Thermo-Hydraulic Analysis in a Assembly Representative of a Feedwater Nozzle

2003 ◽  
Author(s):  
Xavier Boissiere ◽  
Christian Laine ◽  
Louis Doubliez
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Thermal Loading ◽  
Thermal Stresses ◽  
Experimental Testing ◽  
Flow Behavior ◽  
Life Time ◽  
Technical Problems ◽  
Power Stations ◽  
The Impact

One of the ways to reduce costs in modern nuclear power stations is to increase the life time of the steam generator. Much research is being carried out to gain a better understanding of the mechanical and thermal loads, which are generally overestimated in the interest of safety. One of the main technical problems is the feedwater of the reactor at a relatively cold (40 °C) temperature into the hot steam generator (270 °C). The connection zone, i.e. where the cold pipe is connected to the steam generator, is protected by an annulus inside the feedwater nozzle to reduce the thermal stresses. We have to identify the thermo-hydraulic behavior in that zone in order to accurately assess thermal information. This will give reliable boundary conditions for thermo-mechanical calculations. The configuration of the flows is a superposition of well known elementary fluid mechanic problems which interact strongly. No databases are available for this typical configuration, so we need to qualify the flow before using CFD modelisations. Therefore, a specific experimental testing bench was developed. In this paper, we focus our attention on the understanding of the flow in the one eyed cavity. PIV measurements allows us to identify the flow behavior. Our hypothesizes are demonstrated by a frequencies analysis, and finally confirmed with a numerical model. We also present and discuss the impact on a low thermal loading.

Water Flow Analysis in a One Eyed Annulus Cavity Subjected to a Perpendicular Flow

2005 ◽  
Author(s):  
Xavier Boissiere ◽  
Christian Laine
Keyword(s):  
Steam Generator ◽  
Thermal Stresses ◽  
Cold Water ◽  
Experimental Testing ◽  
Flow Analysis ◽  
Cfd Modeling ◽  
Configuration Model ◽  
Feed Water ◽  
Technical Problems ◽  
Hydraulic Behaviour

For safety and economic reasons, one of the greatest concerns in modern nuclear power stations is to increase the service life of the steam generator. Enormous work has been undertaken to obtain better knowledge of the mechanical and thermal loads, which are generally overestimated to ensure maximum safety. One of the technical problems is that the reactor is fed with relatively cold (40 °C) water into the hot steam generator. The connection zone, i.e. where the cold water pipe is connected to the steam generator, is protected by an annular cavity inside the feed water nozzle to reduce thermal stresses. We have to identify the thermo-hydraulic behaviour in that zone in order to get more accurate thermal information for the boundary condition in the thermo-mechanical calculations for this connection. No database is available for this configuration model, so we need to qualify the flow before using CFD modeling. Therefore a specific experimental testing bench along with numerical modeling has been developed. The first requirement is that the flow in the annular cavity is representative of the actual one. For this purpose, we used Reynolds and Strouhal similitudes to set up the flow in the model. Attention has been focused on the understanding of the flow in the annular cavity. PIV measurements enable us to have a good understanding of the flow structure. We also present and discuss the results in terms of velocity profiles. The main result is that the flow is unsteady in the cavity and depends on the length and thickness of the annular cavity. Experimental results give us data to validate our numerical approach, which is used to test several configurations.

Locating Impact on Thick Plates by Acoustic Wave Analysis

2008 ◽  
Author(s):  
Ho-Wuk Kim ◽  
Sang-Kwon Lee
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Group Velocity ◽  
Steam Generator ◽  
Nuclear Power ◽  
Thick Plate ◽  
Impact Load ◽  
Plate Theory ◽  
Thick Plates ◽  
The Impact

Loose parts in a steam generator of a nuclear power plant often impact the wall of the generator and become one of the damage sources in the nuclear power plant. In general, the steam generator of the nuclear power plant is structured by thick plates. This paper presents a novel approach to locating an impact load in a thick plate. The approach is based on an analysis of the acoustic waveforms measured by a sensor array located on the plate surface and theoretically obtained by either the exact elastodynamic or theory the approximate shear deformation plate theory (SDPT). For accurate estimation of the location of the impact source due to loose part, the time differences in the arrival times of the waves at the sensors and their propagation velocities are determined. This is accomplished through the use of a combined higher order time frequency (CHOTF) method, which is capable of detecting signals with lower signal to noise ratio compared to other available methods. The dispersion curves for multi modes of Lamb waves are calculated by using exact plate theory and SDPT. It is difficult to measure directly the group velocity for Lamb mode of acoustic waveform in the thick plate because they are dispersive waves. However, most of the energy in the wave is carried by the flexural waves (A0 mode); the group velocity of this mode is extracted by using the CHOTF technique for estimating the impact source location. The estimates are shown to be in excellent agreement with the actual locations and the technique is applied to the detection of the location of the impact load due to the loose part in a nuclear power plant.

