnostic probe ...

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 water reactors, e.g. Wolsong 2, 3, and 4 ...

International Nuclear Information System (INIS)

It is one of the most important subjects to estimate the flow-related stability of the heat exchanger tubes. A large scale model steam generator has been developed to verify the stability of the tubes in the Japanese PWR steam generators for the two-phase flow-induced vibration and to accumulate related technical data of thermal-hydraulic and flow-induced vibration of U-bend tube bundle. The model steam generator has 230 U-bend tubes of 46 different radius and 5 columns for each of practical diameter and material, and the anti vibration bars are inserted into each spacing between tube arrays. The freon R123 has been used as the secondary side fluid in stead of water-steam two-phase. In the test, void fraction and interfacial velocities in U-bend and straight tube-bundle are ...

Energy Technology Data Exchange (ETDEWEB)

The device is made by a tool at the end of a flexible cable, a motor for the rotation of the flexible, a linear drive with a guide for it between the winder and the tube.

International Nuclear Information System (INIS)

A review of the performance of steam generator tubes in 116 water-cooled nuclear power reactors showed that tubes were plugged at 54 (46 percent) of the reactors. The number of tubes removed from service decreased from 4 692 (0.30 percent) in 1981 to 3 222 (0.20 percent) in 1982. The leading causes of tube failures were stress corrosion cracking from the primary side, stress corrosion cracking (or intergranular attack) from the secondary side and pitting corrosion. The lowest incidence of corrosion-induced defects from the secondary side occurred in reactors that have used only volatile treatment, with or without condensate demineralization.

Energy Technology Data Exchange (ETDEWEB)

In this work, the thermal performance of a conventional collector is improved by inserting porous substrates at the inner walls of the collector tubes. The porous substrates improve the convective heat transfer coefficient between the tube wall and the fluid. This improvement is investigated numerically and its effects on the efficiency and the useful gain of the collector are evaluated. It is found that inserting the porous substrate may raise the collector efficiency considerably, especially at high values of the overall heat loss coefficient.

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)

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 ...

British Library Electronic Table of Contents (United Kingdom)

Prospects for using new structural materials instead of Grade 08Kh18N10T steel for making heat-transfer tubes for horizontal steam generators with the purpose to increase their service life from 30?40 to 60 or more years are considered.

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 orientation is horizontal (the tube`s inlet and ...

International Nuclear Information System (INIS)

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)

Some examples are given of design adaptations, e.g. U-bend support, tube spacer grid, flow distribution baffle, triangular tube pitch, tube-tubeplate sheet connection, which reduce the tendency to various types of corrosion (stress corrosion cracking, tube denting, tube wastage, intergranular cracking). A basic diagram of the secondary circuit is shown. Oxygen, corrosion product and impurity control in the secondary circuit is described. (E.J.). 19 figs., 6 refs.

Energy Technology Data Exchange (ETDEWEB)

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)

Results of calculational investigations into circulation of water and steam-and-water mixture in the PGV-1000 steam generator heat exchanger bundle used at NPPs with the WWER-1000 reactors, are considered. Model of water circulation in horizontal steam generator with submerged heating surface under conditions of steam generation irregularity along the heat transfer tubes is made. On the basis of the obtained data the assumption is made about water essential overflows from the hot collector zone into the cold one. Overflow rate over the upper line of the heat transfer tubes may constitute 0.7 m/s. The conclusion is made about the necessity to set up the vertical barrier which divides hot and cold sections of heat transfer tubes and helps to avoid water transverse overflows.

Energy Technology Data Exchange (ETDEWEB)

The Superphenix 750-MW (thermal), once-throughtype steam generators are the result of over 10 years of intensive research and development. Detailed design and manufacture of the components lasted from 1977 to 1982. The main difficulties encountered during the construction of these prototype units concerned: demonstration of satisfactory high-temperature properties of Alloy 800 tubes; proven resistance of tube welds to thermal fatigue and of structures and shells to sodium/ water reaction effects; prevention of the tube bundle to flow-induced vibration; necessity of manual welding of shells to prevent hot cracking hazards related to the boron content of the steel; and welding of heat exchange tubes. None of these difficulties, however, invalidated initial major design choices, but provided a wealth of technical and technological experience and knowledge for the ...

Energy Technology Data Exchange (ETDEWEB)

'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 ...

Energy Technology Data Exchange (ETDEWEB)

In the course of designing a catalytic porous membrane reactor for experimental studies, both inside and outside of the non-reaction zones as well as the two ends of the membrane need to be completely sealed to ensure that there is no flow across the membrane in the non-reaction zone. Experiments show that up to 50% of the total flow across the membrane may be contributed by the axial flow along the wall of the non-reaction zones if only one side of the membrane is sealed. Another problem that cannot be solved by sealing is the capillary flow of the catalyst along the tube wall into the non-reaction zones when the catalyst is doped on the membrane. One of the best ways to avoid this axial flow of catalyst would be to use non-porous tubes in the non-reaction zones and join them with the porous membrane tube. In doing so, the cost of the membrane reactor could be reduced simply ...

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 orientation is horizontal (the tube's inlet ...

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 sensible to the peculiarities of the horizontal ...

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)

Design considerations are given for the secondary side region of a vertical U-tube nuclear stream generator with an integral preheater. The thermal shield design, the novel recirculating water flow distribution scheme, the high porosity tube supports used in the parallel flow regions, and the U-bend supports are discussed for the Wolsung Plant steam generators. Experimental and analytical development programs undertaken to verify the design features are outlined.

International Nuclear Information System (INIS)

The tube bundle arrangement is suitable as reheating section of a steam generator in combination with a gas-cooled reactor. For any high-temperature section with a straightline type tube bundle, there is one inlet and one outlet collecting main; these two are arranged one opposite to the other and diametrically to the tube bundle. There are a number of lines from the inlet collecting main to the outlet collecting main; these lines are arranged axially one on top of the other and run around the tube bundle helically several times. (GL) 891 GL/GL 892 MB.

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 so far as possible, the materials choices ...

Science.gov (United States)

Numerical solutions for fully developed laminar flow in internally finned tubes with trapezoidal and triangular fin profiles were given with Finite Element Method (FEM): The heat transfer characteristics were obtained and compared under the boundary conditions of uniform heat flux, uniform wall temperature, and the third boundary condition with finite wall thermal conductivity considered. The numerical results show that boundary conditions have pronounced effects on the temperature field. Furthermore, a new mechanism on the heat transfer augmentation of internally finned tubes is proposed.

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 influence on the corrosion behaviour during the whole exploitation period of this equipment. (R.P.)

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.

Energy Technology Data Exchange (ETDEWEB)

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 velocities. With these ...

Energy Technology Data Exchange (ETDEWEB)

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 ...

Energy Technology Data Exchange (ETDEWEB)

Accumulation of Chloride and Cupric ions under the sludge on the steam generator tubing surface in the area between tube sheet and the first tube support is known to be the main reason for the pitting corrosion. In addition to those ions, oxygen, pH, temperature, heat treatment history, contests of minor elements such as C, S or Ti in the alloy are also influencing the material's the pitting corrosion resistance. a series of autoclave pitting test was conducted to evaluate the pitting corrosion characteristics of the domestically produced alloy 690 tubing material and compare with that of INCO manufactured material. 16 refs., 44 figs., 5 tabs. (Author)

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 the perforated sheet on the thermo-hydraulic processes in the ...

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 ...

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 parameters are determined in ...

Energy Technology Data Exchange (ETDEWEB)

The shot peening of the outer surface of Alloy 800 is proposed by the steam generator supplier in order to increase the stress corrosion cracking resistance of the tubes. The benefits and drawbacks of such a treatment are evaluated, focusing on the main concern i.e. a possible enhanced tendency of stress corrosion cracking at the transition of the expanded tube-to-tubesheet joint. Stress corrosion tests as well as residual stresses measurements demonstrate no significant difference between peened and unpeened tubes at that location. The other concerns such as stability of the compressive stresses, increased risk of pitting or intergranular attack and pollution of the secondary circuit by fragmented beads appear the as minor concerns.

Energy Technology Data Exchange (ETDEWEB)

The presentation summarises the highlights of experimental results obtained for VVER type horizontal steam generator heat transfer, primary side flow pattern, and mixing in the hot collector during secondary side boil-off with primary at single-phase natural circulation. The experiments were performed using the PACTEL facility with Large Diameter (LD) steam generator models, with collector instrumentation designed specifically for these tests. The key findings are as follows: (1) the primary to secondary heat transfer degrades as the secondary water inventory is depleted, following closely the wetted tube area; (2) a circulatory flow pattern exists in the tube bundle, resulting in reversed flow (from cold to the hot collector) in the lower part of the tube bundle, and continuous flow through the upper part, including the tubes that have already dried out; and ...

