scholarly journals Investigation of Transient Flow in a Centrifugal Charging Pump during Charging Operating Process

2014 ◽  
Vol 6 ◽  
pp. 860257 ◽  
Author(s):  
Fan Zhang ◽  
Shouqi Yuan ◽  
Qiang Fu ◽  
Feng Hong ◽  
Jianping Yuan
Keyword(s):  
Numerical Simulation ◽  
Transient Process ◽  
Flow Rate ◽  
Nuclear Power ◽  
Power Plants ◽  
Large Scale ◽  
Transient Flow ◽  
Internal Flow ◽  
Operating Conditions ◽  
Variable Operating Conditions

Centrifugal charging pumps are important components of nuclear power plants and must be operated under multioperating conditions for the requirements of the system. In order to investigate the internal flow mechanism of the centrifugal charging pump during the transient transition process of charging operating from Q = 34 m3/h to Q = 110 m3/h, numerical simulation and experiment are implemented in this study. The relationship between flow rate and time is obtained from the experiment and worked as the boundary condition to accurately accomplish the numerical simulation during the transient process. External and internal characteristics under the variable operating conditions are analyzed through the transient simulation. The results show that the liquid viscosity, large scale vortexes exist in the flow passages in the beginning of the variable operating conditions, which indicates flow separation and the sudden changes in direction of velocity. As the flow rate increases gradually, the flow angles of the fluid in inlet accelerate correspondingly and the flow along the blade is more uniform, which leads to a decrease and movements of the vortexes. The contents of the current work can provide references for the design optimization and fluid control of the pump used in the transient process of variable operating conditions.

scholarly journals 1D-3D coupling investigation of hydraulic transient for power-supply failure of centrifugal pump-pipe system

2019 ◽  
Vol 21 (5) ◽  
pp. 708-726 ◽  
Author(s):  
Xiaoqin Li ◽  
Xuelin Tang ◽  
Min Zhu ◽  
Xiaoyan Shi
Keyword(s):  
Turbulent Flows ◽  
Flow Rate ◽  
Power Supply ◽  
Nuclear Power ◽  
Transient Flow ◽  
Internal Flow ◽  
Transition Process ◽  
Pipeline System ◽  
Pipe System ◽  
Rotating Speed

Abstract In the pumping station, the main feedwater system and the reactor system of nuclear power plant, power-supply failure causes strong hydraulic transients. One-dimensional method of characteristics (1D-MOC) is used to calculate the transient process in the pipeline system while three-dimensional (3D) computational fluid dynamics is employed to analyze the turbulent flows inside the pump and to obtain the performance parameters of the pump, and the data exchanges on the boundary conditions of the shared interface between 1D and 3D domains are updated based on the MpCCI platform. Based on the equation of motion of the pump motion parts, the relationship between the external characteristics and the internal flow field in the pump is further investigated because the dynamic behavior of the pump and the detailed fluid field evolutions inside the pump are captured during the transition process, and the transient flow rate, rotating speed, and pressure inside the impeller are comprehensively investigated. Meanwhile, compared with the data gained by experiment and traditional 1D-MOC, the relative errors of rotating speed and the flow rate obtained by 1D-3D coupling method are smaller than those by 1D-MOC. Furthermore, the influences of the main coupling parameters and coupling modes on the calculation results are analyzed, and the cause of the deviation is further explained.

Analyses of Off-Design Point Performances of Simple and Intercooled Brayton Helium Recuperated Gas Turbine Cycles for Generation IV Nuclear Power Plants

2017 ◽  
Author(s):  
Arnold Gad-Briggs ◽  
Pericles Pilidis
Keyword(s):  
Mass Flow Rate ◽  
Mass Flow ◽  
Flow Rate ◽  
Nuclear Power ◽  
Power Plants ◽  
Pressure Ratio ◽  
Operating Conditions ◽  
Outlet Temperature ◽  
Design Point ◽  
Part Load

