CFD Analysis of PWR Surge Line Subjected to Thermal Stratification

2012 ◽  
Vol 468-471 ◽  
pp. 78-82 ◽  
Author(s):  
Athar Rasool ◽  
Zhong Ning Sun ◽  
Jian Jun Wang ◽  
Zeng Fang Ge ◽  
Majid Ali
Keyword(s):  
Thermal Stratification ◽  
Power Plants ◽  
Structural Integrity ◽  
Complex Geometry ◽  
Transient Temperature ◽  
3D Analysis ◽  
Pressurized Water ◽  
Temperature Distributions ◽  
Operational Life ◽  
Surge Line

Thermal stratification effects have been a great concern in a pressurizer surge line of pressurized water reactor (PWR) since 1988. These effects may damage the structural integrity and contribute in reducing the operational life time of pressurizer surge line. Several nuclear power plants operators have so far reported such mechanical damages. To realistically assess the structural integrity of pressurizer surge line subjected to thermal stratification, it is necessary to analyze the transient temperature distribution. Several researchers and scholars have carried out considerable efforts to determine the temperature distributions in the pressurizer surge line. In this study, an effort has been made to simulate the behavior of thermally stratified flow and predict the transient temperature distributions in the pressurizer surge line realistically. To obtain realistic results for such complex geometry of pressurizer surge line 3D analysis is performed using CFX commercially available CFD software. The transient temperature distributions obtained are presented and discussed.

CFD Analysis of Thermally Stratified Flow and Heat Transfer in a PWR Pressurizer Surge Line

2008 ◽  
Author(s):  
Dong Gu Kang ◽  
Jong Chull Jo
Keyword(s):  
Heat Transfer ◽  
Wall Thickness ◽  
Thermal Stratification ◽  
Thermal Stresses ◽  
Complex Geometry ◽  
Heat Transfer Analysis ◽  
Transient Temperature ◽  
Piping System ◽  
Temperature Distributions ◽  
Surge Line

Temperature gradients in the thermally stratified fluid flowing through a pipe may cause undesirable excessive thermal stresses at the pipe wall in the axial, circumferential, and radial directions, which can eventually lead to damages such as deformation, support failure, thermal fatigue, cracking, etc. to the piping systems. Several nuclear power plants have so far experienced such unwelcome mechanical damages to the pressurizer surge lines, feedwater nozzle, high pressure safety injection lines, or residual heat removal lines. In this regard, to determine the transient temperature distributions in the wall of a piping system subjected to internally thermal stratification with accuracy is the essential prerequisite for the assessment of the structural integrity of the piping system subjected to internally thermal stratification. In this study, to predict the transient temperature distributions in the wall of PWR pressurizer surge line with a complex geometry of 3-dimensionally bent piping realistically, 3-dimensional transient CFD calculations involving the conjugate heat transfer analysis are performed for the actual PWR pressurizer surge line subjected to stratified internal flows either during out-surge or in-surge operation using a commercial CFD code. In addition, the wall temperature distributions obtained by taking account of the existence of wall thickness as it is are compared with those by neglecting the existence of wall thickness to identify some requirements for a realistic and conservative thermal analysis.

The Surge Line Stress Analysis Model Setup With the Consideration of Thermal Stratification

2014 ◽  
Author(s):  
Jianjun Wang ◽  
Zengfang Ge ◽  
Zhongning Sun ◽  
Changqi Yan
Keyword(s):  
Stress Analysis ◽  
Transient Process ◽  
Thermal Stratification ◽  
Nuclear Power ◽  
Transient Temperature ◽  
Analysis Model ◽  
Pressurized Water ◽  
Line Model ◽  
Transient Temperature Distribution ◽  
Surge Line

In this paper, we deal with a typical pressurizer surge line in a conventional pressurized water reactor (PWR). This study is performed to develop an understanding of thermal stratification phenomenon, which may occur in the surge line during either normal condition or transient process, in the pressurizer surge line. The pressurizer surge line model of Daya Bay nuclear power plant is used as base analysis model, in which the hot leg is taken into account. The transient temperature distribution required to assess the phenomenon along the pressurizer surge line is obtained through CFD analysis technology using ANSYS FLUENT. The temperature loads are transferred to ANSYS Mechanical for stress evaluation for the heat up transient process. Subsequently, the usage factor is calculated on the basis of ASME Section-III design curve. The possible mitigation scheme for the thermal stratification phenomenon of changing the layout angles is also simulated and analyzed in detail. The results show that the thermal stratification phenomenon will occur both in normal operating condition and in heat up transient process. The circumfluent effect makes the thermal stratification phenomenon exhibit unique profile due to the introduction of the hot leg. The continuous spray mass flow rate may influence both the temperature difference and the occurrence range for the thermal stratification phenomenon. The stress analysis incorporating both temperature load and pressure load is performed for pressurizer surge line model with hot leg for the conservative and complete heat up case.