Wear Scar Progression of Impact-Fretting at Elevated Temperature for Steam Generator Tubes in Nuclear Power Plants

2006 ◽  
Vol 326-328 ◽  
pp. 1251-1254 ◽  
Author(s):  
Chi Yong Park ◽  
Jeong Keun Lee
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Operating Conditions ◽  
Wear Scar ◽  
Fretting Wear ◽  
Test Material ◽  
The Impact ◽  
Steam Generator Tubes

Fretting wear generated by flow induced vibration is one of the important degradation mechanisms of steam generator tubes in the nuclear power plants. Understanding of tube wear characteristics is very important to keep the integrity of the steam generator tubes to secure the safety of the nuclear power plants. Experimental examination has been performed for the purpose of investigating the impact fretting. Test material is alloy 690 tube and 409 stainless steel tube supports. From the results of experiments, wear scar progression is investigated in the case of impact-fretting wear test of steam generator tubes under plant operating conditions such as pressure of 15MPa, high temperature of 290C and low dissolved oxygen. Hammer imprint that is actual damaged wear pattern, has been observed on the worn surface. From investigation of wear scar pattern, wear mechanism was initially the delamination wear due to cracking the hard oxide film and finally transferred to the stable impact-fretting pattern.

Crack Initiation Factor by Impact Fretting Wear at INCONEL Alloy for Steam Generator Tube in Nuclear Power Plants

2007 ◽  
Vol 26-28 ◽  
pp. 1269-1272
Author(s):  
Chi Yong Park ◽  
Jeong Kun Kim ◽  
Tae Ryong Kim ◽  
Sun Young Cho ◽  
Hyun Ik Jeon
Keyword(s):  
Steam Generator ◽  
Wear Mechanism ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Operating Conditions ◽  
Fretting Wear ◽  
Steam Generator Tube ◽  
Inconel Alloy ◽  
The Impact

Inconel alloy such as alloy 600 and alloy 690 is widely used as the steam generator tube materials in the nuclear power plants. The impact fretting wear tests were performed to investigate wear mechanism between tube alloy and 409 stainless steel tube support plates in the simulated steam generator operating conditions, pressure of 15MPa, high temperature water of 290°C and low dissolved oxygen(<10 ppb). From investigation of wear test specimens by the SEM and EDS analysis, hammer imprint, which is known to be an actual damaged wear pattern, has been observed on the worn surface, and fretting wear mechanism was investigated. Wear progression of impact-fretting wear also has been examined. It was observed that titanium rich phase contributes to the formation of voids and cracks in sub-layer of fretting wear damage by impact fretting wear.

The Sixth Royal Society Technology Lecture - Nuclear energy for the steel industry

1974 ◽  
Vol 340 (1621) ◽  
pp. 129-146
Keyword(s):  
Steel Industry ◽  
Nuclear Power ◽  
Iron Ore ◽  
Fossil Fuels ◽  
Raw Materials ◽  
Direct Reduction ◽  
Technical Problems ◽  
Power Stations ◽  
Nuclear Heat ◽  
Cold Metal

The steel industry, as a major consumer of coking coal and hydrocarbons, is exploring ways to reduce its dependence on these potentially expensive raw materials by making direct use of nuclear heat. Of the present two routes for producing steel, the major one (the hot metal route) employing the blast furnace which reduces iron ore to yield molten iron which is subsequently refined by basic oxygen steelmaking, does not lend itself to the application of nuclear heat; in the second (the cold metal route) recycled steel-or a substitute-is melted in an arc furnace where already today a proportion of the electricity used is generated in nuclear power stations. The development of ‘direct reduction’ processes allows iron ore to be converted to a solid pre-reduced iron product. In the conventional prereduction process, fossil fuels are used as both fuel and as chemical reductant. With nuclear heat, the fossil fuel-re-formed to a suitable reductant-is confined to the chemical role and not used as a source of heat. This reduction stage would be followed by arc melting, as in the present cold metal route. This basic process, which at present constitutes the minor route, could become the major one for the manufacture of steel in the long term. The lecture will discuss the various processes and outline a possible configuration for an eventual nuclear steelworks, together with some of the technical problems involved.

Effects of Prior Stress Relaxation on the Prediction of Creep Life Using Time and Strain Based Methods

2013 ◽  
Vol 135 (4) ◽  
Author(s):  
Warwick M. Payten ◽  
Ken U. Snowden ◽  
David W. Dean
Keyword(s):  
Stress Relaxation ◽  
Nuclear Power ◽  
Inelastic Strain ◽  
Life Time ◽  
Fatigue Loading ◽  
Creep Life ◽  
Strain Behavior ◽  
Power Stations ◽  
Creep Fatigue ◽  
Inelastic Design

