Magyar Tudományos Akadémia KFKI Atomenergia Kutatóintézet
1121 Budapest, Konkoly Thege Miklós út 29-33. * Postacím: 1525 Budapest 114, Pf. 49. * Telefon: +36 1 392 2222 * E-mail:

Reaktor Analízis Laboratórium

Traditionally, the activity of the Reactor Analysis Department covers the static and kinetic neutron physical calculations of Light Water Nuclear Reactors, moreover their coupling to thermal hydraulic models. Special attention is paid for the reactors of VVER-440, VVER-1000 power plants and for the VVRSM Research Reactor of KFKI-AEKI which are 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, the validation, the acceleration, the application of the developed algorithms and codes are the most important directions of the research and development in the department.

During an accident (for example LOCA) in a nuclear power plant, large amount of radio nuclides is being discharged into the hermetic compartments which, in case of leaking, can be released also to the adjoining rooms and finally to the environment. In the last years, the development of a complex code has been started for the coupled modeling of the thermal hydraulics and the different complicated processes of the activity transport. The first results are being compared to those obtained from the earlier but much more time consuming calculations.

The developed and validated codes enable to perform the following tasks of the applications:

Design of periodic reloading of the reactor cores

Calculation of the inventory of the radio nuclides, their transport inside of the hermetic compartment, furthermore the release to the other compartments

Safety analyses of the RIA and ATWS events for licensing of new fuel types and power upgrade

Pressure vessel fluence calculations

Three-dimensional kinetic neutronic calculations in simulators

Advisory activity as a background of the Hungarian nuclear authority

Calculation of the core design and the dynamic behavior of the VVRSM type Budapest Research Reactor

Radiation shielding calculations

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.

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, to the HPLWR (High Performance Light Water Reactor). The core design of HPLWR and the safety analysis of reactivity induced incidents were performed in AEKI. The results exceed the results from literature, as they contain the detailed tracing of the 3D phenomena. From 2009, a new project has been started aiming at the development of a multigroup nodal code which is applicable for the fast spectrum metal cooled reactors.

Further new important activities:

Development and application of the burnup credit methodology for the subcriticality analysis of spent fuel storage facilities, quantification of the uncertainties.

Generation IV activities: High Performance Light Water Reactor core design and safety analyses. Design of the safety systems. Development of multigroup nodal code for the core design and safety analysis of the fast spectrum reactors.

Hot channel methodology used for the RIA and ATWS safety analyses, multiphysics approach.

Uncertainty analysis of Reactivity Initiated Accidents calculated by coupled 3D kinetic 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 uncertainties 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.