Analysis Of Coupling Characteristics Between Thermal-Hydraulics And Neutronics For Light-Water Reactor

Various physical phenomena can be encountered during the operation of a nuclear reactor,which probably include the generation and transportation of neutron,the fluid flow and heat transfer,chemical reaction between different materials and the change of mechanical stress and so on.Commonly,specific codes are required to analyze each and every of above-mentioned physical phenomena.As matter of fact,those physical phenomena are coupled and interacted with each other.Among all the interactions,the coupling between the transportation of neutron and the fluid flow and heat transfer is always important.Therefore,the analysis of coupling characteristics between thermal-hydraulics and neutronics is essential for the safety design for reactor core.In the light of the fact that the coupling of deterministic reactor physics code based on Boltzmann neutron transportation equation and quasi-three-dimensional thermal-hydraulics code has been widely incorporated and the calculation accuracy is significantly affected by different structures or geometric models,a universal high-precision computing model is urgently needed to overcome disadvantages from the complex reactor core geometrical structures and the heterogeneity of the materials etc.The coupling between high-fidelity Monte Carlo method and quasi-three-dimensional thermal-hydraulics code may be a feasible option.A package for the purpose of analyzing coupling characteristics of thermal-hydraulic and neutronics is developed on the basis of Monte Carlo method in this paper.Continuous-energy point-wise cross section database is manufactured by using NJOY code.An in-house script was developed based on Python code,which invokes sub-channel code and three-dimensional continuous-energy neutron transportation MCNP code in a so-called external manner.Through comparing the results of the developed codes package to the published data of SERPENT/RELAP5,the method proposed in this thesis is validated.Furthermore,further development of the sub-channel code is conducted,and a high-precision three-dimensional coupling system is developed.The thorough three-dimensional analysis of a fuel rod is realized by using the coupling between MCNP and sub-channel code.Moreover,different cases are studied by using the developed code package,which include the detailed analysis of fuel assembly of simplified boiling water reactor and pressurized water reactor,and the analysis of full size assembly for pressurized water reactor and the whole core.The results show that the distribution of the power,fuel temperature,coolant temperature and density in the boiling-water reactor are quite different from that in the pressurized water reactor because of the sharp gradient of moderator density.If the total power is reduced or the mass velocity is increased,the heterogeneity of such parameters can be reduced to some extent.It is found that the power density in a fuel rod tends to monotonically increase along radial direction via the high-precision coupling algorithm.It can also be concluded that such radial distribution of power density has a flattening effect for the uneven distribution of the temperature field across the rod.The different schemes of axial node have very little influence for the coupling analysis of pressurized water reactor,whereas the different fuel arrangement scheme with different enrichment impacts thermal-hydraulics parameters significantly.The coupling and feedback effects at the core-wide level make the flattening effect of the core radial power distribution extremely obvious.It can also be concluded that averaging treatment across fuel rod cross section is over-conservative and more realistic distributions of key parameters can be provided by using self-developed high-precision coupling algorithm.Furthermore,for reactor design and safety analysis,the results given by high-precision algorithm are more faithful as compared with those by averaging treatment.