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 concentration kinetics of caustic which was ...

International Nuclear Information System (INIS)

An eddy current inspection has been applied for a pre-service and in-service examination of a steam generator in nuclear power plants. The experience from the inspection of steam generators showed that many plants had an excessive number of tubes with eddy current noise signals over several hundreds, which originated from manufacturing anomalies. The plants in U.S suffered significant downstream inspection costs, history reviews, and diagnostic testing because some signals resembled flaws and others masked a flaw. These lessens learned resulted in issuing the guidelines for steam generator tubing specifications and repair, in order to reduce the number of anomalous signals in the tubes and also to provide the requirement of a signal to noise ratio by applying a field type examination with bobbin coil eddy current probes at a manufacturing process. Besides the ...

Energy Technology Data Exchange (ETDEWEB)

Hollow needles made of an electrically conducting material are attached by hinges in the openings of the perforated end of an insert in order to intensify heat and mass transfer. The free sections of the needles are placed outside the insert, and a wick is placed on the wall of the frame in the condensation area. The wick overlaps the inlet openings of the insert.

Energy Technology Data Exchange (ETDEWEB)

A safety analysis code to design a pressurized water reactor and to obtain the licenses including entire proprietary rights is under development in domestic R and D project. The tasks of KAERI is to develop the constitutive relations including models for defining flow regimes and flow regime related models for inter-phase friction, wall frictions, wall heat transfer, and interphase heat and mass transfer in the two-phase three-field equations. In this paper, the process will be presented for choosing the best flow regime maps which occur in gas-liquid two-phase flow in horizontal and vertical tubes.

International Nuclear Information System (INIS)

The development of new electro-optical instrumentation for studying the annular dispersed two-phase flow regime is described. The system measures the thickness of the water film and droplet size and velocity distributions which would be encountered in such a flow regime. The water film thickness is measured by an improved capacitance method with a short time constant using newly developed sensor electrodes. The electrodes are made flush with the inner wall of a cylindrical tube and do not disturb the flow. In the test equipment, steady, laminar flow of water along the inner wall of the tube is controlled by appropriate valves and a porous jacket while droplets are introduced by means of a special spray nozzle.

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)

Flow Accelerated Corrosion (FAC) of Carbon Steel (CS) piping has been one of main issues in Light-Water Nuclear Reactor (LWRs). Wall thinning of CS piping due to FAC increases potential risk of pipe rupture and cost for inspection and replacement of damaged pipes. In particular, corrosion products generated by FAC of CS piping brought steam generator (SG) tube corrosion and degradation of thermal performance, when it intruded and accumulated in secondary side of PWR. To preserve SG integrity by suppressing the corrosion of CS, High-AVT chemistry (Feedwater pH9.8#+-#0.2) has been adopted to Tsuruga-2 (1160 MWe PWR, commercial operation in 1987) in July 2005 instead of conventional Low-AVT chemistry (Feedwater pH 9.3). By the High-AVT adoption, the accumulation rate of iron in SG was reduced to one-quarter of that under conventional Low-AVT. As a result, a tendency to degradation of the SG thermal efficiency was improved. On ...

International Nuclear Information System (INIS)

There are a total of nine operational nuclear plants in Spain totalling 7.350 MWe. These units produced 54.265 x 106 KWh in 1990, 36% of the total generation in Spain. Seven of these plants are of the PWR type. The first plant in operation was Jose Cabrera (ZORITA) in 1968, one loop Westinghouse plant with a model 24 Steam Generator. Due to the design margin and careful operation of the Steam Generator of this plant its performance have been very good, with only 5% tubes plugged after 23 years of operation. This is one of the few units in the world that remains in phosphate chemistry. During the period 1981-1985 a total of four units, two in Almaraz and two in Asco entered in operation. These three loop s Westinghouse units use model D-3 preheater Steam Generators. The poor design and manufacture of the Steam Generators of these units have caused a large number of problems: ...

International Nuclear Information System (INIS)

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.).

Energy Technology Data Exchange (ETDEWEB)

Alloy 800 has been used for steam generator (SG) tubing for more than 30 years, primarily in CANDU reactors 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 specified and appropriate operating conditions. In planning refurbishment of CANDU 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 estimates it based on experimental data specific to CANDU operating conditions. The paper presents a more advanced 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 ...

International Nuclear Information System (INIS)

EPRI's Steam Generator Strategic Management Project issues an annual report on the progress being made in tackling steam generator problems worldwide, containing a wealth of detailed statistics on the status of operating units and degradation mechanisms encountered. A few highlights are presented from the latest report, issued in October 1993, which covers the period to 31 December 1992. (Author).

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.

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 ...

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 ...

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.

Energy Technology Data Exchange (ETDEWEB)

This article describes innovations in new-boiler startups, based on experiences at United Development Group`s 50-MW Niagra cogeneration facility, Niagra Falls, NY. The plant comprises: a circulating coal fluidized bed boiler supplying steam to a nearby factory and electricity to the grid. Before operation the system was flushed with demineralized water, and the boiler degreased, the steam blow relied on a new procedure involving a continuous flow of steam. Startup was then initiated, following manufacturers heatup rate and soak times closely. After startup boiler tube sections were checked, and cleaned if necessary. 1 fig.

Energy Technology Data Exchange (ETDEWEB)

Experiments are performed to investigate the single-phase flow and flow-boiling heat transfer augmentation in 3D internally finned and micro-finned helical tubes. The tests for single-phase flow heat transfer augmentation are carried out in helical tubes with a curvature of 0.0663 and a length of 1.15 m, and the examined range of the Reynolds number varies from 1000 to 8500. Within the applied range of Reynolds number, compared with the smooth helical tube, the average heat transfer augmentation ratio for the two finned tubes is 71% and 103%, but associated with a flow resistance increase of 90% and 140%, respectively. A higher fin height gives a higher heat transfer rate and a larger friction flow resistance. The tests for flow-boiling heat transfer are carried out in 3D internally micro-finned helical tube with a curvature of 0.0605 and a length of 0.668 m. Compared with that in ...

Energy Technology Data Exchange (ETDEWEB)

The reliability of the surface condenser is a key factor in plant performance level and maintenance cost optimization. This is especially the case for thermal nuclear plants where condenser raw wa-ter ingress can introduce contamination into the chemically-controlled, steam/water loop potentially causing damage to sensitive equipment. Two important parameters must be taken into account when attempting to optimize the quality and the reliability of condenser tubing. They include selecting the appropriate material according to the cooling water corrosion level present. A second and equally important parameter is the manufac-turing of the tubing product itself. This paper will identify methods to optimize manufacturing processes and improve tubing quality, according to VALTIMET's 30 years of condenser welded tubing production experience. Those methods complete the core ...

Energy Technology Data Exchange (ETDEWEB)

Since the first commercial operation in 1983, Wolsung unit 1 has experienced several aging phenomena, especially in pressure tubes and feeder pipes. In case of pressure tubes, in-service inspections (ISI) revealed that a major portion of inspected tubes was in contact with calandria tubes. The likelihood of blister formation was a safety concern because it is a potential threat to pressure tube integrity. Wolsung unit 1 was, therefore, subjected to the SLAR (Spacer Location And Reposition) work to separate the contacted pressure tubes from calandria tubes. In this paper, experience of in-service inspection to the pressure tubes will be discussed including the irradiation creep elongation and CIGAR (Channel Inspection and Gauging Apparatus for Reactors) examination of pressure tubes. On the other ...

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)

Steam generators (SG) are among the most critical components of pressurized water Nuclear Power Plants (NPP). SG tubes must provide a reliable pressure boundary between the primary and secondary cooling water. It is because that any leakage from tube defects could result in the release of radioactivity to the environment. Thus degradations of steam generators tubes should be monitored and inspected periodically under nuclear regulatory. In-service inspections of SG tubes are carried out using eddy current test (ECT) and the defected tubes are usually plugged. Because the radioactivity in the internal of SG chambers limits free access of human worker, remote manipulators are required. In South Korea, Manipulators such as the Zetec SM series and the Westinghouse ROSA series have been used. Such manipulators are rigidly mounted to manways or ...

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 nodalization options were introduced. In the ...

Energy Technology Data Exchange (ETDEWEB)

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 ...

Energy Technology Data Exchange (ETDEWEB)

Two mechanical design requirements are defined for the SSC Collider beam tube. First, the vacuum requirement (luminosity lifetime = 150 hrs). It requires the design of a pressure boundary within the cold mass vessel to provide a vacuum tunnel for the proton beam and to minimize the synchrotron radiation gas desorbtion with a suitable material. The Collider beam tube design is under an intensive activity to search for a material that will meet the luminosity requirement without a distributed pump or liner. Second is the tube wall`s resistivity requirement ({sigma}*t = 2E5 {Omega}{sup {minus}1}). For a 4.2 K beam tube the Cu thickness is 100 {mu}m (RRR=30,6.7 T, {sigma}=2E9{Omega}{sup {minus}1}m{sup {minus}1}). The copper yield strength is relatively low in comparison to steel and, therefore, the design of the steel layer is governed by the copper layer yield stress limit. A beam ...