The Design Point (DP) performance of a Nuclear Power Plant (NPP) is fairly straightforward to establish for a given mass flow rate, turbomachinery compressor Pressure Ratio (PR) and reactor Core Outlet Temperature (COT). The plant components are optimum for that point but this is no longer the case if the plant’s operating conditions are changed for part-load performance. Data from tests or previous operating experiences are useful in determining typical part load performance of components based on characteristic maps. However, when individual components are linked in a plant, the range of operating points for part load performances are severely reduced. The main objective of this study is to derive Off-Design Points (ODPs) for the Simple Cycle Recuperated (SCR) and Intercooled Cycle Recuperated (ICR) when considering a temperature range of −35 to 50°C and COTs between 750 to 1000°C, using a modelling & performance simulation tool designed specifically for this study, which calculates the best operational equilibrium ODPs that are critical to the economics of the NPP. Results show that the recuperator High-Pressure (HP) side and reactor pressure losses alter the actual operating parameters (mass flow rate and compressor PR). The SCR yielded a drop in plant cycle efficiency of 1% for a 4% pressure loss in comparison to the ICR (5%) for the same amount of recuperator HP losses. Other parameters such as the precooler and recuperator Low-Pressure (LP) losses still retain the same operating inlet PRs and mass flow rates regardless of the magnitude of the losses. In the absence of characteristic maps in the public domain, the ODPs have been used to produce characteristic trend maps for first order ODP calculations. The analyses intend to aid the development of cycles for Generation IV NPPs specifically Gas Cooled Fast Reactors (GFRs) and Very High Temperature Reactors (VHTRs), where helium is the coolant.

Evaluation of IET Facility Applicability on Simulating SBLOCA in Large-Scale Passive PWR Plant

2018 ◽  
Author(s):  
Haozheng Kong ◽  
Bo Kuang ◽  
Pengfei Liu ◽  
Xia Lu ◽  
Yi Yao ◽  
...  
Keyword(s):  
Flow Rate ◽  
Nuclear Power ◽  
Power Plants ◽  
Large Scale ◽  
Scaling Analysis ◽  
Flow Area ◽  
Injection Flow ◽  
Parameter Evaluation ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant

Focused on the overall experimental verification of important phenomena in small break loss of coolant accident (SBLOCA) of large-scale passive PWR plants, this paper put forward a top-down and bottom-up scaling analysis of SBLOCA process based on H2TS method, by which the scaling conditions of IET facility would be figured out to match with the conditions in real nuclear power plants. Meanwhile, parameter evaluation was conducted between IET facility and real power plants, and finally the results proved that the IET facility could, to a certain degree, represent the conditions in real plants in terms of scaling analysis, and the test data of facility could be applicable for the SBLOCA of real plants. Further, in this paper, scenarios of SBLOCA were simulated and compared between IET facility and AP1000 nuclear power plant by the use of Relap5/MOD3 program. Then, after the conversion of flow area and time ratio, the comparison of results between those two systems was conducted in terms of pressure of reactor coolant system (RCS), injection flow rate of CMT and steam flow rate through ADS-1/2/3/4 valves, which also proved that the IET facility could simulate the SBLOCA of real plants well.

scholarly journals On the Root Causes of the Fukushima Daiichi Disaster from the Perspective of High Complexity and Tight Coupling in Large-Scale Systems

Symmetry ◽  
2021 ◽  
Vol 13 (3) ◽  
pp. 414
Author(s):  
Atsuo Murata ◽  
Waldemar Karwowski
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Large Scale ◽  
High Complexity ◽  
Tight Coupling ◽  
Large Scale Systems ◽  
Safety Issues ◽  
Loss Of Coolant ◽  
Station Blackout ◽  
Fukushima Daiichi

This study explores the root causes of the Fukushima Daiichi disaster and discusses how the complexity and tight coupling in large-scale systems should be reduced under emergencies such as station blackout (SBO) to prevent future disasters. First, on the basis of a summary of the published literature on the Fukushima Daiichi disaster, we found that the direct causes (i.e., malfunctions and problems) included overlooking the loss of coolant and the nuclear reactor’s failure to cool down. Second, we verified that two characteristics proposed in “normal accident” theory—high complexity and tight coupling—underlay each of the direct causes. These two characteristics were found to have made emergency management more challenging. We discuss how such disasters in large-scale systems with high complexity and tight coupling could be prevented through an organizational and managerial approach that can remove asymmetry of authority and information and foster a climate of openly discussing critical safety issues in nuclear power plants.

scholarly journals Optimal Design of Slit Impeller for Low Specific Speed Centrifugal Pump Based on Orthogonal Test

2021 ◽  
Vol 9 (2) ◽  
pp. 121
Author(s):  
Yang Yang ◽  
Ling Zhou ◽  
Hongtao Zhou ◽  
Wanning Lv ◽  
Jian Wang ◽  
...  
Keyword(s):  
Numerical Simulation ◽  
Centrifugal Pump ◽  
Large Scale ◽  
Internal Flow ◽  
Orthogonal Test ◽  
Centrifugal Pumps ◽  
Initial Model ◽  
Low Specific Speed ◽  
Specific Speed ◽  
Four Levels

