Numerical Simulation of Stochastic Point Kinetics Equation in the Dynamical System of Nuclear Reactor

The simulation of 3D thermal-hydraulic problem for the pool type fast reactors, is one of the necessary and great importance. Most system codes can’t be used to simulate multi-dimensional thermal-hydraulics problems, whereas, the CFD method is suitable to deal with these type of simulation challenges. Based on the CFD method, a neutronics and thermohydraulic coupling code FLUENT/PK for nuclear reactor safety analysis by coupling the commercial CFD code FLUENT with the point kinetics model (PKM) and the pin thermal model (PTM) is developed by University of Science and Technology of China (USTC). The coupled code is verified by comparing with a series of benchmarks on beam interruptions in a lead-bismuth-cooled and MOX-fuelled accelerator-driven system. The variations of transient power, fuel temperature and outlet coolant temperature all agree well with the benchmark results. The validation results show that the code can be used to simulate the transient accidents of critical and sub-critical lead/lead-bismuth cooled reactors. Then this coupling code is used to evaluate the safety performance of MYRRHA (Multi-purpose Hybrid Research Reactor for High-tech Applications) at unprotected beam over-power (UBOP) accident, and M2LFR-1000 (Medium-size Modular Lead-cooled Fast Reactor) at the unprotected transient over-power (UTOP) and unprotected loss of flow (ULOF) accident. The transient power, the temperature of coolant and fuel and multi-dimensional flow phenomena in upper plenum and lower plenum are presented and discussed in this paper.

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Scaled neutron point kinetics (SUNPK) equations for nuclear reactor dynamics: 2D approximation

Annals of Nuclear Energy ◽

10.1016/j.anucene.2018.01.020 ◽

2018 ◽

Vol 115 ◽

pp. 377-386 ◽

Cited By ~ 3

Author(s):

Gilberto Espinosa-Paredes ◽

Carlos G. Aguilar-Madera

Keyword(s):

Nuclear Reactor ◽

Reactor Dynamics ◽

Point Kinetics

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Numerical simulation of a turbulent Lead Bismuth Eutectic flow inside a 19 pin nuclear reactor bundle with a four logarithmic parameter turbulence model

Journal of Physics Conference Series ◽

10.1088/1742-6596/1224/1/012030 ◽

2019 ◽

Vol 1224 ◽

pp. 012030

Author(s):

Andrea Chierici ◽

Leonardo Chirco ◽

Roberto Da Vià ◽

Sandro Manservisi

Keyword(s):

Numerical Simulation ◽

Turbulence Model ◽

Nuclear Reactor ◽

Lead Bismuth Eutectic

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A New Accurate Numerical Method Based on Shifted Chebyshev Series for Nuclear Reactor Dynamical Systems

Science and Technology of Nuclear Installations ◽

10.1155/2018/7105245 ◽

2018 ◽

Vol 2018 ◽

pp. 1-15 ◽

Cited By ~ 1

Author(s):

Yasser Mohamed Hamada

Keyword(s):

Nuclear Reactor ◽

Approximate Solutions ◽

Time Interval ◽

Neutron Density ◽

Chebyshev Series ◽

System A ◽

Temperature Feedback ◽

Exponential Rate Of Convergence ◽

Point Kinetics ◽

The Given

A new method based on shifted Chebyshev series of the first kind is introduced to solve stiff linear/nonlinear systems of the point kinetics equations. The total time interval is divided into equal step sizes to provide approximate solutions. The approximate solutions require determination of the series coefficients at each step. These coefficients can be determined by equating the high derivatives of the Chebyshev series with those obtained by the given system. A new recurrence relation is introduced to determine the series coefficients. A special transformation is applied on the independent variable to map the classical range of the Chebyshev series from [-1,1] to [0,h]. The method deals with the Chebyshev series as a finite difference method not as a spectral method. Stability of the method is discussed and it has proved that the method has an exponential rate of convergence. The method is applied to solve different problems of the point kinetics equations including step, ramp, and sinusoidal reactivities. Also, when the reactivity is dependent on the neutron density and step insertion with Newtonian temperature feedback reactivity and thermal hydraulics feedback are tested. Comparisons with the analytical and numerical methods confirm the validity and accuracy of the method.

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