Energy Technology Data Exchange (ETDEWEB)

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.

International Nuclear Information System (INIS)

The objective of the C-E/EPRI project, ''Alternative Steam Generator Materials and Designs,'' was to evaluate the corrosion behavior of contemporary or alternative steam generator materials under prototypic design and secondary fault (high contaminant) water conditions. Two model steam generators built with various support materials and designs were tested under representative thermal and hydraulic conditions. One model operated under seawater faulted all-volatile treatment (AVT) secondary water chemistry conditions. The other model operated under acidified fresh water faulted AVT conditions. This presentation focuses on the tube support and tubesheet corrosion results obtained by destructive examination of both models.

Energy Technology Data Exchange (ETDEWEB)

The Alloy 800 tubes used in CANDU 6 steam generators have not experienced significant corrosion damage to date, which may be attributed to successful water chemistry control strategies. However, it is known that Alloy 800, like other steam generator (SG) tubing materials, is not immune to corrosion, especially pitting, under some plausible but off-specification operating scenarios. Electrochemical measurements provide information on corrosion susceptibility and rate, which are known to be a function of water chemistry. Using laboratory data in combination with chemistry monitoring and diagnostic software it is possible to assess the impact of plant operating conditions on SG tube corrosion for plant life management (PLIM). In this context, this paper discusses the results of electrochemical measurements made to elucidate the corrosion behaviour of Alloy 800 SG ...

International Nuclear Information System (INIS)

This report is a summary of the Steam Generator Group Project progress for 1985. Statistical analyses are presented of data from nondestructive examination (NDE) round robin tests that were performed on the Surry generator in 1984. In the first round robin, teams acquired and analyzed their own data on a selected 10% subset of tubes. In the second round robin, analysts from each participating team analyzed identical sets of data which were provided to them by the Project. Results from several examinations utilizing advanced/alternate NDE technologies including alternate eddy current probes and techniques, ultrasonics, and profilometry are also reported. Criteria are listed for selecting tube specimens that will be removed from the generator. A sampling plan based on these criteria is detailed graphically in the report. The initial steps that have been taken to carry out the plan are described; tooling fabricated for the ...

Energy Technology Data Exchange (ETDEWEB)

The authors investigate the optimisation of solar-heated cavity receivers in which solar energy collected by a paraboloid collector is converted into electrical energy. The focussed solar energy enters the receiver through an aperture to heat the working fluid, which flows inside the receiver tubes and is compressed by the compressor of a gas turbine system, to the highest process temperature. A method of calculating the distribution on the inner receiver surfaces of the solar radiation reflected by the paraboloid collector is presented to begin with. The interchange of radiant energy inside the receiver cavity can be calculated on this basis, including the radiation loads, the temperatures of the inner receiver walls, and the radiative and convective losses. The temperature fields in the tube walls must be known for determining the strength required of receiver tubes; they can be ...

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)

A test methodology for measuring temperature, impedance, pH, and electrochemical potential distributions within a sludge-packed tube support plate crevice in a laboratory test is described. The method successfully showed that there were large concentration gradients between the tube and tube support plate sides of the crevice. The testing also showed that strong bases concentrated more effectively than strong acids, and that the crevice pH, when exposed to seawater-based solutions, increased with increasing superheat and decreasing bulk concentration. The large variations in the crevice chemistry observed under heat transfer were eliminated upon shutdown. These new test data suggest that it might be beneficial to evaluate the variation in the extent of stress corrosion cracking with tube support plate elevation found in some steam generators in light of local chemistry changes, as ...

Energy Technology Data Exchange (ETDEWEB)

Results are presented from a numerical model of the steady-state energy transfer in molten-salt-in-tube solar cavity receivers that includes convective energy transfer at a local (spatially resolved) level. Molten salt energy absorption and gray radiative transfer between all cavity surfaces are also included. This model is applied to the Molten Salt Subsystem Component Test Experiment (MSS/CTE) cavity receiver. Results for this receiver indicate the global (entire cavity) receiver thermal efficiency is invariant within a few percent to most parameters investigated, although front surface temperatures of the nonabsorbing walls vary considerably, and are particularly sensitive to the type of convective submodel used. Absorption efficiencies indicate the effects of the cavity enclosure environment. For all conditions investigated, tube inner wall temperatures remain under 855 K, ensuring that the salt ...

Science.gov (United States)

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 {sup c}losure laws{sup .} Applied approach implies non-equilibrium thermal and flow conditions. The ...

Energy Technology Data Exchange (ETDEWEB)

Two mechanical design requirements are defined for the SSC Collider beam tube. First, the vacuum requirement (luminosity lifetime = 150 hrs). It requires the design of a pressure boundary within the cold mass vessel to provide a vacuum tunnel for the proton beam and to minimize the synchrotron radiation gas desorbtion with a suitable material. The Collider beam tube design is under an intensive activity to search for a material that will meet the luminosity requirement without a distributed pump or liner. Second is the tube wall's resistivity requirement ([sigma]*t = 2E5 [Omega][sup [minus]1]). For a 4.2 K beam tube the Cu thickness is 100 [mu]m (RRR=30,6.7 T, [sigma]=2E9[Omega][sup [minus]1]m[sup [minus]1]). The copper yield strength is relatively low in comparison to steel and, therefore, the design of the steel layer is governed by the copper layer yield stress limit. A ...

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.

Science.gov (United States)

This paper discusses the influence of an endoscope on the peristaltic flow of a couple stress fluid in an annulus under a zero Reynolds number and long wavelength approximation. The inner tube is uniform, rigid, while the outer tube has a sinusoidal wave traveling down its wall. Analytical expressions for the axial velocity, stream function and axial pressure gradient are established. The flow is investigated in a wave frame of reference moving with the velocity of the wave. Numerical calculations are carried out for the pressure rise, frictional forces and trapping. The features of the flow characteristics are analyzed by plotting graphs and discussed in detail.

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)

This paper describes detection of electro-discharged machine (EDM) defects in magnetic steam generator (SG) tubes of Monju fast breeder reactor (FBR). The EDM defects are located under support plate (SP), on the outer tube surface and they are detected by a remote field eddy current probe. Using the experimental measurements and a multi frequency algorithm, the defect signal can be extracted from the SP signal. The parameters of the multi-frequency algorithm were calculated by comparing SP measurements with two-dimensional finite element simulations (FEM). (author)

Energy Technology Data Exchange (ETDEWEB)

This study deals with the application of wavelet analysis to detection and characterization of defects from eddy-current and ultrasonic testing signals of a low signal-to-noise ratio. Presented in this paper are the methods for processing eddy-current testing signals of heat exchanger tubes of a steam generator in a nuclear power plant. The results of processing eddy-current testing signals of tube test pieces with artificial flaws show that the flaw signals corrupted by noise and/or non-defect signals can be effectively detected and characterized by using the wavelet methods. (author)

International Nuclear Information System (INIS)

Mitsubishi is now developing a next-generation Pressurized Water Reactor (PWR) which has the innovative feature of hybrid safety systems (optimum combination of passive safety system and active safety system) and passive core cooling by horizontal steam generators during Loss of Coolant Accident (LOCA). In order to confirm the capability of this passive core cooling system during LOCAs, the thermal-hydraulic tests of horizontal steam generator and the integral thermal-hydraulic tests simulating the LOCAs were performed. The thermal-hydraulic tests of horizontal steam generator consist of a single tube test and a multi-tubes test. On the basis of these test results, the heat transfer characteristics of steam-water two-phase flow with noncondensable gas along a long horizontal tube is understood and the heat transfer correlation including the ...

Energy Technology Data Exchange (ETDEWEB)

The secondary side corrosion of steam generator tubes is the main degradation of components in operating power plants, strongly impacted by chemistry. This is why EDF has largely studied the chemical parameters in its 56 PWRs which might influence corrosion development. The results of 168 hideout returns of chemical species performed on the French plants allowed to draw conclusions on where chemical species are likely to concentrate in steam generators and on the influence of several contaminants on corrosion processes: sodium, chloride, phosphate, organic compounds, etc... Based on laboratory studies and plants feedback, new chemistry specifications were established and are now applied to EDF units to minimize corrosion and operating costs and to provide a good availability while maintaining an excellent safety. Boric acid is added in the secondary water of the 10 oldest units with Inconel 600 MA ...

Energy Technology Data Exchange (ETDEWEB)

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.