CFD Analysis of Thermally Stratified Flow and Conjugate Heat Transfer in a PWR Pressurizer Surgeline

2010 ◽  
Vol 132 (2) ◽  
Author(s):  
Jong Chull Jo ◽  
Dong Gu Kang
Keyword(s):  
Heat Transfer ◽  
Conjugate Heat Transfer ◽  
Thermal Stresses ◽  
Complex Geometry ◽  
Heat Removal ◽  
Heat Transfer Analysis ◽  
Transient Temperature ◽  
Piping System ◽  
Pressurized Water ◽  
Temperature Distributions

Temperature gradients in the thermally stratified fluid flowing through a pipe may cause undesirable excessive thermal stresses at the pipe wall in the axial, circumferential, and radial directions, which can eventually lead to damages such as deformation, support failure, thermal fatigue, cracking, etc., to the piping systems. Several nuclear power plants have so far experienced such unwelcome mechanical damages to the pressurizer surgeline, feedwater nozzle, high pressure safety injection lines, or residual heat removal lines at a pressurized water reactor (PWR). In this regard, determining with accuracy the transient temperature distributions in the wall of a piping system subjected to internally thermal stratification is the essential prerequisite for the assessment of the structural integrity of such a piping system. In this study, to realistically predict the transient temperature distributions in the wall of an actual PWR pressurizer surgeline with a complex geometry of three-dimensionally bent piping, three-dimensional transient computational fluid dynamics (CFD) calculations involving the conjugate heat transfer analysis are performed for the PWR pressurizer surgeline subjected to either out- or in-surge flows using a commercial CFD code. In addition, the wall temperature distributions obtained by taking into account the existence of wall thickness are compared with those by neglecting it to identify some requirements for a realistic and conservative thermal analysis from a safety viewpoint.

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.

Experimental Investigation on Flash Evaporation Related to Pipe Leakage

2019 ◽  
Vol 6 (1) ◽  
Author(s):  
Benan Cai ◽  
Qi Zhang ◽  
Yu Weng ◽  
Hongfang Gu ◽  
Haijun Wang
Keyword(s):  
High Temperature ◽  
Thermal Fatigue ◽  
Weber Number ◽  
Thermal Stratification ◽  
Nuclear Power ◽  
Power Plants ◽  
Response Speed ◽  
Flash Evaporation ◽  
Main Pipe ◽  
Surge Line

Abstract Pipelines such as the surge line and main pipe are easily subjected to thermal stratification and thermal fatigue as a result of the nonuniform temperature distribution in the nuclear power plants. When the surge line or main pipe subjected to thermal stratification and thermal fatigue keeps operating for long time, the pipe leakage may happen due to the existence of pipeline crack. When the fluids with high temperature and pressure leak in the crack, the water will evaporate quickly, which means this process belongs to spray flash evaporation process. The flash evaporation related to pipe leak was experimentally studied in the paper. The experiment was carried out under high temperature and high pressure with low spray rate. The temperature and relative humidity (T&H) variations over time were monitored in the experiment with installing T&H detectors. The T&H variations at different measurement positions and with different spray rates were analyzed, respectively. In addition, the effect of the dimensionless parameters including the Weber number and Jakob number was also investigated. Results indicated that the response speed increased with the increase of the spray flow rate. Higher Weber number and higher Jakob number led to higher evaporation rate. The slight pipe leakage can be predicted by using the (T&H) in the hazardous areas.

Analysis of Thermal Stratification and Fatigue Stress for Pressurizer Surge Line

2010 ◽  
Author(s):  
Xiaofei Yu ◽  
Yixiong Zhang
Keyword(s):  
Cross Section ◽  
Thermal Stratification ◽  
Thermal Stresses ◽  
Structural Integrity ◽  
Bending Moments ◽  
One Dimensional ◽  
Pipe System ◽  
Pipe Cross Section ◽  
Fatigue Stress ◽  
Surge Line

Thermal stratification of pressurizer surge line induced by the inside fluid brings on global bending moments, local thermal stresses, unexpected displacements and support loadings of the pipe system. In order to confirm the structural integrity of pressurizer surge line affected by thermal stratification, this paper theoretically establishes thermal stratified transient and studies the calculation method of thermal stratified stress. A costly three-dimensional computation is simplified into a combined 1D/2D technique. This technique uses a pipe cross-section for computation of local thermal stresses and represents the whole surge line with one-dimensional pipe elements. The 2D pipe cross-section model is used to compute elastic thermal stresses in plane strain condition. Symmetry allows half the cross-section to be considered. The one-dimensional pipe elements model gives the global bending moments including effects of usual thermal expansion and thermal stratification of each model nodes. This combined 1D/2D technique has been developed and implemented to analyze the thermal stratification and fatigue stress of pressurize surge line in this paper, using computer codes SYSTUS and ROCOCO. According to the mechanical analysis results of stratification, the maximum stress and cumulative usage factor are obtained. The stress and fatigue intensity of the surge line tallies with the correlative criterion.

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).

