Research On The Dynamics Method And Transuranium Transmutation For The Molten Salt Reactor Moderated By Zirconium Hydride Rods

Considering its operation temperature,moderating ratio and price,graphite is generally used as moderator in molten salt reactor(MSR).This reactor core is composed of a large number of rectangular or hexagonal graphite cells.However,graphite has poor moderating ability and needs to be replaced regularly after irradiation.Zirconium hydride(ZrH)is considered as an alternative moderator in MSR due to its excellent slowing-down ability,good thermal stability and high radiation resistance.In a ZrH moderated MSR(ZrH-MSR),thousands of ZrH moderator rods surrounded by fuel salt are uniformly fixed in the active core.Due to no solid boundary between adjacent fuel lattice cells in a ZrH-MSR core,an evident axial and transverse fuel salt flow occurs between adjacent fuel channels,which would also lead to the corresponding mixing effect of the delayed neutron precursors(DNPs).In addition,the flowing molten salt in a ZrH-MSR core not only acts as fuel to produce energy in the reactor,but also as coolant to transfer the heat of molten salt and ZrH moderator rods from the core,which results in the form of a strong non-linear coupling between the fuel motion and neutron dynamics.In this work,a new three-dimensional neutronics and thermal-hydraulics coupled code was developed for a ZrH-MSR to simulate the power and temperature fields in complex geometries under axial and transverse fuel salt flow by considering the thermal coupling between the fuel salt and ZrH moderator rods.Then,by applying this coupled code,a series of ZrH-MSR core safety analyses were calculated and discussed.First,in order to deal with the inter channel mixing phenomenon of fuel salt and the thermal coupling between fuel salt and ZrH moderator rods in the ZrH-MSR,a subchannel thermal-hydraulics module(named SubTH hereinafter)was developed by the subchannel method.To verify the reliability of the SubTH code to deal with the inter channel mixing phenomenon,a rectangle assembly with 4 ZrH moderator rods,a hexagonal assembly with 7 ZrH moderator rods and a circular assembly with 7 ZrH moderator rods were selected.The Fluent calculation results were used as the verification benchmark.The accuracy and validity of the SubTH was verified by this indirect approach.Second,based on the Monte Carlo particle transport code MCNP5 and the subchannel thermal-hydraulic code SubTH,a steady-state coupled code named MCNP-SubTH was developed by an external coupling.The accuracy of this code was verified by each relatively independent module.A hexagonal fuel assembly was simulated,which further shows the validity of this code.Besides,by applying this code,the steady-state safety analysis of the 1800 MWth ZrH-MSR was carried out.The k_eff,neutron fluxes,core temperature distribution under various conditions were analyzed,which could provide valuable information for its further optimization.Then,a new neutronics code(named 3DN hereinafter)was developed and also considered the fuel salt convection term not required in the traditional multi-group neutron diffusion equations to reflect the inter fuel salt channel mixing effect.The coupled code(named 3DN-SubTH hereinafter)was established through exchanging the data between the 3DN and the SubTH.The verification results of benchmarks indicate that the 3DN-SubTH code can provide a reliable dynamics simulation for a ZrH-MSR.The 25 MWth ZrH-MSR proposed by Transatomic Power Corporation was selected to evaluate its steady-state and transient characteristics by applying this coupled code.The results show that the control rods fully withdrawn condition under the rated power has the highest fuel salt subchannel temperature(subchannel No.3,1025.53 K),corresponding to the hottest ZrH moderator rod centerline temperature(ZrH moderator rod No.3,1065.21 K).According to the limit for peak ZrH_1.66moderator rod temperature(1073.15 K),it can meet the requirement of ZrH moderator rod safety.Although the changes of the neutron fluxes and DNPs caused by the inner channel mixing effect are not particularly dramatic due to its extremely low power and inlet fuel salt flow velocity.However,it has a significant influence on its core temperature.The maximum temperatures of the hottest fuel salt subchannel and ZrH moderator rod centerline with axial and transverse fuel salt flow are 5.21 K and7.35 K lower than those without the transverse fuel salt flow,respectively.In addition,this reactor has an excellent negative temperature reactivity coefficient,which can ensure its safe shutdown under various transient conditions.Finally,a ZrH-MSR conceptual design aimed to TRU transmutation was proposed and analyzed from the fuel cycle aspect and the neutronics and thermal-hydraulics coupling aspect,respectively.According to the TRU solubility limit and neutron economy,the Li F fuel salt and SVF=0.5 is selected since it can meet the requirement of its TRU solubility limit during 50 years operation and can has the maximal TRU consumption rate of about 252.0 kg/(GWth·year)and the Support factor of 2.9.The total TRU radiotoxicity after discharge is about 63.9%smaller than that without transmutation.The total temperature feedback coefficient can remain negative during the whole operation time,which ensures the safety of the ZrH-MSR.In addition,by applying the MCNP-SubTH code,its steady-state characteristics were evaluated.The highest fuel salt subchannel temperature(subchannel No.3)is 1045.50K,corresponding to the hottest ZrH moderator rod centerline temperature(ZrH moderator rod No.31,1085.63 K).According to the limit for peak ZrH_1.6 moderator rod temperature(1100 K),it can meet the requirement of ZrH moderator rod safety.However,it is necessary to further flatten the core power distribution or redistribute the fuel salt flow distribution due to its small residual safety margin.