scholarly journals Experimental approach to investigate the dynamics of mixing coolant flow in complex geometry using PIV and PLIF techniques

2015 ◽  
Vol 19 (3) ◽  
pp. 989-1004 ◽  
Author(s):  
Ezddin Hutli ◽  
Valer Gottlasz ◽  
Dániel Tar ◽  
Gyorgy Ezsol ◽  
Gabor Baranyai
Keyword(s):  
Heat Transfer ◽  
Structural Integrity ◽  
Complex Geometry ◽  
Measurement Techniques ◽  
Coolant Flow ◽  
Temperature Fields ◽  
Head Part ◽  
Pressurized Water ◽  
Area Of Interest ◽  
Water Reactors

The aim of this work is to investigate experimentally the increase of mixing phenomenon in a coolant flow in order to improve the heat transfer, the economical operation and the structural integrity of Light Water Reactors-Pressurized Water Reactors (LWRs-PWRs). Thus the parameters related to the heat transfer process in the system will be investigated. Data from a set of experiments, obtained by using high precision measurement techniques, Particle Image Velocimetry and Planar Laser-Induced Fluorescence (PIV and PLIF, respectively) are to improve the basic understanding of turbulent mixing phenomenon and to provide data for CFD code validation. The coolant mixing phenomenon in the head part of a fuel assembly which includes spacer grids has been investigated (the fuel simulator has half-length of a VVER 440 reactor fuel). The two-dimensional velocity vector and temperature fields in the area of interest are obtained by PIV and PLIF technique, respectively. The measurements of the turbulent flow in the regular tube channel around the thermocouple proved that there is rotation and asymmetry in the coolant flow caused by the mixing grid and the geometrical asymmetry of the fuel bundle. Both PIV and PLIF results showed that at the level of the core exit thermocouple the coolant is homogeneous. The discrepancies that could exist between the outlet average temperature of the coolant and the temperature at in-core thermocouple were clarified. Results of the applied techniques showed that both of them can be used as good provider for data base and to validate CFD results.

3D Simulation of a Peripheral Adapter J-Groove Attachment Weld in a Vessel Head

2014 ◽  
Author(s):  
Philippe Mourgue ◽  
Vincent Robin ◽  
Philippe Gilles ◽  
Florence Gommez ◽  
Alexandre Brosse ◽  
...  
Keyword(s):  
Complex Geometry ◽  
3D Simulation ◽  
Welding Parameters ◽  
Manufacturing Processes ◽  
Pressurized Water ◽  
Control Rod ◽  
Sequence Modeling ◽  
Component Design ◽  
Water Reactors ◽  
Welding Processes

In Pressurized Water Reactors, most of heavy components and pipes have a large thickness and their manufacturing processes often require multi-pass welding. Despite the stiffness of these components, the distortion issue may be important for operational requirements (e.g. misalignment) or controllability reasons (Non Destructive Examinations have to be achievable, therefore ovalization should be limited). These requirements may be difficult to achieve by simply adjusting welding processes. Indeed because of the complexity of mechanisms involved during a welding operation and the high number of influencing parameters, this process is still essentially based on the experience of the welder. Furthermore the experimental estimation of the stress and distortion level in the component remains a difficult task that is subject to errors even if techniques are currently improved to become more accurate. These are the reasons why AREVA has put a large effort to improve welding numerical simulations, in order to have a better understanding of the involved physical phenomena and also to predict the residual state through the structure. Computational welding mechanics is used to qualify the manufacturing processes in the very early phase of the welded component design. Within the framework of a R&D program whose main objective was to improve tools for the numerical simulation of welding regarding industrial needs, AREVA has decided to validate new methodologies based on 3D computation by comparison with measurements. For this validation task the chosen industrial demonstrator was a Control Rod Drive Mechanism (CRDM) Nozzle with a J-groove attachment weld to the vessel head. For such an application, operations of post-joining straightening have to be limited, if not prohibited, because of their cost or the impossibility to use them in front of a steel giant. The control of distortion during welding operations is a key issue for which simulation can be of great help. Regarding distortion issues, both accurate metal deposit sequence modeling and respect of the real welding parameters are mandatory, especially for multi-pass operation on such a complex geometry. The aim of this paper is to present the simulation of the distortion of a peripheral adapter J-groove attachment weld mock-up. This new full 3D simulation improves the result of the previous one based on lumped pass deposits. It is the result of a fruitful collaboration between AREVA and ESI-Group.