Marine centrifugal pumps are mostly used on board ship, for transferring liquid from one point to another. Based on the combination of orthogonal testing and numerical simulation, this paper optimizes the structure of a drainage trough for a typical low-specific speed centrifugal pump, determines the priority of the various geometric factors of the drainage trough on the pump performance, and obtains the optimal impeller drainage trough scheme. The influence of drainage tank structure on the internal flow of a low-specific speed centrifugal pump is also analyzed. First, based on the experimental validation of the initial model, it is determined that the numerical simulation method used in this paper is highly accurate in predicting the performance of low-specific speed centrifugal pumps. Secondly, based on the three factors and four levels of the impeller drainage trough in the orthogonal test, the orthogonal test plan is determined and the orthogonal test results are analyzed. This work found that slit diameter and slit width have a large impact on the performance of low-specific speed centrifugal pumps, while long and short vane lap lengths have less impact. Finally, we compared the internal flow distribution between the initial model and the optimized model, and found that the slit structure could effectively reduce the pressure difference between the suction side and the pressure side of the blade. By weakening the large-scale vortex in the flow path and reducing the hydraulic losses, the drainage trough impellers obtained based on orthogonal tests can significantly improve the hydraulic efficiency of low-specific speed centrifugal pumps.

Finite Element Analyses of Residual Stresses in Typical Welded Joints Used in Nuclear Power Plants and Comparisons With Experiments

2010 ◽  
Author(s):  
Dean Deng ◽  
Kazuo Ogawa ◽  
Nobuyoshi Yanagida ◽  
Koichi Saito
Keyword(s):  
Numerical Simulation ◽  
Residual Stress ◽  
Residual Stresses ◽  
Welded Joints ◽  
Nuclear Power ◽  
Power Plants ◽  
Butt Joint ◽  
Welding Residual Stress ◽  
Dissimilar Metal ◽  
Water Reactors

Recent discoveries of stress corrosion cracking (SCC) at nickel-based metals in pressurized water reactors (PWRs) and boiling water reactors (BWRs) have raised concerns about safety and integrity of plant components. It has been recognized that welding residual stress is an important factor causing the issue of SCC in a weldment. In this study, both numerical simulation technology and experimental method were employed to investigate the characteristics of welding residual stress distribution in several typical welded joints, which are used in nuclear power plants. These joints include a thick plate butt-welded Alloy 600 joint, a dissimilar metal J-groove set-in joint and a dissimilar metal girth-butt joint. First of all, numerical simulation technology was used to predict welding residual stresses in these three joints, and the influence of heat source model on welding residual stress was examined. Meanwhile, the influence of other thermal processes such as cladding, buttering and heat treatment on the final residual stresses in the dissimilar metal girth-butt joint was also clarified. Secondly, we also measured the residual stresses in three corresponding mock-ups. Finally, the comparisons of the simulation results and the measured data have shed light on how to effectively simulate welding residual stress in these typical joints.

Development of 1500-r/min 75-Inch Ultra-Long Last Stage Blades for Nuclear Steam Turbine

2017 ◽  
Author(s):  
Deqi Yu ◽  
Jiandao Yang ◽  
Wei Lu ◽  
Daiwei Zhou ◽  
Kai Cheng ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Large Scale ◽  
Vibration Monitoring ◽  
Steam Turbines ◽  
Element Analysis ◽  
Rotating Blade ◽  
Lifetime Analysis ◽  
Computational Fluid Dynamics Cfd ◽  
Last Stage

The 1500-r/min 1905mm (75inch) ultra-long last three stage blades for half-speed large-scale nuclear steam turbines of 3rd generation nuclear power plants have been developed with the application of new design features and Computer-Aided-Engineering (CAE) technologies. The last stage rotating blade was designed with an integral shroud, snubber and fir-tree root. During operation, the adjacent blades are continuously coupled by the centrifugal force. It is designed that the adjacent shrouds and snubbers of each blade can provide additional structural damping to minimize the dynamic stress of the blade. In order to meet the blade development requirements, the quasi-3D aerodynamic method was used to obtain the preliminary flow path design for the last three stages in LP (Low-pressure) casing and the airfoil of last stage rotating blade was optimized as well to minimize its centrifugal stress. The latest CAE technologies and approaches of Computational Fluid Dynamics (CFD), Finite Element Analysis (FEA) and Fatigue Lifetime Analysis (FLA) were applied to analyze and optimize the aerodynamic performance and reliability behavior of the blade structure. The blade was well tuned to avoid any possible excitation and resonant vibration. The blades and test rotor have been manufactured and the rotating vibration test with the vibration monitoring had been carried out in the verification tests.

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