Best Estimate Plus Uncertainty Study Of AP1000 Integral Test Facility SBLOCA

When China introduce the GEN-III+ passive pressure water reactor(PWR)technology and develops autonomous reactors,it also needs to carry out codes and experimental verification to ensure safety.The typical test facility APEX can provide a large amount of data.There are few studies on APEX using BEPU method for complete analysis,so it is necessary to make experimental verification and code quantitative evaluation of passive safety characteristics based on BEPU method.This thesis develops a new type of M-PIRT based on traditional PIRT,and uses two main codes for modeling and transient analysis of typical accident conditions for APEX facility.Finally,it combines experimental data to verify and quantify a variety of uncertainties.Firstly,the key step——PIRT in BEPU was fully studied,the development history of PIRT was summarized systematically,and the guiding role of PIRT for the research of key models in the reactor scaled facility design and code development,which is currently concerned,was summarized.The new M-PIRT was proposed by judging the importance of the phenomenon through the size of the ball in the three-dimensional coordinate,and many typical reactors’ M-PIRT were shown in this thesis,including the high-level phenomena in PIRT under SBLOCA of the AP1000.This thesis analyzed the reasons for the ranking of the high-level phenomena to provide a basis for the selection of input parameters for subsequent uncertainty analysis.Secondly,based on the investigation of domestic and international typical PWR facilities,SBLOCA conditions and RELAP5 learning,the steady state model of RELAP5 about APEX facility was established,and its transient was verified based on the 0.5inches break LOCA.Then another break size(2 inches)was set in the RELAP5 to compare the safety characteristics under different break sizes.In addition,in order to compare the simulation capabilities of two different codes under the same transient,the conversion process of RELAP5 and TRACE models was combined and the TRACE was used to model the APEX facility.After verifying the steady state,the 2 inches break LOCA was set for simulation.Compared with the same conditions of RELAP5 simulations and experimental results,analyzed the important thermal-hydraulic phenomena and passive safety characteristics in the SBLOCA,and found that the TRACE and RELAP5 can better simulate the PWR in SBLOCA,which proved the rationality of the model in this thesis.Finally,the uncertainty analysis was carried out to ensure the reliability of best estimate.After comparing the lowest core level,the random sampling statistical method,sensitivity analysis numerical method,and traditional error transfer analysis method were respectively selected to perform uncertainty analysis.The random sampling statistical method with the highest core level value was selected for further quantitative analysis.The uncertainty bands and sensitivity analysis of the key parameters were obtained,and it was found that the uncertainty of different input parameters has different effects on the uncertainty bands of different output parameters at different times.The Spearman rank correlation coefficient calculated by DAKOTA showed that the input parameters which have a large impact on minimum core level are decay heat and break area.Based on the APEX facility,two main thermal-hydraulic codes were used to simulate different SBLOCA conditions,make different comparisons,and analyze the transient calculation by three uncertainty methods.The analysis results can provide a reference for the safety performance evaluation of the GEN-III+ advanced reactors(such as CAP1400,HPR1000,etc.),the application of BEPU,and provide a basis for the verification of the thermal-hydraulic codes.