Fuel Management Method For Research Reactor Based On Coupled Monte Carlo Particle Transport And Depletion Calculation

Research reactor is different from nuclear power plant.There are variety of irradiation samples in research reactor core.And the change frequency of startup and shutdown in research reactor is much higher than that of nuclear power plant.Especially it would be demanded to restart research reactor in Iodine well.So the calculation method and software can not totally meet the demands of complicated research reactor analysis.A new fuel management system for this kind of research reactor should be developed,which can use to track depletion of fuel assemblies and protect it from break because of beyond time of depletion.A new fuel management method is developed based on coupling of neutron transport,depletion calculation,criticality search and multiple temperature cross sections.And a Monte Carlo fuel management code named MCBMPI(Monte Carlo Burnup code based on Message Passing Interface)is developed.Benchmarks and experiments are conducted to verify this code.The results show that the code agrees well with both other benchmark results and experiment results.And it can play an important role in reactor operation.The contents of this research include the following three parts.Firstly,a Monte Carlo fuel management code named MCBMPI is developed And this code applies some advanced algorithm and methods,such as predictor-corrector coupling scheme,ultra-fine energy group method for depletion cross section calculation,adaptive criticality search algorithm and so on.And some calculation cases are used to verify these new algorithms and methods.Secondly,this code is verified by calculating some benchmark problems.And the eigenvalues and nuclides concentrations are compared with other codes.Then two experiments on SPRR-300 are conducted to compare the calculation values by MCBMPI.One is equilibrium Xenon and Iodine well experiment,and the other one is burnup reactivity experiment.Finally,a graphic user interface(GUI)for reactor core refueling is developed.The GUI can meet all the functions of fuel assembly change and moving in reactor core.In conclusion,according to above research,we find that MCBMPI can do well in benchmark problem and its speedup performance is even better than MCNP5MPI.The ultra-fine energy group method for depletion cross section calculation and adaptive criticality search algorithm can make Monte Carlo fuel management calculation faster.A case calculation of 50000 groups is 27 times faster than calculation including reaction rate tally for 234 nuclides.And the adaptive algorithm is 5.6 times faster than routine interpolation method for criticality search of SPRR-300 first core loading.The difference of calculation and experiment for burnup reactivity on SPRR-300 is 8%,which shows great agreement between calculation and experiment.