Nonlinear Dynamic Analysis of RCS LOCA Based on Secondary Development of ANSYS

2011 ◽  
Vol 105-107 ◽  
pp. 334-338
Author(s):  
Huan Huan Qi ◽  
Zhong Xiu Zeng
Keyword(s):  
Nonlinear Dynamic ◽  
Nuclear Reactor ◽  
System Modeling ◽  
Nonlinear Dynamic Analysis ◽  
Secondary Development ◽  
Modular System ◽  
Coolant Pump ◽  
Modular Modeling ◽  
Reactor Coolant ◽  
Loss Of Coolant Accident

Research of loss of coolant accident (LOCA) postulated on the nuclear reactor coolant system (RCS) was investigated with ANSYS program. Secondary development of ANSYS was performed to form the customized module for implementing effective and efficient RCS LOCA nonlinear analysis. A standard analysis procedure was established. It has following functions, such as parameters and modular system modeling, automatically set break, static pilot analysis, LOCA nonlinear dynamic calculation and automatic reports generation. Main pipes, Steam generator and reactor coolant pump models were molded by parameterized modular modeling method. Those models considered the nonlinear factors, such as material nonlinearities, gap and so on, constructed component model libraries of RCS. Comparing the results calculated by ANSYS and program-specific, it is showed that the results are generally consistent.

Research on Modular Modeling Method of Reactor Coolant System Based on THEATRe

2014 ◽  
Author(s):  
Bo Shi ◽  
Zhao-Fei Tian
Keyword(s):  
Real Time ◽  
Data Exchange ◽  
Reactor Core ◽  
Simulation Software ◽  
Modeling Method ◽  
Time Step ◽  
Coolant Pump ◽  
Modular Modeling ◽  
Reactor Coolant ◽  
Coolant System

At present, research on the reactor coolant system is less yet, though modular modeling method has been widely used in the second-loop system of reactor. This paper takes the reactor coolant system of Qinshan-1 nuclear power plant as the object of study, analyses and researches on modular modeling method of reactor coolant system based on THEATRe, which is a large Thermal-Hydraulic real time simulation software developed by GSE Company and adopts NMNP (Nodal Momentum Nodal Pressure) solving method. This research establishes the modular model of the reactor coolant system equipments (including reactor core, main coolant pump, pressurizer, steam generator) using the THEATRe code. Due to each module is wrote into through different input cards, they can be solved by using their own matrix of velocity-pressure to guarantee the independence of the numerical calculation for different modular modules. THEATRe code does not have its own TDV like relap-5, meanwhile it also needs to ensure the pressurizer module can play a role in the multi-pressure node system. So this paper modifies solving method of the THEATRe source code to get suitable pressure boundary and flux boundary for RCS equipment modular module, and selects reasonable time step and data exchange frequency to achieve the data exchange of boundary pressure, flux and enthalpy among the equipment modules, which lays the foundation of establishing the real-time modular simulation model of the reactor coolant system in the future.

scholarly journals Flow Unsteadiness and Pressure Pulsations in a Nuclear Reactor Coolant Pump

2016 ◽  
Vol 62 (4) ◽  
pp. 231-242 ◽  
Author(s):  
Dan Ni ◽  
Minguan Yang ◽  
Bo Gao ◽  
Ning Zhang ◽  
Zhong Li
Keyword(s):  
Nuclear Reactor ◽  
Pressure Pulsations ◽  
Coolant Pump ◽  
Reactor Coolant ◽  
Flow Unsteadiness

Influence of Guide Vane Wrap Angle Key Design Parameters on Hydraulic Performance of Nuclear Reactor Coolant Pump

2014 ◽  
Vol 721 ◽  
pp. 73-77 ◽  
Author(s):  
Wei Nan Jin ◽  
Rong Xie ◽  
Mu Ting Hao ◽  
Xiao Fang Wang
Keyword(s):  
Flow Field ◽  
Nuclear Reactor ◽  
Guide Vane ◽  
Three Dimensional ◽  
Internal Flow ◽  
Hydraulic Performance ◽  
Design Parameters ◽  
Coolant Pump ◽  
Reactor Coolant ◽  
Wrap Angle

To study the effects of guide vane with different vane wrap angles and relative positions of outlet edge on hydraulic performance of nuclear reactor coolant pump, three-dimensional steady numerical simulations were performed by using CFD commercial software Numeca. The results show that the vane wrap angle changes the head and power characteristics by changing the relative velocity angle in vane outlet. The inner flow field changes while the wrap angle changes. With the wrap angle increases, the shock loss in volute is reducing, but the friction loss in vane passages is getting large. So there exists an optimum wrap angle and relative positions of outlet edge that corresponds to the highest efficiency of a pump. Numerical simulation is performed with the two key design parameters optimized through surrogate model, the internal flow field is improved and then the hydraulic efficiency is improved.

scholarly journals Influence of Circumferential Placement Position of Guide Vanes on Performance and Dynamic Characteristics of Nuclear Reactor Coolant Pump

2020 ◽  
Vol 2020 ◽  
pp. 1-12
Author(s):  
Xiaorui Cheng ◽  
Boru Lv ◽  
Chenying Ji ◽  
Ningning Jia ◽  
Dorah N
Keyword(s):  
Nuclear Reactor ◽  
Guide Vane ◽  
Equivalent Stress ◽  
Hydraulic Loss ◽  
Stress Strain ◽  
Coolant Pump ◽  
Guide Vanes ◽  
Reactor Coolant ◽  
Maximum Equivalent Stress ◽  
Exit Edge

In order to study the influence of the circumferential placement position of the guide vane on the flow field and stress-strain of a nuclear reactor coolant pump, the CAP1400 nuclear reactor coolant pump is taken as the research object. Based on numerical calculation and test results, the influence of circumferential placement position of the guide vane on the performance of the nuclear reactor coolant pump and stress-strain of guide vanes are analyzed by the unidirectional fluid-solid coupling method. The results show that the physical model and calculation method used in the study can accurately reflect the influence of the circumferential placement position of the guide vane on the nuclear reactor coolant pump. In the design condition, guide vane position has a great influence on the nuclear reactor coolant pump efficiency value, suction surface of the guide vane blade, and the maximum equivalent stress on the hub. However, it has a weak effect on the head value, pressure surface of the guide vane blade, and the maximum equivalent stress on the shroud. When the center line of the outlet diffuser channel of the case is located at the center of the outlet of flow channel of the guide vane, it is an optimal guide vane circumferential placement position, which can reduce the hydraulic loss of half of the case. Finally, it is found that the high stress concentration area is at the intersection of the exit edge of the vane blade and the front and rear cover, and the exit edge of the guide vane blade and its intersection with the front cover are areas where the strength damage is most likely to occur. This study provides a reference for nuclear reactor coolant pump installation, shock absorption design, and structural optimization.

Sign in / Sign up

Export Citation Format

Share Document