The Sensitivity of Risk-Informed Reactor Structural Integrity Analysis Results to Various Interpretations of Warm Pre-Stress

2009 ◽  
Author(s):  
Mark T. EricksonKirk ◽  
Terry L. Dickson
Keyword(s):  
Structural Integrity ◽  
The United States ◽  
Integrated Analysis ◽  
Total Risk ◽  
Pressurized Water ◽  
Best Estimate ◽  
Water Reactors ◽  
Using Data ◽  
Accident Scenarios ◽  
Reactor Operating

Warm pre-stress, or WPS, is a phenomenon by which the apparent fracture toughness of ferritic steel can be elevated in the fracture mode transition if crack is first “pre-stressed” at an elevated temperature. Taking proper account of WPS is important to the accurate modeling of the postulated accident scenarios that, collectively, are referred to as pressurized thermal shock, and to the accurate modeling of routine cool-down transients. For both accident and routine cool-downs the transients begin at the reactor operating temperature (approximately 290°C for pressurized water reactors in the United States) and proceed to colder temperatures as time advances. The probabilistic fracture mechanics code FAVOR, which is being used by the NRC to provide the technical basis for risk-informed revisions of 10 CFR 50.61 and 10 CFR 50 Appendix G, adopts a model of WPS as part of its fracture driving force module. In this paper we assess the conservatism inherent to the FAVOR WPS model relative to a best-estimate WPS model constructed using data recently produced by the European Commission “SMILE” project and published by Moinereau and colleagues. Assessments of the conservatisms inherent to the so-called “conservative principle” WPS model, and also to a classic LEFM model that does not credit WPS are also made. The data presented herein demonstrate that, for an integrated analysis of PTS risk, the FAVOR and conservative principle WPS models both over-estimate the vessel failure risk by a factor of between 2 and 3× relative to the best estimate model. Our examination of the effect of WPS models on the predictions of individual transients reveals that for the severe transients that dominate risk there is little difference (usually less than 4×) between the conditional probabilities of crack initiation and of through wall cracking predicted by the different WPS models. There are considerable differences in the predictions of the various WPS and non-WPS models for low severity transients, however, the contribution of these transients to the total risk of vessel failure is small.

Coordinated Research Projects at IAEA Concerning RPV and Structural Integrity

2006 ◽  
Author(s):  
William Server ◽  
Timothy Hardin ◽  
Milan Brumovsky´
Keyword(s):  
Ferritic Steel ◽  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Current Status ◽  
Research Projects ◽  
Energy Agency ◽  
Pressurized Water ◽  
Toughness Testing ◽  
Water Reactors

The International Atomic Energy Agency (IAEA) has had a series of reactor pressure vessel (RPV) structural integrity programs that started back in the 1970s. These Coordinated Research Projects most recently have focused on use of the Master Curve fracture toughness testing approach for RPV and other ferritic steel components and on the issue of pressurized thermal shock (PTS) in operating pressurized water reactors. This paper will provide the current status for these projects and discuss the implications for improved safety of key ferritic steel components in nuclear power plants (NPPs).

A Simplified Evaluation of Stress Intensity Factors for a Small Diameter Pipe Penetrating a Thick Plate

2003 ◽  
Author(s):  
John M. Emery ◽  
Katsumasa Miyazaki ◽  
Anthony R. Ingraffea
Keyword(s):  
Stress Intensity ◽  
Stress Intensity Factors ◽  
Complex Geometry ◽  
Three Dimensional ◽  
Small Diameter ◽  
Boiling Water Reactors ◽  
Intensity Factors ◽  
Pressurized Water ◽  
Diameter Pipe ◽  
Water Reactors

Recent trouble with stress corrosion cracking in the internal parts of boiling water reactors and similar issues found in pressurized water reactors has prompted interest in developing simplified methods to determine stress intensity factors for such cracks. Currently, there are many practical and accurate simplified methods to calculate stress intensity factors for a surface crack in plates and pipes. However, there are none that deal with the complex geometry that can arise within the reactors. The complex geometry found within the vessels often entails reentrant comers, welds, holes, and other stress amplifiers. This paper sets forth a means by which some commonly known and accepted simplified solutions to cracks in pipes and plates can be modified to improve the accuracy of stress intensity factors when applied to this complex geometry. The effort to do so included axisymmetric and fully three-dimensional numerical modeling of both the cracked and uncracked body with a variety of assumed surface flaws. It was confirmed that the simplified methods lead to exceedingly conservative estimates for the stress intensity factors of the complex geometry. Finally, a correction factor based on the axisymmetric analyses was applied to the three-dimensional results to improve the accuracy of the simplified solutions.