Energy Technology Data Exchange (ETDEWEB)

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)

The following method is proposed as a supplement to main patent 2419107, which deals with the filling of tubular plates surrounding the lead cores: the flat materials which are used to produce the jacket tubes are prepared before forming into jacket tubes by a film producing aqueous carboxy-methyl cellulose solution closing the pores. The preparation can be done by coating the flat material with a soluble foil. After pressing the paste in, the tubular plates are dried and after cooling are dipped into sulphuric acid with a specific gravity of 1.24 g/cm/sup 3/. The carboxy-methyl cellulose is dissolved and the permeability of the tube walls for electrolyte is restored after only 15 minutes.

International Nuclear Information System (INIS)

The detection, in European steam generators, of intergranular stress corrosion cracking initiating from the primary side (PWSCC) and the somewhat similar detection of intergranular stress corrosion (IGSCC) and intergranular corrosion (IGA) initiated from the secondary side in Japanese steam generators has led to a growing awareness of the potential for such corrosion forms in United States PWR steam generators. What had been a minor occurrence at several units is now a cause of concern and repair measures at more than 30 sites in the United States, and in some cases, is the major cause of steam generator unavailability. The United States nuclear utilities and EPRI formed the first Steam Generator Owners' Group (SGOG) in 1977 in response to the prospect of continued denting corrosion. The occasional appearance of PWSCC, IGSCC and IGA in domestic steam generators ...

International Nuclear Information System (INIS)

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 describes work leading to the model generator cleaning demonstrations and ...

Energy Technology Data Exchange (ETDEWEB)

The failure of receivers has been one of the main operating problems at the White Cliffs solar thermal power plant. This Technical Note reports the results of an initial investigation that identifies the cause as having been their thermal fatiguing of the tube walls. The fatigue appears to be caused by unstable heat transfer at vapor qualities below the point where critical heat flux is generally exceeded. Methods for avoiding this problem are tested.

Energy Technology Data Exchange (ETDEWEB)

In this study, deformation modes and precipitations have been characterized in test pieces made of alloy 800, grade 2 hyper-hardened state and age-conditioned for 3000 h at 550/sup 0/C, used for steam generator tubes of the Super Phenix Reactor, after continuous fatigue and fatigue-relaxation tests in the oligocyclic range. This microstructural study has provided an interpretation of the fatigue behaviour of the material.

Energy Technology Data Exchange (ETDEWEB)

Possible threats against the leak tightness of the reactor containments, due to pipe whips from hypothetical pipe ruptures in the steam- and feedwater systems, have been investigated for Forsmark 3/Oskarshamn 3, Ringhals 1, Oskarshamn 1 and Barsebaeck 2/Oskarshamn 2. Based on available drawings, such as installation drawings and isometric views of pipes, the pipe systems have been put together in new drawings with their bracing supports and containment walls. This inventory shows that pipe whips can occur on a number of places on the containments walls after hypothetical pipe ruptures in the steam- and main feedwater systems. In order to find out whether these pipe whips are real threats against the leak tightness, further analysis needs to be made but are out of the scope of this investigation.

International Nuclear Information System (INIS)

Results from a heavily fouled 19 tube C-E model boiler test to investigate the potential for egg crate corrosion in aggressively fouled AVT chemistry are reported. Substantial support plate and egg crate corrosion was produced in this test. Carbon steel drilled support plates exhibited extensive denting which resulted in flow hole ligament cracking. Corrosion of the carbon steel egg crate, through-wall at areas of tube contact, resulted in denting of the Alloy 600 heat transfer tubes. Corrosion performance of the 409 stainless steel egg crate was improved compared to the carbon steel egg crate although localized through-wall corrosion was noted. The results from the above test and previously reported tests were compared based on the following simplifying assumptions: maximum dent size and/or corrosion penetrations utilized, average bulk water chloride concentrations, pilling ...

Energy Technology Data Exchange (ETDEWEB)

Recently the increase of steam temperature and pressure of power plant is required to enhance the thermal efficiency and reduce the CO{sub 2} emission. For the application to advanced USC (Ultra Super Critical) boiler with steam temperature around 700 C, the application of Ni-based alloy such as Alloy617 will be necessary. A new Ni-based alloy USC141 (20Cr-10Mo-2Ti-Al-bal.Ni) with excellent creep rupture strength and low thermal expansion has been developed by Hitachi Ltd. and Hitachi Metals Ltd. as the candidate material for 700 C USC turbine components. In present work, to investigate the possibility for boiler tube application of USC141, its weldability and high temperature strength properties were experimentally examined. The tested material as solution-treated condition shows higher creep rupture strength than that of Alloy617. GTAW (Gas Tungsten Arc Welding) trials of tubular specimen using NIMONIC263 filler wire were ...

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)

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 polarization measurements was verified with the respect of stress corrosion cracking ...

International Nuclear Information System (INIS)

Small scale mock-ups of the primary first wall, the baffle first wall, the shield block and a partial model for the edge of the primary first wall module were designed and fabricated incorporating most of the key design features of the ITER shielding blanket. All mock-ups featured the DSCu heat sink, the built-in SS coolant tubes within the heat sink and the SS shield block. CFC tiles was used as the protection armor for the baffle first wall mock-up. The small scale shield block mock-up, integrated with the first wall, was designed to have a poloidal curvature specified in the ITER design. Fabrication routes of mock-ups were decided based on the single step solid HIP of DSCu/DSCu, DSCu/SS and SS/SS reflecting the results of previous joining techniques development and testing. For attaching the CFC tiles onto DSCu heat sink in the fabrication of the baffle first ...

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Within the scope of the German reactor safety research the thermohydraulic computer code ATHLET which was developed for accident analyses of western nuclear power plants is more and more used for the accident analysis of VVER-plants particularly for VVER-440,V-213. The experiments with the HORUS-facilities and the analyses with the ATHLET-code have been realized at the Technical University Zittau/Goerlitz since 1991. The aim of the investigations was to improve and verify the condensation model particularly the correlations for the calculation of the heat transfer coefficients in the ATHLET-code for pure steam and steam-noncondensing gas mixtures in horizontal tubes. About 130 condensation experiments have been performed at the HORUS-II facility. The experiments have been carried out with pure steam as well as with noncondensing gas injections into the steam mass flow. The ...

International Nuclear Information System (INIS)

General corrosion testing of Alloy 800 tube material by the General Electric Co. using out of pile loops with steam temperatures in the range 430 to 770 deg.C and 70 at. pressure with a surface heat flux of 55 W.cm"-"2 and flow velocities of 30 to 60 m.sec"-"1 showed after about 5000 h (a) a relatively moderate loss of metal at a fairly high corrosion rate, (b) a relatively high loss of oxide to the steam phase, and (c) the formation of a zone at the metal surface depleted in chromium by diffusion to the oxide film. It is important to establish whether these results for Alloy 800 were typical of those to be expected under heat transfer, and whether they could be used as a basis for extrapolation to the higher heat fluxes and longer exposure time relevant to superheat applications. Additional corrosion tests in superheated steam of 1000 h duration under isothermal conditions were therefore undertaken and ...

International Nuclear Information System (INIS)

In response to European operating experience regarding Alloy 800 steam generator (SG) tube cracking, New Brunswick Power Nuclear (NBPN) recognized two important shortcomings in the industry SG technical support framework. First, information exchange networks between utilities with Alloy 800 SG tubing should be strengthened. Second and more importantly, there is a need for consistent industry guidance for Alloy 800 SG tubing degradation management, supported by clear technical bases. It is NBPN's perspective that this guidance can be improved immediately using the extensive amount of laboratory and field data that is available for Alloys 600, 690, and 800, and EPRI methodology developed to predict the performance of Alloy 690. NBPN has initiated the development of a performance-based inspection plan. This effort is based on demonstrated improvement factors for Alloy 800, defined by the EPRI approach, and ...

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 polarization measurements was verified with the respect of stress corrosion cracking ...

British Library Electronic Table of Contents (United Kingdom)

A procedure based on electrophoretic deposition (EPD) was developed to coat metal plates with powder catalysts. The method was tested on stainless-steel plates with three Ni-based catalysts for the steam reforming of ethanol. The catalysts (Ni/La2O3/g-Al2O3) contained 15% Ni and 8% La, and were prepared using three types of g-alumina with different textural properties. The powder catalysts were suspended in isopropanol, and EPD deposition was performed with a voltage of 100V and a distance between electrodes of 2cm. Deposition time was varied between 3 and 7min, which gave a thickness of the catalyst layer from around 30 to 100mm. The morphology of the catalyst layer was dependent on the textural characteristics of the g-Al2O3 used to prepare the catalyst. The activity of the catalyst plat...

