Research On Analysis Method Of Kinetics For Channel-Type Molten Salt Reactor With Coupled Neutronics And Thermal-Hydraulics

In order to pave a way for nuclear energy getting out the development dilemma in terms of safety,spent assembly reprocessing and fuel source shortage,widespread research on design and analysis of innovative reactors has been performed all over the world.As one of the six candidates for Generation IV reactors,Molten Salt Reactor(MSR)adopts the liquid fuel,and is characterized by water-free cooling,inherent safety,physically Nuclear Non-Proliferation,feasibility of employing Thorium-Uranium fuel cycle,ease of small modular design,etc.Nonetheless,the criticality and dynamics characteristics of MSR are substantially different from that of a traditional solid-fueled reactor because of its unique features in the aspects of fuel flow,self-cooling,etc.By comparison with the traditional reactor types,the coupling effect of multiphysics field in MSR including neutron flux,temperature,flow rate,stress and nuclide concentration is more noteworthy.The bases for multi-physics coupling analysis on an MSR are neutronics and Thermal-Hydraulics(TH).Therefore,research on analysis method of kinetics with coupled neutronics and TH exhibits an important academic signification and engineering value.In this work,an implicit strategy is implemented during the process of neutronics and TH coupling for the channel-type MSR.The ‘assembly-core' two-step method based on the deterministic theory is employed in the neutronics calculation to satisfy the need of iterative neutron flux calculation from dynamic analysis.The feedback model based on the few-group parameters of assemblies is the key to the neutronics and TH coupling calculation.Suitable equivalent homogenization models are developed for different types of assemblies,and the parameterization process of the macroscopic cross-sections at a series of discrete condition points is performed by mean of the least square fitting(LSF)method.For the neutronics aspect,considering the influence of fuel salt flow,the fourth order standard nodal expansion method(NEM)is implemented to obtain the neutron flux,and the time-dependent diffusion equations are discretized and solved by an unconditionally stable fully-implicit backward difference scheme and the exponential transformation method(ETM).The above-mentioned methods can effectively improve the calculation accuracy and save the calculation cost.Taking into account the special flow and heat transfer features of channel-type MSR,the parallel multi-channel model and the single channel heat transfer model are adopted for TH calculation.Under the steady state,the pressure-flow equations are decoupled based on the ‘prediction-correction' theory,after which the distributions of flow rate,pressure and temperature are obtained.Simultaneously,the distribution of graphite temperature is obtained analytically according to the convective heat transfer relationship between fuel salt and graphite;Under the transient condition,the system of nonlinear equations including pressure and flow rate parameters are solved iteratively by a Quasi-Newton method,and the effective heat transfer coefficient method is introduced to solve the heat conduction equation of graphite.Based on above-mentioned models,a dynamic analysis code named TMSR3 D is developed for the channel-type MSR.Under the un-flow condition,the verification process of the steady-state and transient diffusion benchmark problems for solid-fueled reactors proves the availability of the developed code in criticality calculation and neutron kinetics calculation;the verification and validation(V&V)process based on the operation data of the Molten Salt Reactor Experiment(MSRE)indicates that the few-group parameter model,the flow model of delayed neutron precursors,the multichannel TH model and the neutronics-TH coupling model implemented in TMSR3 D are correct,and the developed code can provide a reliable description of the dynamics behavior for the channel-type MSR;The further probe on MSRE also proves the inherent safety of its core design.Finally,steady-state and transient analyses for a 2 MWth thorium-based MSR(TMSR-LF)are carried out based on the TMSR3 D code.The steady-state results indicate that under the rated condition,the key parameters including the highest core temperature and the effective delayed neutron fraction are all lower than the design limited values.The transient results indicate that the dynamic response of the core to the variation of fuel flow rate in primary loop is not only related to the temperature feedback effect,but also is defined by the direct reactivity change caused by the flow variation.During the transient processes of loss of heat sink and overcooling the inlet fuel temperature,the reactor can achieve the steady state or automatically shut-down with safety from transients without the action of protection system.During the accident of reactivity insertion,the dynamic response is determined by the ratio between the reactivity inserted and the effective delayed neutron fraction(?/ ?),temperature feedback effect,and the cooling capacity of the core.The power response becomes more severe when ?/? increases.Within certain limits,the temperature feedback effect is more significant and the core takes less time to achieve the steady state as the initial power or the inserted reactivity increases.In general,research on aspects of the equivalent homogenization parameter model,kinetics model based on nodal expansion method and exponential transformation and the TH model based on parallel multi-channel approximation is performed in this work,and the dynamic analysis code TMSR3 D with coupled neutronics-TH for the channeltype MSR is developed.V&V procedure for this code is carried out based on corresponding benchmark problems and the experimental data of MSRE,and steadystate and transient analyses are conducted for the concept design of the first thoriumbased MSR(TMSR-LF)to be built in China,for which the neutronics and TH coupling mechanism and the dynamic characteristics are probed.The numerical model and analysis method proposed in this work can be applied to the design and safety analysis for a series of MSRs.