Applicability Assessment For SuperMC Code In Neutronic Analysis Of Advanced Sodium-cooled Reactors

An increasing trend towards and interest in aspiration of safe and peaceful nuclear energy is overwhelming across the world these days.Due to its ability of breeding fuel and transmutation of radioactive waste,fast neutron spectrum has become one of the popular research focuses in the world.As the core neutronics plays a key role in supporting the safety and dynamics of the reactor,it is necessary to conduct such simulations with high fidelity.Many researchers carry out core neutronics simulations for SFRs using both deterministic and Monte Carlo(MC)codes.On the basis of these calculations when compared with benchmarked experiments,it can be found that results of MC codes are commonly observed to be closer to the measurements(more accurate)as compared with that of deterministic ones.FDS Team(INEST,CAS,China)developed a highly efficient and general purpose code-the SuperMC-which is capable to model and simulate the complex nuclear systems.In order to validate and apply the SuperMC code/program in the area of designing and safety analysis of fast reactors,the high-fidelity core neutronics simulations including transport and burnup calculations were performed and analyzed based on two benchmarked Sodium-cooled Fast-spectrum Reactors(BFS-62-3A and BN-600).The study could essentially be divided in to two major parts:I.The benchmarking of SuperMC program for Sodium-cooled Fast Reactors' neutronic analysis with BFS-62-3A which is a benchmarked experiment conducted at BFS-2 facility,IPPE,Russia.This work incorporates three distinctive subjects including:a.Validation of SuperMC code for its neutron transport calculation capability.During the validation,various parameters were calculated-The spectral indices,for instance,gave C/E values 0.9927 and 0.9990 for F49/F25 and F28/F25.The average discrepancies for radial fission rate distributions in the fuel region and control rod worths were found to be 2.7%and less than 5%respectively.b.The testification of competence of indigenously developed point-wise nuclear data library,the HENDL/MC.The simulations agreed well with the experiment and this fraction of our study hence enabled the HENDL/MC library to be benchmarked.c.A detailed investigation on core's neutronic analysis of BFS-62-3A:including estimation of the impact of reflector density on fission rates,sodium void reactivity effect,sensitivity analysis,the effects of various data libraries on the criticality and fission rates,and code-to-code verification(using the available results of three Monte Carlo codes-SuperMC,MCNP,and Serpent-and a deterministic code DYN3D-MG).For reflector's density dependence of reaction rates in peripheral region of the assembly,for instance,a decrease of density by 5%was found to be in good agreement with the experiment.For code-to-code verification,control rod worth(CRW)and fission rates were calculated.For CRW,the average deviations are 8.19%and 4.97%for the Serpent and SuperMC respectively.For fission rates,the average discrepancies between SuperMC and other codes were about 1.8%and 1.6%for 239Pu and 235U,respectively.II.The neutronic analysis of BN-600 hybrid-core reactor using SuperMC program.The prototype reactor was reconstructed for criticality studies and fuel burnup analysis.The calculations,keeping the shim rods(SHRs)mid-core inserted and scram rods(SCRs)fully out/withdrawn,involved estimation of change in isotopic concentrations and prediction of burnup effect and burnup reactivity loss during reactor operation.The results so obtained were compared with the reference data available in IAEA's technical document(IAEA-TECDOC-1623)and other published articles.The comparison gave a good agreement of the SuperMC's results with available reference data.It is to conclude that SuperMC code is quite competent and capable to perform the neutronic analysis of fast spectrum sodium-cooled reactors.During the neutronics study of BFS reactor(described in "I" above),the sensitivity analysis renders an interesting fact that the reactivity of the core is substantially sensitive to fuel materials;including U-90%(present in HEZ),U-36%(old)(present in LEZ)and Pu-95%(present in PEZ);while the reactivity is comparatively less sensitive to UO2-36%(present only in LEZ with a few number of pellets).This thesis presents a work based on a strong motivation to help alleviate the boundless and comprehensive adoption of efficient,safe and espousal fast-reactor technology(with a specific emphasis on advanced sodium-cooled reactors).The work performed would be playing an integral role in furnishing input to almost all aspects related to design and safety evaluation,and operation of an SFR.The future work is intended to apply learning from the study and highly efficient Monte Carlo code for Lead-cooled fast reactor,the CLEAR-0 and other variants designed by INEST,CAS,China;and studying the subcritical systems such as Ukraine-based Accelerator-driven KIPT subcritical nuclear assembly.