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The dielectric wall accelerator (DWA) system being developed at the Lawrence Livermore National Laboratory (LLNL) uses fast switched high voltage transmission lines to generate pulsed electric fields on the inside of a high gradient insulating (HGI) acceleration tube. High electric field gradients are achieved by the use of alternating insulators and conductors and short pulse times. The system is capable of accelerating any charge to mass ratio particle. Applications of high gradient proton and electron versions of this accelerator will be discussed. The status of the developmental new technologies that make the compact system possible will be reviewed. These include, high gradient vacuum insulators, solid dielectric materials, photoconductive switches and compact proton sources.

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A new theory has been developed to calculate the steady state temperature profile in a cylindrical sample positioned along the entire axis of a cylindrical microwave cavity. Temperature profiles were computed for alumina rods of various radii contained in a cavity excited in one of the TM{sub 0n0} modes with n = 1, 2 or 3. Calculations were also performed with a concentric outer cylindrical tube surrounding the rod to investigate hybrid treating. The parametric studies of the total sample center and surface temperatures were performed as a function of the total power transmitted into the cavity. Also, the total hemispherical emissivity was varied at boundaries of the rod, surrounding tube, and cavity walls. The results are discussed in the context of controlling the average rod temperature and the temperature distribution in the rod during microwave processing.

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RODbaffle heat exchanger is a kind of shell-and-tube heat exchanger with eminent performance. Because of the characteristics of tube bundle support structure, fluid flow on shellside is longitudinal and periodical, and fluid flow is of symmetry in landscape orientation. According to the fluid flow characteristics on shellside and ignoring the impact of inlet, outlet and shell wall, a periodic flow unit duct was taken as the simplified model of the shellside to perform numerical simulation by using CFD code, FLUENT. It was found that the both errors in magnitude of the main fluid velocities and heat transfer coefficients between results of simulation and that of experiment or correlations are less than 10%, and the errors of pressure drop between simulation and correlation are no more than 20%. The detailed characteristics and relation of fluid flow and heat transfer on shellside of the RODbaffle heat exchanger were analyzed ...

International Nuclear Information System (INIS)

Experiments have been performed with the PMK integral-type facility, a model of WWER-440 type PWRs, to investigate two-phase natural circulation behaviour. The phenomena to be expected in this reactor type are different from those in PWRs with vertical steam generators mainly due to the loop seal in the hot leg and the horizontal layout of the steam generator heat transfer tubes. The experiments showed that the system is repressurized when the water level drops to the hot leg elevation due to the effect of the loop seal. Opening of the loop seal can be smooth, but may lead to oscillations depending on the power and the mass inventory. Natural circulation recovers after the hot leg loop seal is opened, but then decreases with further mass inventory decrease. (orig.).

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)

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Accurate prediction of tube surface temperature is important for determining whether accelerated superheater corrosion will occur in kraft recovery boilers. A heat transfer model, which combines an iterative numerical solution scheme with a more realistic flue gas flow field obtained from the CFD code FLUENT, has been developed to predict the tube, steam and flue gas temperatures in the superheater region. Results of tests performed for a recovery boiler show that temperatures predicted using the model are in good agreement with actual temperatures measured using thermocouples. The presence of a large flue gas recirculation zone above the bullnose is shown to have an adverse effect on the heat transfer in the superheater region. 9 refs., 12 figs., 1 tab.

International Nuclear Information System (INIS)

The cold-worked Zr-2.5%Nb alloy is used as material for the pressure tubes in CANDU nuclear reactors. During the service life in reactor, diffusion of hydrogen and/or deuterium in the pressure tubes wall occur. Below a certain temperature, a stable hydride of zirconium is formed, as a brittle phase which can lead to catastrophic failures. For this reason, it is very important to be able to investigate the hydrogen effect on the micro structural properties of zirconium alloys. In the present paper a non-destructive testing technique is used, known as ultrasonic spectral analysis. When an ultrasonic signal traverses a medium, the frequency components associated with the input signal are altered. By frequency analysing the reflected signals, it is possible to study and compare the material properties. The two major parameters measured in ultrasonic spectroscopy are the attenuation and the velocity of the waves. Attenuation is ...

International Nuclear Information System (INIS)

The risk analysis was carried out in consideration of conditions prevailing at the Kalkar site analogous to the investigations in phase A of DRS (German Reactor Study). Earthquake design loads include the probabilities of upper deviations of the site intensities to be expected. The calculations of dynamic loads for select buildings are made using models and computational methods. Component analyses were performed analogous to DRS for the supports of large components, supports of the roof construction of the reactor building taking into account support reserves due to plastic work capacity, wall disks in steam generator buildings and switchboard plant buildings. (DG).

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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 led to an ...

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 where feedwater was polluted with sea or river water. Specific effects 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 hrs for seawater 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 seawater pollution. Soaks cannot stop denting if they are not followed by an on-line ...

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During 17 years since the start of operation of the first commercial LWR in Japan, many LWRs have experienced various corrosion damages, but the causes of them were clarified, and the counter-measures were executed effectively in actual plants, as the results, the cause of corrosion damage decreased remarkably, and now, the high rate of operation has become to be maintained. In this paper, the major cases of corrosion damage experienced in LWRs in Japan and foreign countries, the causes of them and the countermeasures, the problems of hereafter and so on are described. The corrosion damage of metallic materials in the environment of LWRs occurs in the parts in contact with high temperature, high pressure water and steam, such as stainless steel piping in the primary cooling system of BWRs, and nickel alloy heating tubes of steam generators, carbon steel feed water piping and zirconium alloy fuel cladding ...

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Nuclear steam generator tubes have become flawed in ways that challenge conventional eddy current probes. In response to the shortcomings of conventional probes, array probes have been developed to improve measurement capabilities. However, the commercially available array probes have exhibited several weaknesses that offset the advantages and limit the applications in steam generator inspections. A primary weakness is the relatively high rate of probe failure coupled with the high unit cost for each probe. This can be costly for a utility in the time lost for probe replacement and increased radiation exposure in addition to the probe costs. Other weaknesses which make array probes undesirable for routine use are: poor mechanical and electrical characteristics; difficulty in operation and calibration; and incomplete coverage of the tube circumference. Several prototype array probes have been built to ...

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Short communication.

Energy Technology Data Exchange (ETDEWEB)

Short communication

Energy Technology Data Exchange (ETDEWEB)

The conceptual design, parametric analysis, cost and performance analysis, and commercial assessment of a 100-MWe line-focus solar central receiver power plant are reported. This volume contains the appendices: (a) methods of determination of molten salt heat-transfer coefficients and tube-wall temperatures, (b) inputs for STEAEC programs, (c) description of system analysis computer program, (d) receiver analysis program, and (e) heliostat production plan and design methodology. (WHK)

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)

This paper describes the results obtained from two large experimental facilities built at Sandia National Laboratories in Albuquerque, NM. FLAME (Flame Acceleration Measurements and Experiments) is a large horizontal rectangular channel designed to study flame acceleration, transition to detonation, simulation of combustion in containment geometries, component heating, and other problems in hydrogen combustion relevant to reactor safety. The Heated Detonation Tube has been designed to study detonations in hydrogen-air-steam mixtures. Both facilities have been in operation for just over a year. 12 figures.

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)

During the service in nuclear power plants, steam generator (SG) tubing will suffer from a vibration with a frequency from 30-40 Hz. This means that SG tubing will undergo a very high cycle fatigue process during the design life time (25 years=about 3x10"1"0cycles). The material degradation and the very high cycle fatigue life of SG tubing materials due to the vibration are therefore of concern. In this investigation, the fatigue properties of Alloy 690 and Alloy 800 SG tubing in the very high cycle fatigue regime (higher than 10"7 cycles) and strain controlled fatigue at temperatures up to 330"oC have been investigated. The fatigue properties of SG tubing of Alloy 690 are comparable to that of wrought material. The strain controlled fatigue properties of Alloy 690 and Alloy 800 are slightly better at 288"oC than at room temperature (RT). At the stress near the ...

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 materials used, including Alloy 800. Reasons ...

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 extension to 18 ...

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 improvements, while flushing and ...

Energy Technology Data Exchange (ETDEWEB)

The two-phase heat-transfer and pressure-drop characteristics of offset-strip-fin heat-exchanger geometries are examined using both experimental and analytical modeling techniques. Newly-obtained boiling-flow heat-transfer data are presented for two large-scale offset fin geometries at low-to-moderate wall-superheat conditions. Data for which nucleate boiling appeared to be completely suppressed were analyzed together with similar results for other offset fin surfaces to study the effects of channel dimensions on the convective evaporation component of the local heat transfer coefficient. Although these data were found to be well correlated in terms of modified forms of the F and Martinelli parameters used in previous studies of round-tube flows, a single F-factor correlation valid for all combinations of fin and channel dimensions did not appear to exist. The two-phase pressure-drop characteristics of offset-fin surfaces were found to be much ...

