Study On The Fuel Self-sustainability Of Thorium Molten Salt Reactor In Online Plutonium Addition Mode

Molten salt reactor is the only liquid fuel reactor,which was selected as one of the six types of reactors in the Generation IV advanced nuclear energy system.It has many remarkable characteristics:inherent safety,proliferation resistance,neutron economy and excellent fuel cycle.Molten salt reactors are considered to be the best type of utilization of thorium resources.The thorium-uranium cycle study of molten salt reactors plays an important role in solving the long-term supply of nuclear fuel and reducing the amount of spent fuel.At present,the research on thorium and uranium cycles in molten salt reactors mainly focuses on two processing modes:?1?Online reprocessing to separate neutron poisons and recover fuel and carrier salts.In this mode,the best breeding performance of thorium and uranium fuel can be achieved.However,the ²³²Th-²³³U breeding capability of that highly relies on the online reprocessing technology,which requires long time research and development?R&D?and is difficult to realize in a short time.?2?One-through mode.The thorium fuel mixed with low-enriched uranium or reactor-grade plutonium is usually used to start the reactor,and online replenishment of low-enriched uranium or plutonium fuel to maintain the balance of reactivity.Spent fuel is unloaded once at the end of core lifetime,and it can be geologically disposed or recovered by batch processed.The technology of this model is relatively mature,and the fuel for starting the reactor can be obtained.However,the fuel conversion ratio of that is relatively low,which is equivalent to the one-through of the pressurized water reactor.Therefore,it is only suitable for the initial utilization stage of the ²³²Th-²³³U fuel cycle of the molten salt reactor.This article proposes a sustainable fuel cycle program under the online plutonium addition mode:Using ²³³U/Th to start the reactor,the reactivity supplement and the self-sustainment of ²³³U in the reactor are achieved by online feeding plutonium.Uranium,thorium and carrier salts are recovered by fluorination and evaporation,vacuum distillation and electrochemical methods at the end of core lifetime;the recovered fuel can start the next self-sustained molten salt reactor in the same way.This model replaces online reprocessing technology with plutonium addition and offline batch processing,while achieving the self-sustainability of Th/U fuel and the plutonium burning capability from the pressurized water reactors.The key problems solved in this work are:?1?Feasibility analysis of achieving self-sustainment of thorium and uranium under this mode;?2?The dependence relationship between thorium uranium self-sustaining and the amount of added plutonium;?3?The burning efficiency of plutonium itself.The research content of this article is mainly divided into two parts,which are the fuel sustainability of two types of reactors,graphite-moderated molten salt reactor and molten salt fast reactor under online plutonium addition mode.The research methods of the two reactor types are similar,which start with the single cell model and analyze the general rules of ²³³U self-sustaining and plutonium utilization,and then study the core model according to the analysis results of the single cell model.Finally,an appropriate core model is established for detailed fuel utilization analysis.In both models,²³³U/Th is used as the fuel for starting the reactor,a bubbling system is used to remove the insoluble and gas fission products in the depletion process,and the reactor-grade plutonium is added to maintain the core criticality.FLiBe is selected as the carrier salt in the graphite moderation model,and three carrier salts,which are FLiBe,LiF and NaCl respectively,are compared and analyzed in the fast spectrum model.In this work,a developed Molten-salt Reactor Refueling and Reprocessing System analysis code?MSR-RRS?based on SCALE 6.1 is adopted for the burning calculation.In the calculation of graphite-moderated single cell,by changing the volume ratio of molten salt?representing the neutron energy spectrum?and neutron loss rate?representing the absorption of structural materials and leakage,etc.?,the self-sustaining performance of ²³³U and the incineration efficiency of plutonium are analyzed.The analysis results indicate that the self-sustaining of ²³³U can be achieved in most of the range of molten salt volume ratio of about 10%-85%.and the breeding efficiency of ²³³U is the best when the molten salt volume ratio is about 43%.On the contrary,the two regions with lower and higher molten salt volume have less dependence on plutonium and relatively less added amount,while the middle region has higher dependence on plutonium.In addition,the burning rate of plutonium is higher in the thermal spectrum area,and the burning efficiency is lower in the fast spectrum area.In general,the thermal spectrum area is more conducive to the self-sustaining of thorium and uranium and the burning of plutonium.Therefore,based on the single cell analysis results and the temperature reactivity coefficient analysis under the core model,the core model which is finally selected as follows:The volume ratio of molten salt is 20%,the cell margin is 10cm,and the thermal power is 250 MW.In the 10 years fuel consumption,the breeding ratio of ²³³U is 1.22,that is,the end of life²³³U is increased by 22%compared with the beginning of life;the cumulative addition of plutonium is 639.9 kg,which is 2.2 times the initial uranium loading,and the consumption rate of plutonium is 54.4%;at the same time,32 kg of MA was newly produced,accounting for 0.18%of the total heavy metal mass.During the decade of burnup,the temperature reactivity coefficient is always negative and decreases with time.In the calculation of fast spectrum single cell,the model is pure molten salt fuel,which only changes the neutron loss rate,and the change range is 0%-15%;The self-sustaining time of ²³³U under the two ways of adding ²³³U/Th and Pu/Th online is studied.The results show that under the conditions of online addition of ²³³U/Th,²³³U self-sustainment can be established under FLiBe,LiF,and NaCl carrier salts,but the neutron loss rate is low.In the FLiBe salt,the self-sustaining can not be mantained when the neutron loss rate is 5%.However,when Pu/Th is added online,the self-sustainment time of ²³³U can be significantly improved.By contrast,LiF carrier salt can still achieve 45 years of self-sustaining time when the neutron loss rate is 7.5%,but in the case of ²³³U/Th feeding,it can no longer be self-sustaining;under the same neutron loss rate addition,the self-sustaining time of FLiBe carrier salt is shorter,but the amount of plutonium added is higher.When the neutron loss rate is 7.5%,the self-sustaining time of the three carrier salts FLiBe,LiF and NaCl are 25 years,45 years and 126 years and the amount of plutonium added are 4.1 kg/MW,2.6 kg/MW and 2.1kg/MW,respectively.Therefore,the LiF and NaCl carrier salts are more suitable for fast-spectrum solutions.Based on the analysis results of the single cell,the relationship between the neutron loss rate and conversion ratio with the core size is further analyzed,it is found that the neutron loss rate of NaCl carrier salt is much greater than LiF carrier salt under the condition of same core size,and the conversion ratio is much smaller than LiF carrier salt.Therefore,LiF is finally selected as the carrier salt,and a simple core model with a height and diameter of 280 cm is established,whose core thermal power is 1200 MW.It is concluded from the burnup analysis that the core can achieve self-sustainment of ²³³U between 10 and 32 years.Compared with the graphite-moderated reactor,the relative addition of plutonium in the fast reactor model and the relative production of MA is less.Specially,in the 32nd year,the cumulative addition of plutonium is 2.7 tons,accounting for 36%of the initial uranium loading,the output of MA is 110 kg,accounting for 0.16%of the total heavy metal mass,but the consumption rate of plutonium itself is very low,only 33.6%.In this work,an online plutonium addition solution is adopted to solve the problem that the fuel of Thorium Molten salt Reactor is difficult to self-sustainment under the condition of Off-line reprocessing,and at the same time analyzed the corresponding problems of plutonium dependence and incineration efficiency,which has some reference significance.