Study On The Migration Behavior Of Bubbles In A Lead-Cooled Fast Reactor After Steam Generator Tube Leakage

It has been proved that Lead or Lead-alloy cooled fast reactors(LFRs)have the potential to be the most promising reactor which will be commercially realized firstly.That's because of its outstanding fuel breeding and waste transmutation performance.However,in the pool-type structure design of LFR,steam generator units(SGU)are directly placed inside the reactor vessel.Compact structure arrangement,complex thermal-hydraulic phenomena,corrosion and erosion of the structures by lead alloys along with large pressure differential between primary and secondary system raise safety concerns related to steam generator tube leakage and/or rupture(SGTL/R).When the accident happens,the cooling water of the secondary system enters the primary system under pressure,and directly contacts the high-temperature liquid lead-bismuth alloy.The cooling water boils quickly,generating an instantaneous pressure shock wave that will have a big damage to the surrounding structure.At the same time,a large number of bubbles,generated by the heat exchange between the two-phase-flow,threatens the local heat transfer of the heat exchanger and even the core under the entrainment of the liquid lead-bismuth alloy.Hence,it is very important to carry out numerical simulations in this area.This paper mainly studies the steam generator tube rupture/leakage accidents in the SNCLFR-100 small natural circulation lead-cooled fast reactor and M2LFR-1000 forced circulation lead-cooled fast reactor,which are independently designed by the University of Science and Technology of China.Researches mainly focus on the migration and accumulation behavior of bubbles generated under the action of liquid lead and cooling water in the primary system.The main work of the thesis includes:(1)Several commonly used multiphase flow models are compared with each other.For the leakage of small bubbles in the heat exchanger,the DPM model(Discrete Phase Model,DPM)is selected with detailed introduction.It is verified with abundant experimental data of terminal bubble rise velocity at low dynamic viscosity.Finally,the Tomiyama drag coefficient correlation is selected.On this basis,the study on the factors affecting the maximum size of entrained bubbles in liquid lead is carried out and compared with the simulation results.It is concluded that the faster liquid lead flows,the larger bubbles that can be entrained is,and the temperature effect is not obvious.(2)According to the steady-state model of small natural circulation lead-cooled fast reactor SNCLFR-100,the trajectory tracking calculation of the bubbles after leakage is carried out.The monitoring surfaces are set up in each part of the reactor in order to count the number of bubbles arriving at these surfaces.The migration depth of bubbles under different conditions(bubble sizes,leakage heights and pure/contaminated systems)is analyzed,along with the possible bubble accumulation position of the reactor.Finally,the flow rate of bubbles entering the core under leakage accident is calculated quantitatively.The accumulation probability and escape probability of bubbles at different positions in the core is obtained.In comparison.When the reactor is working normally,the small-sized bubbles produced by the leakage enter the core circulation and escape,and the large-sized bubbles directly float up and escape.The bubbles that cannot be entrained may accumulate above the partition and in front of the core entrance.(3)For the forced-cycle lead-cooled fast reactor M2LFR-1000,a quarter-reactor model is established.Its steady-state circumstance is calculated by commercial software ANSYS FLUENT and is proved correct after the comparison with the design value.On this basis,DPM model was loaded to carry out the trajectory tracking calculation of the bubbled after the leakage.Still,the migration depth of bubbles under different conditions(bubble sizes,leakage heights and pure/contaminated systems)is analyzed.The probability of the bubble entering the core,remaining in the core,re-entering the inlet of the heat exchanger and directly ascending is calculated under the corresponding conditions.The bubble flow rate that entered the core,remained in the core,and re-entered the inlet of the heat exchanger under the leakage accident was quantitatively calculated.It is concluded that liquid lead with low purity can significantly enhance the entrainment ability of small-sized bubbles.Finally,the bubble tracking calculation results of SNCLFR-100 and M2LFR are compared in order to analyze and discuss and the reasons for the differences.In summary,this paper carries out simulation studies on the trajectory of bubbles generated after steam generator tube leakage accident of the lead-cooled fast reactor heat exchanger,grasps the basic law of these bubbles migration in the reactor,and compares the migration results under different conditions.The accumulation probability and escape probability of bubbles in important positions in the reactor are obtained.It provide relevant data for the follow-up study of the physical,thermal,and chemical corrosion effects of bubbles on important positions in the reactor.It also provides support for further safety analysis of leakage/break accident.