Proposal for Evaluation With Strain-Based Criterion of Impact Phenomena Between Steel Structure and Tornado Missile

2016 ◽  
Author(s):  
Yuko Sakamoto ◽  
Koji Shirai ◽  
Toshiko Udagawa ◽  
Shunsuke Kondo
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Impact Analysis ◽  
Structural Integrity ◽  
Steel Structures ◽  
Severe Accident ◽  
Numerical Code ◽  
Pressurized Water ◽  
Impact Phenomena ◽  
The Impact

In Japan, nuclear power plants must be protected from tornado missiles that are prescribed by Nuclear Regular Authority (NRA). When evaluating the structural integrity of steel structures in the plant with impact analysis by numerical code, strain-based criteria are appropriate because the tornado missiles have huge impact energy and may cause large deformation of the structures. As one of the strain-based criteria, the Japan Society of Mechanical Engineers (JSME) prescribes limiting triaxial strain for severe accident of Pressurized Water Reactor (PWR) steel containment. To confirm whether or not this criterion is appropriate to the evaluation of the impact phenomena between the steel structures and the tornado missiles, a free drop impact experiment to steel plates (carbon steel and austenitic stainless steel) was carried out with heavy weights imitated on one of the tornado missiles, followed by an impact analysis of the experiment with AUTODYN code and the JSME strain-based criterion. Consequently, it was confirmed that the strain-based criterion of JSME standard was for evaluating the fracture of steel structures caused by tornado missiles.

A New Method to Weaken Thermal Stratification Phenomena in Pipes of Nuclear Power Plant

2013 ◽  
Author(s):  
Shengfei Wang ◽  
Yuxin Pang ◽  
Xiaojing Li ◽  
Dandan Fu ◽  
Yang Li ◽  
...  
Keyword(s):  
Thermal Stratification ◽  
Nuclear Power ◽  
Thermal Stresses ◽  
Heat Pipes ◽  
New Method ◽  
Simple Experiment ◽  
Pressurized Water ◽  
Piping Systems ◽  
Surge Line ◽  
Water Reactors

Thermal stratification phenomena are observed in piping systems of pressurized water reactors, especially in the pressurizer surge line. As a result of the thermal stratification induced thermal stresses, fatigue problems can occur in the pipework. US NRC requirements have also identified flow stratification in surge lines as a phenomenon that must be considered in the design basis of surge lines. In this paper, a new method to reduce thermal stratification is proposed. As we all know, heat pipe is a simple device with no moving parts and can transfer large quantities of heat over fairly large distance. The new method is that using heat pipes to weaken the thermal stratification. In order to validate the new method, a simple experiment and theoretical analysis was taken. The results show that, the temperature difference of thermal stratification with heat pipes is smaller than the stratification without heat pipes. A design scheme was also given at the end of paper.

Comparison of Analytical Methods to Study the Impact of a Postulated Closure Head Drop Accident on the Structural Integrity of the Bottom-Mounted Instrumentation System

2010 ◽  
Author(s):  
William C. Castillo ◽  
Geoffrey M. Loy ◽  
Joseph M. Remic ◽  
David P. Molitoris ◽  
George J. Demetri ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
The United States ◽  
Reactor Vessel ◽  
Primary Coolant ◽  
Heavy Load ◽  
Pressurized Water ◽  
System Pressure ◽  
The Impact

During typical nuclear power plant refueling activities for a pressurized water reactor (PWR), the reactor vessel closure head assembly must be removed from the reactor vessel (RV), transported for storage, and returned to the RV after refueling. This is categorized as a critical heavy load lift in NUREG-0612 [1] because a drop accident could result in damage to the components required to cool the fuel in the RV core. In order to mitigate the potentially severe consequences of a closure head drop, the United States Nuclear Regulatory Commission (USNRC) has mandated that nuclear power plants upgrade to a single failure-proof crane, show single failure-proof crane equivalence, or perform a head drop analysis to demonstrate that the core remains covered with coolant and sufficient cooling is available after the head drop accident. The primary coolant-retaining components associated with the RV are the inlet and outlet nozzles and the hot and cold leg main loop piping. Typical head drop analyses have considered these components to ensure that their structural integrity is maintained. One coolant-retaining component that has not been included in head drop evaluations on a consistent basis is the bottom-mounted instrumentation (BMI) system. In a typical Westinghouse PWR, 50 to 60 BMI nozzles are connected through the bottom hemisphere of the RV to one-inch diameter guide tubes which run under the vessel to a seal table above. Failure of the BMI system has the potential to adversely affect core coolability, especially if multiple failures are postulated within the system. A study was performed to compare static and dynamic methods of analyzing the effects of a head drop accident on the structural integrity of the BMI system. This paper presents the results of that study and assesses the adequacy of each method. Acceptability of the BMI system pressure boundary is based on the Nuclear Energy Institute Initiative (NEI 08–05 [2]) criteria for coolant-retaining components, which are based on Section III, Appendix F of the ASME Code [3].

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