Nuclear Safety of MOX Fueled VVER
System Analysis of Uncertainties of Nuclear Safety Characteristics for VVER Reactors with Various Fraction of MOX Fuel in the Core
Tech Area / Field
- FIR-EXP/Experiments/Fission Reactors
- FIR-NSS/Nuclear Safety and Safeguarding/Fission Reactors
8 Project completed
Senior Project Manager
Tocheny L V
Federal State Unitary Enterprise Research and Development Institute of Power Engineering named after N.A.Dollezhal, Russia, Moscow
- Kurchatov Research Center, Russia, Moscow\nITEF (ITEP), Russia, Moscow
- FRAMATOME, France, Paris La Défense\nGesellschaft für Anlagen und Reaktorsicherheit mbH, Germany, Köln
Project summaryThe project proposed is devoted to the solution of the same problems of the provision of safety of power reactors of VVER type under the use of military-grade and reactor-grade plutonium for their MOX load, and belongs to the main group of works planned to be carried out in Russia. The project participants are the major Russian reactor Institutions (RRC ‘KI”, IPPE, ITEP, RDIPE). The project is coordinated with the scientific leader of MOX program in Russia Academician N.N. Ponomarev-Stepnoi and is approved by Minatom RF.
Еhe project plan is discussed also and coordinated with participants of the international CEG MOX meeting (Brussels, October 2000).
The main goal of the project investigations is to analyze the problems and to find the solutions of how to provide a necessary accuracy of calculation of neutron-physical characteristics of VVER reactors with MOX fuel, namely, the characteristics that define their nuclear safety, in particular, the origination and development of severe accidents.
The methodical basis for the project proposed is formed by the results of the already fulfilled ISTC project #116 «Development of methodical and calculation technology verification of nuclear data bases used in the calculation of neutron-physical characteristics and in analysis of nuclear safety of reactor facilities and nuclear conversion technological processes».
In ISTC project #116 developed was such a combination of methods, codes and data bases that allows to perform calculation analysis of VVER reactors with MOX load with an accuracy comparable to that provided by reference methods. For example, the universal combined library of evaluated nuclear data B645 provides an essentially more high accuracy of calculation of neutronic characteristics of VVER reactor with U, Pu, mixed U-Pu loads than the files used as its sources (ENDF/B-4,5,6, JENDL-3, JEF) : error of calculation eff 0.5 % , the accuracy of calculation of reaction rates (power) in a fuel assembly as provided by these methods is different from the reference evaluations by only 1-2%, the accuracy of reactivity effects – by 0.1%,
The main goals of the project proposed are:
- to finish as a product for practical use the methods, codes and data bases developed in the fulfilled project #116 and others , that would enable to improve the guaranteed accuracy of simulation of neutronic characteristics of VVER –type reactors with MOX fuel;
- to analyze the impact of errors and uncertainties of different nature upon safety parameters of VVER reactors with MOX load (effects and coefficients of reactivity, reactivity balance);
- to investigate the initial conditions of development of more typical scenarios of severe accidents. This, in turn, would require to calculate the kinetics parameters for transient processes of VVER reactor with MOX fuel in the course of severe accident development.
Thus, the main goals of the project are:
· To perform a calculation and analytical investigation of the impact of errors and uncertainties of different nature upon predicted neutron-physical characteristics of VVER reactors with MOX fuel, to define the ways of provision of the required accuracy;
· To study the impact of uncertainties upon initial conditions of development of accidents, including the severe ones, with the goal to eliminate such accidental situations.
As a result of the project fulfillment, obtained would be the set of methods, codes, and nuclear data libraries the practical application of which would provide an essential increase in the accuracy of calculation of neutronic characteristics of VVER reactors with plutonium load for practically any point in reactor fuel lifetime, particular results of calculation investigations of the impact of errors of different nature upon predicted neutron-physical characteristics of VVER reactors with MOX fuel, results of reactivity balance/error analysis for possible scenarios of severe accidents, etc.
The International Science and Technology Center (ISTC) is an intergovernmental organization connecting scientists from Kazakhstan, Armenia, Tajikistan, Kyrgyzstan, and Georgia with their peers and research organizations in the EU, Japan, Republic of Korea, Norway and the United States.
ISTC facilitates international science projects and assists the global scientific and business community to source and engage with CIS and Georgian institutes that develop or possess an excellence of scientific know-how.