Reactor Analysis Department
Traditionally, the activity of the Reactor Analysis Department covers the development, validation and application of static and kinetic neutron physical calculations tools for Light Water Nuclear Reactors, moreover their coupling to thermal hydraulic and thermal mechanical models. Special attention is paid to the reactors of VVER-440, VVER-1000 power plants and to the VVRSM Research Reactor of KFKI-AEKI which can be analyzed by the own developed codes: the KARATE code system for the core design, the KIKO3D dynamic code for the safety analyses. The enhancements of the calculation accuracy, its quantification, the acceleration, the application of the developed algorithms and codes are the most important directions of the research and development in the department. The developed and validated codes enable to perform the following investigations: • Safety analyses of the RIA and ATWS events for licensing of new cores, fuel types and power upgrade • Design of the reactor cores including periodic reloading • Calculation of the inventory of the radio nuclides, their transport inside of the hermetic compartment, furthermore their release to the other compartments • Pressure vessel and reactor internal fluence calculations • Real time three-dimensional kinetic neutronic calculations for simulators • Advisory activity as a background of the Hungarian Atomic Energy Agency • Calculation of the core design and the dynamic behavior of the VVRSM type Budapest Research Reactor • Radiation shielding calculations of the reactor and the storage and transport devices One of the major innovations of coming decades in nuclear engineering will be the development of Gen IV NPPs. Up to 2009, the department was exclusively related to one type of Gen IV, namely to the HPLWR (High Performance Light Water Reactor) cooled by water above superctitical pressure. The core design of HPLWR and the safety analysis of reactivity induced incidents were performed. The results exceed the results from literature, as they contain the detailed tracing of the 3D phenomena. From 2009, anew projects have been started aiming at the development of a multigroup nodal code which is applicable for the fast spectrum metal cooled reactors which is presently applicable also for the above tasks. Further new important, perspecitivic activities: • Development and application of the burnup credit methodology for the subcriticality analysis of spent fuel storage facilities, quantification of the uncertainties. • Hot channel methodology applied for the safety analyses of RIA and ATWS events by using the multiphysics approach . • Uncertainty analysis of Reactivity Initiated Accidents calculated by coupled 3D neutronic and system thermal hydraulic codes. • Investigation of future (“closed”) fuel cycles leading to lower amount on radioactive waste. Fuel utilization scenarios by reprocessing and using different reactor types. • Sensitivity and uncertainty analyses of hot channel calculations for RIA and ATWS events. • Evaluation of the uncertainties originating from the cross section data in the frame of the Working Party of Reactor Systems and in the Working Party of Nuclear Criticality Safety. • Multi-physics: thermal mechanics, reactor physics, thermal hydraulics; small scale effects impact on DNBR. In the last years, Paks NPP initiated several projects aiming at increased maximum allowed power, more economic fuel cycles, lower pressure vessel fluence, maneuvering regime. For supporting these efforts, several modifications of the VVER fuel construction were introduced. Increased average enrichment, modification of the lattice pitch, fuel length and diameter, profiled enrichment, application of burnable absorber, shielding the absorber assembly coupler part with Hf plates can be mentioned in this respect. In spite of the mentioned fuel modifications, utilizing the new possibilities in the more economic core designs, such local safety relevant limitations like maximum linear heat rate, sub-channel enthalpy, hoop stress, cladding fatigue are still challenged. Having regard to the above situation, continuous development and validation of the own developed codes is indispensable for the repeated quantification of the uncertainties and the margins of the calculated of the above mentioned limitations. Special uncertainty methodology is used for this purpose.