To increase the efficiency and reliability of the thermodynamics analysis of the hydraulic system, the method based on pseudo-bond graph is introduced. According to the working mechanism of hydraulic components, they ...To increase the efficiency and reliability of the thermodynamics analysis of the hydraulic system, the method based on pseudo-bond graph is introduced. According to the working mechanism of hydraulic components, they can be separated into two categories: capacitive components and resistive components. Then, the thermal-hydraulic pseudo-bond graphs of capacitive C element and resistance R element were developed, based on the conservation of mass and energy. Subsequently, the connection rule for the pseudo-bond graph elements and the method to construct the complete thermal-hydraulic system model were proposed. On the basis of heat transfer analysis of a typical hydraulic circuit containing a piston pump, the lumped parameter mathematical model of the system was given. The good agreement between the simulation results and experimental data demonstrates the validity of the modeling method.展开更多
This paper introduces a powerful design and analysis tool named SIMCAT, that is developed to support applications to license a CANDU nuclear reactor, refurbish projects, and support the existing CANDU stations. It con...This paper introduces a powerful design and analysis tool named SIMCAT, that is developed to support applications to license a CANDU nuclear reactor, refurbish projects, and support the existing CANDU stations. It consists of the CATHENA （Canadian Algorithm for Thermo-Hydraulic Network Analysis）, the control logics from C6SIM （CANDU 6 Analytical Simulator）, and a communication protocol, PVM （parallel virtual machine）. This is the first time that CATHENA has been successfully coupled directly with a program written in another language. The independence of CATHENA and the C6SIM controllers allows the development of both CATHENA and C6SIM controller to proceed independently.展开更多
In case of accident at a nuclear power plant, water sources may not be available for a long period of time and the core heats up due to the residual power. Any attempt to inject water during core degradation can lead ...In case of accident at a nuclear power plant, water sources may not be available for a long period of time and the core heats up due to the residual power. Any attempt to inject water during core degradation can lead to quenching and further fragmentation of core material. The fragmentation of fuel rods and melting of reactor core materials may result in the formation of a ＂debris bed＂. The typical particle size in a debris bed might reach few millimeters （characteristic length-scale： 1-5 mm）. The two-phase flow model for reflood of the degraded core is briefly introduced in this paper. It is implemented into the ICARE-CATHARE code, developed by IRSN （Institut de radioprotection et de surete nucleaire）, to study severe accident scenarios in pressurized water reactors. Currently, the French IRSN sets up two experimental facilities to study debris bed reflooding, PEARL and PRELUDE, and validate safety models. The PRELUDE program studies the complex two phase flow （water/steam）, in a porous medium （diameter 180 mm, height 200 mm）, initially heated to a high temperature （400℃ or 700℃）. On the basis of the experimental results, thermal hydraulic features at the quench front have been analyzed. The two-phase flow model shows a good agreement with PRELUDE experimental results.展开更多
Boiling heat transfer condition has significance for pool-type research reactors cooled by natural circulation.It has important effect on the fuel element safety of reactor.On the basis of heat transfer characteristic...Boiling heat transfer condition has significance for pool-type research reactors cooled by natural circulation.It has important effect on the fuel element safety of reactor.On the basis of heat transfer characteristics of the Xi'an pulsed reactor(XAPR),fuel conduction,single-phase convection and boiling heat transfer,and void fraction models of the core are constructed.To validate the correctness of the physical models presented in the paper,numerical calculation based on a subchannel analysis method of XAPR is carried out,and the temperature fields are measured in some reactor coolant channels.The comparison between the calculated and experimental results verifies the effectiveness of the models.These physical models are used to calculate the thermal-hydraulic parameters of XAPR at the rated power(for XAPR the rated power is 2.0 MW in steady-state operation).The results indicate that subcooled boiling occurs in the XAPR core but it exhibits a subcooling degree which is considerably higher than that of saturation boiling.Subcooled boiling improves the efficiency of heat transfer between the fuel element surface and coolant,as well as effectively protects fuel elements.This research is also a beneficial reference in thermal-hydraulic analysis for other natural circulation reactors.展开更多
基金Project(51175518)supported by the National Natural Science Foundation of China
文摘To increase the efficiency and reliability of the thermodynamics analysis of the hydraulic system, the method based on pseudo-bond graph is introduced. According to the working mechanism of hydraulic components, they can be separated into two categories: capacitive components and resistive components. Then, the thermal-hydraulic pseudo-bond graphs of capacitive C element and resistance R element were developed, based on the conservation of mass and energy. Subsequently, the connection rule for the pseudo-bond graph elements and the method to construct the complete thermal-hydraulic system model were proposed. On the basis of heat transfer analysis of a typical hydraulic circuit containing a piston pump, the lumped parameter mathematical model of the system was given. The good agreement between the simulation results and experimental data demonstrates the validity of the modeling method.
文摘This paper introduces a powerful design and analysis tool named SIMCAT, that is developed to support applications to license a CANDU nuclear reactor, refurbish projects, and support the existing CANDU stations. It consists of the CATHENA （Canadian Algorithm for Thermo-Hydraulic Network Analysis）, the control logics from C6SIM （CANDU 6 Analytical Simulator）, and a communication protocol, PVM （parallel virtual machine）. This is the first time that CATHENA has been successfully coupled directly with a program written in another language. The independence of CATHENA and the C6SIM controllers allows the development of both CATHENA and C6SIM controller to proceed independently.
文摘In case of accident at a nuclear power plant, water sources may not be available for a long period of time and the core heats up due to the residual power. Any attempt to inject water during core degradation can lead to quenching and further fragmentation of core material. The fragmentation of fuel rods and melting of reactor core materials may result in the formation of a ＂debris bed＂. The typical particle size in a debris bed might reach few millimeters （characteristic length-scale： 1-5 mm）. The two-phase flow model for reflood of the degraded core is briefly introduced in this paper. It is implemented into the ICARE-CATHARE code, developed by IRSN （Institut de radioprotection et de surete nucleaire）, to study severe accident scenarios in pressurized water reactors. Currently, the French IRSN sets up two experimental facilities to study debris bed reflooding, PEARL and PRELUDE, and validate safety models. The PRELUDE program studies the complex two phase flow （water/steam）, in a porous medium （diameter 180 mm, height 200 mm）, initially heated to a high temperature （400℃ or 700℃）. On the basis of the experimental results, thermal hydraulic features at the quench front have been analyzed. The two-phase flow model shows a good agreement with PRELUDE experimental results.
文摘Boiling heat transfer condition has significance for pool-type research reactors cooled by natural circulation.It has important effect on the fuel element safety of reactor.On the basis of heat transfer characteristics of the Xi'an pulsed reactor(XAPR),fuel conduction,single-phase convection and boiling heat transfer,and void fraction models of the core are constructed.To validate the correctness of the physical models presented in the paper,numerical calculation based on a subchannel analysis method of XAPR is carried out,and the temperature fields are measured in some reactor coolant channels.The comparison between the calculated and experimental results verifies the effectiveness of the models.These physical models are used to calculate the thermal-hydraulic parameters of XAPR at the rated power(for XAPR the rated power is 2.0 MW in steady-state operation).The results indicate that subcooled boiling occurs in the XAPR core but it exhibits a subcooling degree which is considerably higher than that of saturation boiling.Subcooled boiling improves the efficiency of heat transfer between the fuel element surface and coolant,as well as effectively protects fuel elements.This research is also a beneficial reference in thermal-hydraulic analysis for other natural circulation reactors.
