Development Of System Analysis Code With Three-dimensional Hot Pool Model For Pool-type Fast Breeder Reactor

Aiming at developing system analysis code independently, a System Analysis Code for Pool-type Fast Reactor in China (SAC-CFR), with neutron kinetics model, core and hot pool model, intermediate heat exchanger model, primary and intermediate heat transport system, steam generation system, was developed based on Compaq Visual Fortran to analyze the thermal-hydraulic characteristic of fast reactor under steady state operation. Then the transient calculation function of SAC-CFR was developed with further development of component transient model, plant control and protection system model, transient calculation logic and stability criteria.The newly developed SAC-CFR is characterized by three aspects. Firstly, the conventional point reactor neutron kinetics model was chosen to calculate the time dependent fission power with prompt jump approximation (PJA) method and piecewise polynomial approximation method. Secondly, the hot pool was analyzed with three-dimensional model developed based on SIMPLE algorithm on stagger grid under Cartesian coordinates and cylindrical coordinates. Thirdly, The three-dimensional hot pool analysis model was updated after incorporating porous medium model of penetration components and complex geometry component. Penetration components included intermediate heat exchanger (IHX), primary pump, decay heat exchanger (DHX), and radial shielding. Then the updated three-dimensional model was embedded into SAC-CFR.Then the newly developed code was used to analyze system thermal-hydraulic characteristic of China Experimental Fast Reactor (CEFR) under steady state operation of different power level. The good agreements between the simulation results and data in CEFR safety analysis report (SAR) showed that the present model is effective. The transient started from turbine trip test at45%thermal output in the Monju plant was analyzed with SAC-CFR. The simulation results, such as mass flow rate in primary loop, coolant temperature at inlet and outlet of main vessel, coolant temperature at outlet of intermediate heat exchanger (IHX), were compared with the data tested in reactor. A good agreement between the calculated results and the test data was obtained.Finally, the system analysis code with three-dimensional hot pool model was used to analyze the thermal-hydraulic behavior in upper plenum of "MONJU" FBR during reactor scram transient, and the flow field in CEFR hot pool under steady-state operation condition and after scram. Some thermal stratification characteristics in Monju was founded. The agreement between the computational flow field and hot pool geometry of CEFR showed the effectiveness of porous medium model. Also, some characteristic of flow field was founded.As the extension of SAC-CFR development, an analysis code to perform thermal-hydraulic calculation in Passive Direct Reactor Auxiliary Cooling System (PDRACS) was developed through the coupling of thermal-hydraulic response in direct heat exchanger (DHX), air heat exchanger (AHX) and circuits of PDRACS. After steady-state initialization, a transient starting from air damper opening at the instant of reactor scram was simulated. Key parameters, such as mass flow rate in intermediate loop under natural circulation, heat removed by PDRACS, air temperature at AHX outlet, coolant temperature at both side of DHX, are calculated, which preliminary showed the effectiveness of the code. And the newly developed code is now ready to be incorporated into SAC-CFR to perform the interactive response between the PDRACS and the whole system after scram.