A critical requirement for both next generation conventional and nuclear plants is the development of simplified inelastic design and fitness for purposes procedures that give a reasonably accurate prediction of the complex multiaxial time dependent stress strain behavior. The accumulation of this inelastic strain in the form of coupled creep-fatigue damage over time is one of the principal damage mechanisms which will eventually lead to crack initiation in critical, high temperature equipment. Two main procedures that address creep-fatigue loading are generally used, either a time fraction or a ductility exhaustion approach. It is generally accepted that these methods enable conservative predictions within a factor of 2 to 3 and hence are reliable methods for code based design and fitness for purpose type assessments. However, for complex cycles, this may not be the case; for example, prior relaxation cycles are found to accelerate the creep rupture of the material with the result that a significant reduction in creep life can be observed. An investigation was undertaken into the influence of prior relaxation on resultant failure using a typical low alloy ferritic power station steel. Both time-based and strain based methods were used to predict the damage caused by the stress relaxation cycles followed by operation at steady state. The predictions found that while ductility exhaustion methodologies based on mean properties were adequate in predicting the failure life, time fraction methods were found to be extremely nonconservative for mean properties and only lower bound solutions provided an estimate of remaining creep life. The ASME time fraction approach, using isochronous curves was found to be extremely conservative for K = 0.67, but was able to predict similar damages to ductility exhaustion when K = 1 was used. The Monkman-Grant approach resulted in predictions that erred on the conservative side. The results have implications for both current and future conventional and nuclear power stations as it may be difficult for time based approaches to account accurately for complex cycling, shakedown conditions or stress relaxation at welds.

French PWR Nuclear Power Plants Steam Generators and Reactor Pressure Vessel Heads Aging Management Criteria and Replacement Strategy

2004 ◽  
Author(s):  
Georges Bezdikian
Keyword(s):  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Life Time ◽  
Reactor Pressure Vessel ◽  
Steam Generators ◽  
Life Management ◽  
Power Stations ◽  
Reactor Pressure

The approach used by the French utility, concerning the Aging Management system of the Steam Generators (SG) and Reactor Pressure Vessel Heads, applied on 58 PWR NPPs, involves the verification of the integrity of the component and the Life Management of each plant to guarantee in the first step the design life management and in the second step to prepare long term life time in operation, taking into account the degradation of Alloy 600 material and the replacement of these materials by components made with Alloy 690. The financial stakes associated with maintaining the lifetime of nuclear power stations are very high; thus, if their lifetime is shortened by about ten years, dismantling and renewal would be brought forward which would increase their costs by several tens of billions of Euros. The main objectives are: • to maintain current operating performances (safety, availability, costs, security, environment) in the long term, and possibly improve on some aspects; • wherever possible, to operate the units throughout their design lifetime, 40 years, and even more if possible. This paper shows the program to follow the aging evaluation with application of specific criteria for SG and for Vessel Heads, and the replacement of the Steam Generators and Vessel Heads at the best period. The strategy of Steam Generators Replacement are developed and Vessel Head program of monitoring and replacement are detailed.

Effects of Prior Stress Relaxation on the Prediction of Creep Life Using Time and Strain Based Methods

2010 ◽  
Author(s):  
Warwick M. Payten ◽  
Ken U. Snowden ◽  
David W. Dean ◽  
Samuel R. Humphries ◽  
Lyndon Edwards
Keyword(s):  
Stress Relaxation ◽  
Nuclear Power ◽  
Inelastic Strain ◽  
Life Time ◽  
Fatigue Loading ◽  
Creep Life ◽  
Strain Behavior ◽  
Power Stations ◽  
Creep Fatigue ◽  
Inelastic Design

A critical requirement for both next generation conventional and nuclear plants is the development of simplified inelastic design and fitness for purposes procedures that give a reasonably accurate prediction of the complex multi-axial time dependent stress strain behavior. The accumulation of this inelastic strain in the form of coupled creep-fatigue damage over time is one of the principal damage mechanisms which will eventually lead to crack initiation in critical high temperature equipment. Two main procedures that address creep-fatigue loading are generally used, either a time fraction or a ductility exhaustion approach. It is generally accepted that these methods enable conservative predictions within a factor of 2 to 3 and hence are reliable methods for code based design and fitness for purpose type assessments. However, for complex cycles, this may not be the case, for example prior relaxation cycles are found to accelerate the creep rupture of the material with the result that a significant reduction in creep life can be observed. An investigation was undertaken into the influence of prior relaxation on resultant failure using a typical low alloy ferritic power station steel. Both time based and strain based methods were used to predict the damage caused by the stress relaxation cycles followed by operation at steady state. The predictions found that while ductility exhaustion methodologies based on mean properties where adequate in predicting the failure life, time fraction methods were found to be extremely non-conservative for mean properties and only lower bound solutions provided an estimate of remaining creep life. The Monkman-Grant approach resulted in predictions that erred on the conservative side. The results have implications for both current and future conventional and nuclear power stations as it may be difficult for time based approaches to accurately account for complex cycling, shakedown conditions or stress relaxation at welds.

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