Research On Time Parallel Method Of Space Time Neutron Kinetics And Transient Charactericstic Analysis Of Traveling Wave Reactor

Neutron Kinetics is an important part of the reactor safety analysis.However,it's expensive to solve these equations,and it's solved sequentially by step by step usually,which is hard to use the multi-core computers.Travelling Wave Reactor(TWR)is an advanced and innovative reactor,which can breed fertile fuel into fissile fuel greatly.TWR can make use of natural or depleted uranium,and improve the utilzation of uranium resources.In this thesis,neutron kinetics and TWR are concerned,and the time integration methods of neturon kinetics and the transient characteristic of TWR will be studied in details.The main contents of this thesis are listed below:Firstly,the numerical solution of the neutron kinetics is studied.Four time integration methods like the backward Euler method,the backward differential formulas(BDF)and so on,are used to solve the neutron kinetics equations.The new adaptive time-step algorithm is implemented in the BDF,which improves the efficiency of the BDF.The numerical tests show that adaptive time-step BDF owns good accuracy and efficiency.Secondly,the time parallel integration method Parareal is researched and applied to the neutron kinetics equations.In tradition,the ODE equations are solved step by step,which can not utilize multi-core processors.Parareal algorithm is a parallel method,which uses two kinds of the grids and calculates iteratively on the grids.It is fast convergent,and implemented in the MPI parallel enviroment.Two strategies are studied to implement the Parareal algorithm.They are called master-slave and distributed algorithm.The numerical results show that the distributed algorithm owns a better speed-up ratio than the master slave algorithm.The neutron kinetics equations discretized in one time step are also solved in OpenMP parallel way.Combined this technique and Parareal algorithm,a MPI-OpenMP hybrid way is proposed,and can improve the speed-up ratio further.Then the coupled three dimensional neutronics/thermal-hydraulics code is studied for the fast reactor core transient analysis.The thermal property and the heat transfer model of the liquid sodium is added into the sub-channel analysis code COBRA.For the special reactivity feedbacks of fast reactors,the assembly's cross-sections library is established before the core calculation and the mesh size is also adjusted during calculation because of the expansion feedback.The cross-section library contains the enough range of the fuel and coolant temperature which might appear in the transient process.The coupled code HEXCOB is developped by coupling three dimensional neutronics and thermal-hydraulics sub-channal code COBRA.The adaptive BDF and the Parareal algorithm are also implemented in the code HEXCOB.The numerical results test and verify the code.Finally,the transient characteristic of TWR is studied by the means of the coupled code HEXCOB.The transient processes such as uncontrolled control rod withdrawal under full power state or 0.1% full power state,the control rods with different value dropping into the core respectively,and so on,are then simulated and analysed.The maximum temperatures of the fuel and the clad are concerned during the transient process.The simulations show that the safety criterion is satisfied under the condition of the protected measure.