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Computing Nuclear Reactor Cells

#3267


Development of Calculation Methodology Used in Computing Nuclear-Physics Characteristics of Nuclear Reactor Cells and Assemblies Containing MOX Fuel, and in Unstructured Grids for Nuclear Safety of Reactor Facilities

Tech Area / Field

  • FIR-MOD/Modelling/Fission Reactors

Status
3 Approved without Funding

Registration date
22.04.2005

Leading Institute
Kurchatov Research Center, Russia, Moscow

Collaborators

  • CEA/DEN/SAC / Systems and Structures Modeling Department, France, Gif-sur-Yvette Cedex\nUniversity of Arizona / College of Engineering and Mines, USA, AZ, Tucson\nPennsylvania State University, USA, PA, University Park

Project summary

The project focuses upon development of assembly level computational tools related to utilization of (LWR) technologies, such as pressurized water reactors PWR and water-water energetic reactor VVER for plutonium disposition in spent reactor fuel. Heterogeneous assembly configuration consisting of a standard LWR assembly loaded with both UO2 and mixed oxide (MOX) fuel pins is under consideration in this study.

Crucial undertaking in reactor control and safety analysis is availability of suitable calculation routines for predicting neutron behavior and reactor performance under normal conditions and during unforeseen accidents. One of the most important component of the present project is developing computational methods and codes for assembly-level calculation with their consequent utilization in the global reactor analysis.

Analysis of benchmark tests has indicated that current available engineering reactor codes often need unjustified adjustments to predict a correct pin power distribution in the vicinity of the boundary between MOX and UO fuel cells. In addition, traditional reactor analysis methods fail to analyze some important problems in principle: for example the problem of 3-D fine mesh neutron flux distribution which is of primary importance in estimation of neutron flux at the core-reflector interface, and over finite-uniform axial regions due to control rod presence as well as in analysis of sensor element responses of Inherent Reactor Control System.

Another problem raised in LWR reactor is a proper treatment of anisotropic scattering effects, particularly, in closed grids, heterogeneous assemblies with different types of fuel rods, and in proximity of control rods. Taking into account anisotropic scattering is important for calculation of the 2-D cell and assembly-level characteristics, such as pin-power distribution at the boundary of different fuel cells and for computing axial neutron leakage from the assembly. To overcome the problem outlined above, we propose to elaborate and realize consequent approach to numerical methods currently implemented in reactor lattice codes

As a basis for this project, the development of numerical methods for the solution of the neutron transport equation in 2-D complicated geometry systems and 3-D geometry systems has to be done. Taking into account the strong progress of computer techniques, these methods have good perspectives for the creation of a new generation of engineering codes, i.e. such codes which have accuracy closed to accuracy of precision codes while requiring a much smaller computing time satisfying engineering code demands.

In order to be able to predict reactor behavior during severe accidents, additional efforts have to be made to investigate reactor characteristics in the following situations:

  • Presence of cavities due to removal of one or several fuel cells or assemblies, when the reactor core is damaged
  • Partial absence of water because of coolant leakage under accident conditions

These situations cannot be treated properly by traditional methods, because the localized effects described above can lead to prompt neutron reactivity conditions, and, therefore, should be treated as accurately as possible. For this purpose the development of a methodology that allows to analyze the physical approximations and to take into account higher-order terms in the neutron transport equation expansion is necessary.

As results of the completed project the system of methodologies, verified codes, libraries of input and output data of mathematical benchmark-tests that are the most important for VVER and PWR reactor cells and assemblies will be obtain. Firstly, these methodologies and codes are intended for obtaining diffusion characteristics of nuclear reactor cells and assemblies for use in global reactor analysis. Thus, it will be possible to achieve a consistent analysis of the influence of the physical approximations in the most important reactor core parameters that define safety conditions of reactor operation, mainly the reactivity coefficients.

This consistent analysis will permit to obtain a verified methodology for the estimation of safety conditions of reactor operation. This methodology will be based on the complex of the codes which has the accuracy of reference codes with calculation times of current engineering codes. It is important that the calculation accuracy can be chosen according to the demands of the different stages of calculation analysis, such as estimation analysis, design, precision analysis

Furthermore, the results of calculation research of this project concerning anisotropic scattering analysis are very important both as new knowledge about the influence of anisotropic scattering into neutron diffusion in complicated heterogeneous media and from the point of view of creation and elaboration of optimal methodology for nuclear reactor cell codes.

The developments for the problem of plutonium disposition as a component of nuclear reactor fuel and the development methodology for nuclear reactor safety as well meet ISTC goals and objectives because they are directed at humanity peaceful purpose: energy production, non-proliferation of nuclear weapon materials and environmental protection problems

Realization of this project supposes development of the contacts between Russian and west researchers: definition of joint benchmark-tests, change of the results obtained both with Russian and west codes. Joint seminars are supposed to be held regularly. The results of the researches are supposed to be discussed at the seminars and presented in international meetings.

Full activity of the project is 3820 person-days, activity of weapon scientists who possess knowledge and skills related to weapons of mass destruction is 1920 person-days.


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