International Nuclear Information System (INIS)

Lappeenranta University of Technology (LTKK) and VTT Energy carried out a series of preliminary tests in 1999 to study the behavior of noncondensable (NC) gases in VVER geometry. The tests aimed at studying the effect of NC gases on system thermal-hydraulics and on heat transfer in a horizontal steam generator (HSG). The system behavior can be affected by hydrogen produced in the core in case of a severe accident, by nitrogen from hydro-accumulators released into the primary circuit in case of a loss-of-coolant accident (LOCA) and more generally by any NC gas in all cases where cooling is ensured by natural circulation. A secondary objective of the tests - the first series of tests ever performed with NC gas with PACTEL - was to find out, if the instrumentation of PACTEL was adequate for this type of tests and if it was functioning properly. This paper presents the measured and calculated (CATHARE code version V15a mod 2.1) results of the test NCg-l. It was carried ...

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Steam generators (SGs) in PWRs and CANDUs are designed for at least a 30-year operating life. However, in the 15-25 years that SGs tubed with Alloy 600 have operated commercially, they have experienced reduced reliability, mainly due to SG tubing degradation. One of the degradation mechanisms that Alloy 600 SG tubing has suffered from is lead-induced stress corrosion cracking (PbSCC) in AVT and near-neutral SG environments. In contrast to Alloy 600 tubing, test data obtained in high temperature water indicate that Alloy 800 is resistant to cracking by lead and lead compounds. For Alloy 800 (cold worked, shot peened), the most aggressive environment is reported to be alkaline with minor concentration of PbO (80 ppm). Work by Max Helie et al. also concluded that Alloy 800 is not sensitive to lead assisted SCC for pH values close to neutrality, whereas it could be affected in high ...

International Nuclear Information System (INIS)

A new generation of transmit-receive single-pass probes, denoted as C6 or X probe, was field tested during the Gentilly II, 2000 steam generator tube inspection. This probe has a performance equivalent to rotating probes and can be used for tubesheet and full-length inspection at an inspection speed equivalent to that of bobbin probes. Existing C3 transmit-receive probes have been demonstrated to be effective in detecting circumferential cracks. The C5 probe can detect both circumferential and axial cracks and volumetric defects but cannot discriminate between them. The C6 probe expands on the capabilities of both probes in a single probe head. It can simultaneously detect and discriminate between circumferential and axial cracks to satisfy different plugging criteria. It has excellent coverage, good defect detectability, and improved sizing and characterization. Probe data is displayed in C-scan format so that the amount of data to be analyzed ...

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Traditional two-phase flow-regime criteria based on the gas and liquid superficial velocities may not be suitable to the analyses of rapid transient or entrance flows by the two-fluid model. Under these conditions, it is postulated that direct geometrical parameters such as the void fraction are conceptually simpler and therefore more reliable parameters to be used in flow-regime criteria than the traditional parameters. From this point of view, new flow-regime criteria for upward gas-liquid flow in vertical tubes have been developed considering the mechanisms of flow-regime transitions. These new criteria can be compared to existing criteria and experimental data under steady-state and fully developed flow conditions by using relative velocity correlations. The criteria showed reasonable agreements with the existing data for atmospheric air-water flows. Further comparisons with data for steam-water in round tubes and a ...

International Nuclear Information System (INIS)

Liquid sodium is used as coolant in Fast Breeder Reactors (FBR). There is a likelyhood of sodium spillage in ambient air in the Steam Generator Building (SGB) of the FBR plant. Due to high chemical reactivity with oxygen, especially at temperatures greater than 573 K, it catches fire very easily. In order to carryout safety related experimental studies for different modes of sodium fires and to develop suitable mathematical models for the assessment of their consequences, an experimental facility (SFEF, Sodium Fire Experimental Facility) is being setup a IGCAR, Kalpakkam. The SFEF is having a 540 m"3 volume experimental hall. Stainless steel linear will be provided on the inside surfaces of experimental hall walls, ceiling and floor. Analysis has been carried out for enclosed sodium pool fire scenarios in SFEF by using sodium pool fire code SOFIRE II, which estimates the thermal transients like pressure rise, gas temperature rise, cell ...

International Nuclear Information System (INIS)

In this paper, a phenomenological model of the thermal hydraulics of convective boiling in the post-critical-heat-flux (post-CHF) regime is developed and discussed. The model was implemented in the TRAC-PF1/MOD2 computer code (an advanced best-estimate computer program written for the analysis of pressurized water reactor systems). The model was built around the determination of flow regimes downstream of the quench front. The regimes were determined from the flow-regime map suggested by Ishii and his coworkers. Heat transfer in the transition boiling region was formulated as a position-dependent model. The propagation of the CHF point was strongly dependent on the length of the transition boiling region. Wall-to-fluid film boiling heat transfer was considered to consist of two components: first, a wall-to-vapor convective heat-transfer portion and, second, a wall-to-liquid heat transfer representing ...

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This paper presents the results of an experiment conducted in the BN-350 fast reactor in Kazakhstan that involved the irradiation of argon-pressurized thin-walled tubes (0--2000 MPa hoop stress) constructed from Fe-16Cr-15Ni-3Mo-Nb stabilized steel in contact with the sodium coolant, which enters the reactor at {approximately}270 C. Tubes in the annealed condition reached 75 dpa at 335 C, and another set in the 20% cold-worked condition reached 81 dpa at 360 C. Upon disassembly all tubes, except those in the stress-free condition, were found to have failed in an extremely brittle fashion. The stress-free tubes exhibited diameter changes that imply swelling levels ranging from 9 to 16%. It is expected that stress-enhancement of swelling induced even larger swelling levels in the stressed tubes. The embrittlement is explained in terms of the sensitivity of the ...

International Nuclear Information System (INIS)

The Supercritical Water Cooled Reactor(SCWR) operates in a pressure around 25MPa and temperature of 293#approx#510 .deg. C. In order to study the heat transfer behaviors and good comparisons between the various fluids, a heat transfer test loop(SPHINX) using CO_2 has been constructed in KAERI as a part of international research program, I-NERI. At a supercritical pressure, the heat transfer coefficient is much larger than that estimated from the Dittus-Boelter correlation for a relatively large flow rate with moderate wall heat flux conditions. This phenomenon was explained by the rapid variations of the physical properties near the wall with the temperature. On the contrary, the heat transfer becomes worse when the bulk fluid enthalpy is below the pseudo-critical enthalpy under a low flow rate with large heat flux conditions. This phenomenon is called 'deteriorated heat transfer', and which is explained as the modification of the shear stress ...

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The titanium stabilized austenitic steel similar to the type of 321 is sensitive to the stress corrosion crackingunder horizontal steam generator operating condition. SCC was observed under crevice corrosion parameters and has resulted in the transgranular or intergranular cracking at the both, components primary collectors and heat exchange tubes. The crevice environment is characterized by aggressive impurities and 'non aggressive' compounds. Sulfates and chlorides as aggressive species and silicates and alumino-silicates as 'non aggressive' species on the other hand are present in significant amount in the crevice environment under operating condition. Local water chemistry parameters were evaluated with MULTEQ Code. As input data the measured operational values of local and bulk environments have been used. The determined parameters were compared with the results of thread hole environment analyses and ...

International Nuclear Information System (INIS)

Computer programs have been written to allow the analysis of different types of eddy-current probes and their performance under different steam generators 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 intergranular attack. 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)

For the advanced fusion reactor FFHR2 (Force Free Helical Reactor) that has been proposed by NIFS, molten salt Flibe (LiF:BeF2=64:36) breeder blanket system is selected because of Flibe's features such as chemical stability, low-pressure operation and low electric conductivity. The Flibe is however high Prandtl number fluid since it has high viscosity and low thermal conductivity. Therefore its heat transfer performance is low compared with liquid Li or Pb-Li. In addition to heat removal of 1MW/m2 on the first wall, electrolysis of molten salt due to MHD effect will take place under high flow rate condition. This indicates that heat transfer enhancement under low flow rate is essential for the Flibe blanket system. In our laboratory, heat transfer characteristics of molten salt HTS (KNO3:NaNO2:NaNO3=53:40:7), have been evaluated, which is used as a simulant fluid of Flibe from the points of view of Be's toxicity and similar Prandtl number. In this paper, we adopt ...

International Nuclear Information System (INIS)

Alloy 800 has been used for steam generator (SG) tubing for more than 30 years, primarily in CANDU reactors and in reactors in Germany. The grade of Alloy 800 tubing used for this service has a controlled Ti/C ratio (#>=#12 for CANDU SGs), and this specification is sometimes termed Alloy 800 M or Alloy 800 NG . There have been very few corrosion-related flaws detected in this material in SG service, and, until recently, no incidences of cracking. There has been extensive R and D carried out on Alloy 800 tubing, both in Canada and elsewhere, under a variety of operating conditions, including shutdowns, which show that it has excellent resistance to corrosion-related degradation under specified and appropriate operating conditions. These R and D findings are reflected in the in-service experience. It has been shown from the R and D that Alloy 800 is susceptible to corrosion under acidic conditions, in ...