Assessment of Post-CHF Heat Transfer Models for the SPACE Code

2009 ◽  
Author(s):  
K. Y. Choi ◽  
B. J. Yun ◽  
H. S. Park ◽  
S. K. Moon ◽  
K. D. Kim
Keyword(s):  
Heat Transfer ◽  
Mode Selection ◽  
Film Boiling ◽  
Model Assessment ◽  
Field Equations ◽  
Pressurized Water ◽  
Water Reactors ◽  
And Performance ◽  
Physical Phenomena ◽  
Transfer Models

The SPACE (Safety and Performance Analysis Code) which is based on a multi-dimensional two-fluid, three-field model is under development for a licensing purpose of pressurized water reactors in Korea. A total of 12 wall-to-fluid heat transfer modes were defined and a heat transfer mode selection logic was developed according to the noncondensable gas quality, the void fraction, the degree of subcooling and the wall temperature. Among the constitutive equations of the SPACE code, post-CHF heat transfer models are considered to have the most significant uncertainties because their physical phenomena are not fully understood yet. Though a variety of models and correlations for the post-CHF heat transfer are available, there is no reliable model to reproduce the heat transfer rate as realistically as possible. Several post-CHF heat transfer models are implemented into the SPACE code; critical heat, transition and film boiling models. The present paper describes the model assessment progress which was done for the transition and film boiling models of the SPACE code. A heat flux partition into the continuous liquid, entrained droplet and vapor fields should be taken into account in order to be in line with physical phenomena as the SPACE code has three-field equations in its hydraulic solver. Existing energy partitioning methods were peer-reviewed in order to determine the best model which can be applicable to the SPACE code. The present work will help to consolidate the developed wall-to-fluid heat transfer package of the SPACE code.

Development of Simulation Codes for Some Accidents in Pressurized Water Reactors

2008 ◽  
Author(s):  
X. B. Yang ◽  
G. H. Su ◽  
S. Z. Qiu
Keyword(s):  
Heat Transfer ◽  
Physical Property ◽  
Friction Model ◽  
Physical Models ◽  
Pressurized Water Reactors ◽  
Transfer Model ◽  
Feed Water ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactors

An analysis code has been developed for evaluating the transient thermo-hydraulic behaviors of the pressurized water reactor system. A series of mathematical and physical models is considered in this code, such as the point reactor neutron kinetics model, the heat transfer model, the friction model, the thermo-physical property model and so on. All possible flow and heat transfer conditions in some accidents have been considered and their corresponding models are supplied. Gear’s method is adopted for a better solution to the stiff equations. In this paper, some general accidents in the pressurized water reactors have been investigated, including the station blackout accident (SBO), the loss of flow accident (LOFA), the loss of feed water accident (LOFWA) and the reactivity insertion accident (RIA). The calculated results have been verified by the RELAP5/Mod3 and the results are satisfactory.

Flow and Heat Transfer Simulation in a Complete PWR Fuel Assembly Using Wall-Modelled Rans

2021 ◽  
Author(s):  
Blaž Mikuž ◽  
Ferry Roelofs
Keyword(s):  
Heat Transfer ◽  
Pressure Drop ◽  
Flow Field ◽  
Fuel Assembly ◽  
Hot Spots ◽  
Complex Geometry ◽  
Computational Domain ◽  
Pressurized Water ◽  
Flow And Heat Transfer ◽  
Significant Difference

Abstract Reproduction of turbulent flow and heat transfer inside a pressurized water reactor (PWR) fuel assembly is a challenging task due to the complex geometry and the huge computational domain. Capability of a wall-modelled RANS approach has been examined, which had already been validated against the measurements of the MATiS-H experiment. The method is here expanded to a larger computational domain aiming to reproduce flow and thermal field in the entire PWR fuel assembly. Namely, in the first part of the present study, wall-modelled RANS is performed in a relatively short section of the representative PWR fuel assembly containing one single mixing grid with an array of 15×15 fuel rods. Linear and nonlinear eddy-viscosity turbulence models have been applied, however no significant difference is observed in the predicted pressure drop in the fuel assembly. The obtained predictions revealed an interesting pattern of swirl flow as well as diagonal cross flow downstream the mixing grid, which is driven by the applied design of split-type mixing vanes. In the second part, the computational model is extended to a domain representative of a complete PWR fuel assembly with ten mixing grids, inlet and outlet sections. Pressure drop and flow field are analysed together with the predicted temperature and potential hot spots. In spite of a relatively coarse spatial resolution of the applied approach, the wall-modelled RANS provided promising results at least for the qualitative prediction of the pressure, flow field and location of hot spots.

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