Thermal-hydraulic Research On Loss Of Coolant Accidents Of Large Advanced Passive PWR

In nuclear power field,the passive technology has been extensively researched and widely applied in engineering in recent years.For passive pressurized water reactor(PWR),its special safety system operation characteristics are different from those of the conventional PWRs when loss of coolant accident(LOCA)occurs.Therefore,thermal-hydraulic research on LOCA of large advanced passive PWR is one of the important tasks to master the characteristics of its passive systems.This thesis mainly focuses on the following three aspects:(1)A simulation model is built by the mechanistic code of the IFBVHC experimental apparatus in SJTU.The objective is to evaluate the capability of the code simulating low pressure and natural circulation systems by comparing the calculation results with the experimental data.According to the design characteristics of large advanced passive PWR,the thermal-hydraulic analysis model of large advanced passive PWR is developed by the mechanistic code,including reactor cooling system(RCS),passive core cooling system(PXS),auxiliary system and simplified secondary system.Systematic steady state debugging is conducted through comparing with design parameters to verify the detailed analysis model of large advanced passive PWR.(2)The detailed analysis model is used to analyze a series of small-break LOCAs(SBLOCA)and a large-break LOCA(LBLOCA).SBLOCA spectrum is including inadvertent ADS actuation,50.8mm cold leg break,double-ended rupture of direct vessel injection line(DEDVI)and 254 mm cold leg break,while LBLOCA is analyzed the double-ended cold leg guillotine(DECLG)accident.Different break sizes and locations are selected and the results of these calculations can cover the design basis accidents of LOCA.Results demonstrate that PXS has the capability of depressurizing the RCS,injecting borated water into the reactor pressure vessel,cooling the core in time and effectively remove the decay heat.Further,in order to assess the impact of different containment back pressure and single failure assumptions on overall system response,sensitivity study is performed.(3)Sequence of events for an SGTR demonstrates PXS can effectively depressurize and cool down the RCS,remove the decay heat and automatically terminate the break flow before the ruptured SG being overfilled.Finally,the analyses of different break types,locations and number of the ruptured tubes are established by sensitivity studies to investigate their influence on the thermal-hydraulic response.Research shows that simultaneously rupture of five tubes may result in the countercurrent flow from the secondary side of the ruptured SG to the RCS.The thermal-hydraulic research on LOCA of large advanced passive PWR provides certain reference and support to understand the engineering design characteristics of PXS and its thermal-hydraulic phenomena.