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


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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.



Edited by:

Paul P. Lin and Chunliang Zhang




H. H. Qi and Z. X. Zeng, "Nonlinear Dynamic Analysis of RCS LOCA Based on Secondary Development of ANSYS", Applied Mechanics and Materials, Vols. 105-107, pp. 334-338, 2012

Online since:

September 2011




[1] Standard Review Plan. U.S. NRC(1997).

[2] Prinja NK, Chitkara NR. Post Collapse Cross2sectional Flattening of Thick Pipes in Plastic Bending. Nucl Eng Des, Vol. 83(1984), pp.113-121.


[3] Prinja NK, Chitkara NR. Large Rotation, Large Strain Analysis of Pipe Whip With Flow Choking. Nucl Eng Des, Vol. 93(1986), pp.69-81.


[4] Kurihara R. Experimental Studies of 42inch Pipe Whip Test Under BWR LOCA Conditions. Nucl Eng Des, Vol. 76(1983), pp.23-33.

[5] Hua Yun-long, Yu Tong-xi. Analytical and Numerical Methods of Pipe Whip Problems in Nucear Power plants(in Chinese). Computational Structural Mechanics and Applications, Vol. 1(1988), pp.105-112.

[6] Zhao Guoqiao. Dynamic Response Analysis of Nuclear Reactor Piping System(in Chinese). Beijing: Department of Engineering Mechanics, Tsinghua University, (1992).

[7] Yu Ru-hong. An Aanlysis of Primary Reactor Coolant Loop Systems for the Protective Design Against Pipe Break Effects(in Chinese). Nuclear Power Engineering, Vol. 7(1986), pp.49-56.

[8] Zhang Xi-wen, etc. Nonlinear Dynamic Response Analysis in Piping System for a Loss of Coolant Accident in Primary Loop of Pressurized Water Reactor(in Chinese). Atomic Energy Science and Technology, Vol. 34(2000), pp.385-390.

[9] Mao Qing, Zeng Zhong-xiu, etc. Preliminary Study of Reactor Coolant System LOCA Nonlinear Dynamic Analysis(in Chinese), the 11th Proceeding of Structural Mechanics in Reactor Technology, (2000).

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