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Multi-walled carbon nanotubes (MWNTs) were modified with poly(hexamethylene adipamide) (also known as Nylon 66) via a controlled polymer solution crystallization method. A 'nanohybrid shish kebab' (NHSK) structure was found wherein the MWNT resembled the shish while Nylon 66 lamellar crystals formed the kebabs. These Nylon 66-functionalized MWNTs were used as precursors to prepare polymer/MWNT nanocomposites. Excellent dispersion was revealed by optical and electron microscopies. Nitric acid etching of the nanocomposites showed that MWNT formed a robust network in Nylon 66. Non-isothermal DSC results showed multiple melting peaks, which can be attributed to lamellar thickness changes upon heating. The crystallite sizes L{sub 100} and L{sup 010} of Nylon 66, determined by WAXD, decreased with increasing MWNT contents. Isothermal DSC results showed that crystallization kinetics increased first and then decreased with increasing MWNT contents in ...

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A safety analysis code to design a pressurized water reactor and to obtain the licences including entire proprietary rights is under development in domestic research and development project. The purpose and scope of this report is to develop the flow regimes related models for inter-phase friction, wall frictions, wall heat transfer, and inter-phase heat and mass transfer in two-phase three-field equations. In order to choose choose the flow regime criteria, we have investigated various exiting best-estimate T/H codes in this chapter 2. They are the RELAP5-3D, TRAC-M, CATHARE, MARS codes. Around 500 references used in these codes have been collected and reviewed. Also we have investigated eleven papers in detail. In chapter 3, based on the selected flow regimes, the flow regime maps for a gas-liquid flow in horizontal and vertical tubes have decided including the mechanisms of flow regime transition regions. Conclusively, ...

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A novel compact CT-guided intensity modulated proton radiotherapy (IMPT) system is described. The system is being designed to deliver fast IMPT so that larger target volumes and motion management can be accomplished. The system will be ideal for large and complex target volumes in young patients. The basis of the design is the dielectric wall accelerator (DWA) system being developed at the Lawrence Livermore National Laboratory (LLNL). The DWA uses fast switched high voltage transmission lines to generate pulsed electric fields on the inside of a high gradient insulating (HGI) acceleration tube. High electric field gradients are achieved by the use of alternating insulators and conductors and short pulse times. The system will produce individual pulses that can be varied in intensity, energy and spot width. The IMPT planning system will optimize delivery characteristics. The system will be capable of being sited in a conventional linac vault ...

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A joint US/Russia program is being conducted to develop, fabricate, launch, and operate a solar dynamic demonstration system on Space Station Mir. The goal of the program is to demonstrate and confirm that solar dynamic power systems are viable for future space applications such as the International Space Station Alpha. The major components of the system include a heat receiver, a closed Brayton cycle power conversion unit, a power conditioning and control unit, a concentrator, a radiator, a thermal control system, and a Space Shuttle carrier. This paper discusses the design of the heat receiver component. The receiver comprises a cylindrical cavity, the walls of which are lined with a series of tubes running the length of the cavity. The engine working fluid, a mixture of xenon and helium, is heated by the concentrated sunlight incident on these tubes. The receiver incorporates integral thermal storage, using a eutectic ...

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An approximate analyzing model was developed to predict the azimuth of zigzag streamline, the distribution ratio between two neighboring layers and the radial distribution among the layers of the gas passing through the shaft of blast furnace loaded in layers. The theoretical basis of the approximate analysis is composed of the material balance derived from the assumption that any stream line surrounded with two streamlines is isolated and no material transfer occurs through the tube wall, and of the energy theory that the stream tube and passage are formed so that the total pressure drop in the whole system is the minimum. The effects of the apparent angle of repose between two layers, the packing volume and the passage resistance on the non-uniform flow in the layer were evaluated on the basis of the model. The result calculated by the approximate analysis agrees with the result by the numerical analysis based on the ...

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Condensing two-phase channel flow occurs in many industrial applications, including heating and refrigeration systems. It can also occur in certain nuclear reactor accidents. For example, during a small-break loss-of-coolant accident in a pressurized water reactor, following the partial depletion of the primary coolant, condensation of steam on the primary side of the steam generator tubes can provide a heat sink for disposal of the decay heat generated in the reactor core. Condensing two-phase flow can also play an important role in the operation of the passive emergency cooling system in the advanced simplified boiling water reactor. Here, steady-state condensation in the presence of a noncondensable in a concurrent two-phase channel flow is analyzed using a two-fluid model. The effect of noncondensables on the combined heat transfer at the liquid-gas mixture interphase is accounted for by using the stagnant film model, ...

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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 evaluated by means of ...

International Nuclear Information System (INIS)

PWR plants have made efforts to maintain the long-term integrity of the steam generators (SG) by reducing the amount of corrosion products entering the secondary side of the SG. Iron entered the SG can cause several problems: degraded heat conductivity of the SG tubes in locations where iron is deposited, water level oscillations in the SG due to tube support plate hole blockage, and initiation and propagation of inter-granular attacks (IGA) and stress corrosion cracking (SCC). One of the most effective measures, high all-volatile treatment (AVT) chemistry has been applied to actual plants to reduce the flow-accelerated corrosion (FAC) coming from the carbon steel piping. The secondary water chemistry at Genkai NPS 1 and 2 changed, from the Low AVT chemistry to the High AVT chemistry, in November 2006. In this paper, we will describe the results of experiments in applying the use of High pH water in the secondary water ...

International Nuclear Information System (INIS)

Fretting wear test in room temperature water was performed to evaluate the wear coefficient of CANDU (CANadian Deuterium Uranium) steam generator (SG) tube material (Alloy 800) against 410 type martensitic stainless steels. The main focus is to compare the wear behaviors between Alloy 800 and Inconel 690. Test conditions are 10#approx#30N of normal load, 200#approx#450mm of sliding amplitude and 30Hz of frequency. The result indicated that the wear rate of Alloy 800 was higher than those of Inconel 690 at various test condition such as normal loads, sliding amplitudes etc. From the results of SEM observation, there was little evidence of plastic deformation layer that were dominantly formed on the worn surfaces of Inconel 690. Also, wear particles in Alloy 800 were released from contacting asperities deformed by severe plastic flow during fretting wear. Main cause of wear rate between Alloy 800 and Inconel 690 may be due to the difference of ...

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Direct Flame Impingement involves the use of an array of very high-velocity flame jets impinging on a work piece to rapidly heat the work piece. The predominant mode of heat transfer is convection. Because of the locally high rate of heat transfer at the surface of the work piece, the refractory walls and exhaust gases of a DFI furnace are significantly cooler than in conventional radiant heating furnaces, resulting in high thermal efficiency and low NOx emissions. A DFI furnace is composed of a successive arrangement of heating modules through or by which the work piece is conveyed, and can be configured for square, round, flat, and curved metal shapes (e.g., billets, tubes, flat bars, and coiled bars) in single- or multi-stranded applications.

British Library Electronic Table of Contents (United Kingdom)

An evaporative cooling system developed for operation and qualification testing of silicon pixel and microstrip detectors for the inner tracking detector of the CERN ATLAS spectrometer is described. Silicon detector substrates must be continuously operated between 0 and ???7?C in the high radiation environment near the circulating beams at the CERN Large Hadron Collider (LHC). This requirement imposes unusual constraints on the cooling system and has led to the choice of perfluoro-n-propane (C3F8) refrigerant, which combines good chemical stability under ionizing radiation with high dielectric strength and nonflammability. Since the silicon detectors must also be of extremely light construction to minimize undesirable physics background, coolant tubes are of thin (200 ?m) aluminum wall, wh...

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Steam turbines of the new power plant generation enable the realisation of over-critical steam parameters with 600 C and 300 bar by means of simple and double resuperheating. Extensive research work in the field of material technology has provided the pre-conditions. Forged pieces, castings, high-temperature ferritic 9 - 10 % CrMoV steels, which are additionally alloyed with niobium, nitrogen and wolfram, are used in steam turbines. A proof of strength for shaft and cast steels is available by means of a long-term stress-rupture test of up to max. 60,000 hours. The strength of turbine steels exceeds the one in tube and header steels (x 10 CrMoVNbN 9 1). (orig./MM) [Deutsch] Dampfturbinen der neuen Kraftwerksgeneration erlauben heute die Realisierung von ueberkritischen Dampfparametern mit 600 C und 300 bar mit einfacher und zweifacher Zwischenueberhitzung. Umfangreiche werkstofftechnische ...

