Core Thermal Calculation Of Fast Reactor Refueling Process Based On Subchannel Program

In order to meet the energy demand of the national economy and achieve the goal of carbon neutrality and carbon peak,active and orderly development of nuclear energy is currently a national policy of China.In order to ensure the safety of nuclear energy,it is necessary to conduct a thermal safety analysis of the reactor core to meet the design criteria.The fast reactor sub-channel analysis program is an important technical mean to carry out the above analysis.At present,the PWR sub-channel program is relatively mature.Because the layout of the core fuel assembly and the coolant physical properties of the sodium-cooled fast reactor are quite different from that of the PWR,the sub-channel program of the sodium-cooled fast reactor requires special research and development.At home and abroad,a series of sub-channel analysis programs have been developed for liquid metal cooling fast reactor cores.These programs can meet the requirements of core thermal analysis to a certain extent,but there is a lack of verification by a large number of experiments and they are still in the perfect stage.In addition to some differences in mathematical models and solving methods for different sub-channel calculation models,there are also differences in the selection of empirical relations.Therefore,a lot of research on sub-channel programs is needed to improve the accuracy of the program.In response to the above problems,we independently developed SAC-SUB(System Analysis Code-Subchannel),a sub-channel analysis program for the thermal-hydraulic calculation of the sodium-cooled fast reactor.In order to verify the applicability and accuracy of the self-developed fast reactor sub-channel analysis program,the calculation and analysis of the FFM(2A)(Fuel Failure Mockup)experimental benchmark questions of the Oak Ridge National Laboratory in the United States were carried out.The temperature of the coolant at the outlet of the sub-channel under the three working conditions of the FFM experiment is analyzed,and the temperature of the sodium coolant at the outlet of each sub-channel is calculated by SAC-SUB.By comparing with the experimental data.the maximum deviation of the coolant outlet temperature is within 5%,which verifying the applicability and accuracy of the program.The verified SAC-SUB program was used to carry out the thermal-hydraulic calculation of the different cycles of the China Experimental Fast Reactor(CEFR)core from uranium dioxide fuel to MOX fuel.Key thermal-hydraulic parameters under first cycle,second cycle.third cycle,the fourth cycle and equilibrium state were obtained,such as average outlet temperature of the hottest fuel assembly sub-channel in each flow zone,the highest sodium temperature of the sub-channel,the temperature rise of the cladding,the temperature rise of the membrane,the temperature rise of the fuel,and the sub-channels average temperature rise at the outlet,the highest sodium temperature rise in the sub-channels,etc.The following results were obtained by being compared with the design value:the maximum deviation of the average temperature of the hottest fuel assembly sub-channel outlet is 1.56%;the maximum deviation of the channel maximum sodium temperature is 1.18%;and the maximum sodium temperature rise of the sub-channel largest deviation is 1.181%,which meets the requirement that the deviation from the design value is not more than 2%;the maximum deviation of the cladding temperature rise is 7%,and the maximum deviation of the membrane temperature rise is 9.58%,and the maximum deviation of the fuel temperature rise is 1.81%,which meets the requirement that the absolute deviation from the design value is less than 10%.The fast reactor sub-channel calculation program SAC-SUB can be used for the thermal-hydraulic calculation of the sub-channel of the sodium-cooled fast reactor,which can provide a certain theoretical analysis tool for the future design and research of the sodium-cooled fast reactor and the development of the sub-channel program of the sodium-cooled fast reactor.At the same time,the thermal-hydraulic calculation results of the refueling process of the China Experimental Fast Reactor can provide certain reference value for the future CEFR MOX fuel core and its transition process.