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The steam generator inspection and maintenance device has an articulated arm formed at least by two rigid segments connected by a pivot and two plates on the ends of the arms by articulated joints which enable them to rotate about two orthogonal axis. One of the axis is parralel to the axis of the pivot between the arm sections while the other is perpendicular to the surface of the corresponding plate. Each of the plates carries coupling elements which are able to engage with the perforations of the heat exchanger tube plates and also a support for a maintenance tool.

British Library Electronic Table of Contents (United Kingdom)

The present paper deals with instabilities of long homogeneous and isotropic thin shells, characterized by geometrical non-linearities and imperfections, with reference to a bent helicoidally geometrical shape of particular interest for the helicoidal steam generators tube bundle. Apparently no data exist in the literature to describe the non-linear buckling behaviour of curved thin shells under external pressure, thus, the theoretical analyses based on the classical linear elastic theory, as expected, might be inadequate to evaluate the collapse load especially if the curvature is rather large. To the purpose of determining the buckling pressure load the effects of a pre-existing level of geometrical and technological imperfection, unavoidably caused by various manufacturing processes wer...

International Nuclear Information System (INIS)

Alloy 800 has been used by Siemens since 1968 and is now used by KWU for U-tubed steam generators in PWR's. The particular grade of alloy 800 that is used is within the ASTM-Specification B 163, but there have been modifications in composition to improve the corrosion resistance. First the permitted upper limit of carbon was reduced to 0.04% and was then further reduced to 0.03% and a stabilisation ratio of Ti : C >= 12 and Ti : C + N >= 8 was specified. The minimum permitted chromium and nickel levels were increased to 20% and 32% respectively. The maximum permitted levels of other elements or impurities were reduced. At the final fabrication stage peening with glass beads on the outer surface was specified to introduce a compressive stress to depth greater than that allowed for production flaws. An account is given of the behaviour of the alloy in service in four power plants already in operation, and future developments are discussed. ...

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The bobbin probe technique is basically one of the important ECT methods for the steam generator tube integrity assessment that is practiced during each plant outage. The bobbin probe also is the essential component which consists of the whole ECT examination system, and provides a decisive data for the evaluation of tube integrity in compliance with acceptance criteria described in specific procedures. The selection of probe is especially important because the quality of acquired ECT data is determined by the probe design characteristics, such as geometry and operation frequency, and has an important effect on examination results. The Electric Power Research Institute (EPRI) has recently defined the procedures for the qualification of eddy current hardware and technique. These procedures provide two basic methods for qualification. Flawed tube removed from operation, or artificial flaw is required for ...

International Nuclear Information System (INIS)

Several types of corrosion have damaged alloy 600 tubing in the secondary side of steam generators. The types of corrosion include wastage, denting, intergranular attack, stress corrosion, erosion-corrosion, etc. The environments which cause attack may originate from leaks of cooling water into the condensate, etc. When the contaminated feedwater is pumped into the generator, the impurities may concentrate first 200 to 400 fold in the bulk water, depending on the blowdown, and then further to saturation and dryness in heated tube support plate crevices. Characterization of local solution chemistries is the first step to predict and correct the type of corrosion that can occur. The pH is of particular importance because it is a major factor governing the rate of corrosion reactions. The pH of a solution at high temperature is not the same as the ambient temperature, since ionic dissociation constants, solubility and ...

International Nuclear Information System (INIS)

Alloy 690 and Alloy600 are used as a material for the steam generator tubing in the pressurized water reactor(PWR) of nuclear power plants due to its high corrosion resistance. Although those are a highly corrosion resistance material, their stress corrosion cracking(SCC) have been found on occasion, which are deeply related to a surface oxide film on a base material which have occurred on the primary side as well as the secondary side of a tubing. And The SCC is accelerated in the existing Pb which is the impurity of secondary steam generator components. The Oxide on a steel surface in an aqueous solution above 100 .deg. C is composed of a duplex film structure. The inner layer of the oxide is dense and less porous, which is formed by a growth of the oxide layer on the metal surface. The outer layer of the oxide is less adhesive, which is formed by a dissolution and precipitation mechanism. Growth ...

International Nuclear Information System (INIS)

Source term release characteristics are analyzed for the severe SGTR (Steam Generator Tube Rupture) sequences beyond the design basis accidents in Wolsong 2/3/4 plants which are of CANDU6 type reactor. In PWRs, SGTR sequences have long been recognized to be important and are distinctly different from the non-bypass sequences since there is a direct fission product release path from the primary system to the environment bypassing the containment gas volume. Meanwhile, a SGTR in a CANDU reactor is analyzed not to provide a complete and direct path into the environment for the source term resulting from a severe accident. This is because the majority of the fission product released arises from heatup and interactions of the disassembled fuel channel segments and debris in the calandria tank rather than from fuel heatup in the fuel channel. These fission products are released from the calandria tank into the containment atmosphere through the four ...

International Nuclear Information System (INIS)

The Steam Generator (SG) serves as an interface between primary and secondary cycle in Pressurized Heavy Water Reactor (PHWR). Failure of steam generator tubes result in leaking of active heavy water in the secondary closed loop. In Tarapur Atomic Power Station-3 and 4 (TAPS- 3 and 4), Scintillator detectors are provided to detect on line heavy water leakages in SG and moderator heat exchangers by monitoring Nitrogen-16 ("1"6N) and Oxygen-19 ("1"90) activities. Efficacy of detection of these activities at designed detector position on SG blow down line in presence of background radiation field is analysed theoretically. The count rate of "1"9O and "1"6N estimated at the detector position inside Reactor Building (RB) shows that detectors only respond to very high leak rates due to presence of high background even though sensitivity is appreciably good. For detector position out side the RB containment, the travel time for ...

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The Electrohydrodynamic (EHD) is an active heat transfer augmentation technique which utilizes the effect of secondary motions generated through the application of an electrostatic potential to a dielectric fluid. Net result is better momentum and heat transfer between the fluid and the heat transfer wall through destabilization of the thermal boundary layer and better mixing of the fluid adjacent to the heat transfer surface. EHD enhancement of refrigerant/refrigerant oil mixtures heat transfer using the Electrohydrodynamic (EHD) technique is the subject of a three-year experimental investigation in a project funded by the US Department of Energy, effective June 1, 1993. For the interim period between November 1992 and June 1993 when the DOE funds became available, the Air-Conditioning and Refrigeration Technology Institute (ARTI) provided partial funding for our EHD research program with the aim of accomplishing three major tasks: (1) conduct a comprehensive ...

International Nuclear Information System (INIS)

To overcome the limitations of conventional bobbin style eddy-current probes, it has become common practice to perform some portion of a steam generator tube inspection program with array coil probes. Typically, eight surface-riding pancake coils are utilized in a probe head, either as eight independent coils (8 x 1) or in a parallel combination of four serially connected coils (4 x 4). The advantages of the array coils are improved sensitivity to intergranular attack (IGA), reduced interference from tube denting, and improved sensitivity to circumferential cracking. The current status of commercially available probes is such that several weaknesses in array coils have limited the widespread application of this technology. Task S 301-6 was conceived to address each of the weaknesses in order to produce a more reliable and cost-effective probe. The emphasis was on reliability, since the cost associated with downtime for a ...

International Nuclear Information System (INIS)

This paper describes the construction of PWR horizontal steam generator in Soviet Union, the water chemistry treatment for secondary side, the design of steam separator, the test of heat transfer characteristics and operation. (author).

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We show that the eigenvalue equations describing a cylindrical ideal magnetophydrodynamicsw (MHD) plasma interacting with a thin resistive wall can be put into the standard mathematical form: ??? = ??? ?. This is accomplished by using a finite element basis for the plasma, and by adding an extra degree of freedom corresponding to the electrical current in the thin wall. The standard form allows the use of linear eigenvalue solvers, without additional interations, to compute the complete spectrum of plasma modes in the presence of a surrounding restrictive wall at arbitrary separation. We show that our method recovers standard results in the limits of (1) an infinitely resistive wall (no wall), and (2) a zero resistance wall (ideal wall).

International Nuclear Information System (INIS)

Liquid metal cooling for the first wall and blanket of a magnetic confinement fusion reactor has various advantages. However, it has the disadvantages of large magnetohydrodynamic pressure drops and heat transfer deterioration under a strong magnetic field. Thus, the present authors have proposed cooling with a helium-lithium annular mist flow as well as the cooling with a liquid metal boiling flow, and as fundamental studies, investigated the effect of a magnetic field on the flow characteristics and heat transfer of liquid metal two-phase systems since the 1970s. In the present paper we summarize the important findings obtained from our experimental studies for (i) an air-mercury stratified flow in a horizontal rectangular channel, (ii) a helium-lithium annular mist flow in a horizontal rectangular channel, (iii) the mercury pool boiling on a horizontal surface, and (iv) air-mercury upward flows in a vertical circular tube. Based on the ...