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Cross –Verification of Neutron-Physical Codes


Cross-Verification and Analysis of Accuracy of Neutron-Physical Codes for Nuclear Reactors RBMK and CANDU

Tech Area / Field

  • FIR-MOD/Modelling/Fission Reactors

3 Approved without Funding

Registration date

Leading Institute
Kurchatov Research Center, Russia, Moscow


  • Atomic Energy Canada Limited, Canada, ON, Chalk River\nAdvanced Energy Technologies Consulting Inc., Canada, ON, Oshawa\nUniversity of Ontario Institute of Technology / Faculty of Energy Systems and Nuclear Science, Canada, ON, Oshawa

Project summary

Nuclear Energy is one of the most important energy technologies. Priority problem of development of Nuclear Energy is increasing its safety and economic efficiency. New variants of improved nuclear reactor core design have been proposed. These new directions of development of the designs are known as Generation IV Nuclear Reactors. It seems that using reactor coolant with supercritical parameters is one of such improvement. Also it seems that conception of new variant of core design as channel core design with light water supercritical coolant (so called as “Criss-Cross conception) is perspective. Every conception needs serious studying before becoming as viable, so a lot of theoretical speculations and experimental analysis are needed. Carrying out of experiments both on critical assemblies and on heating stands is very expensive so it is impossible to overestimate the role of mathematical modeling of neutron physical processes in nuclear reactor and thermo hydraulic processes as well. Availability of suitable calculation tool that could simulate neutron behavior and reactor performance under normal conditions and during unforeseen accidents is necessary component for elaboration of a conception of new reactor design.

The analysis of estimation of the accuracy of neutron-physical calculations is important in modeling / testing of various new designs of a reactor core or some parts of the core. This problem is really an issue in many current challenges, such as development of next-generation reactors in frame of Generation IV project, and, actually, in any full core simulation of channel-type reactors.

The objective of this proposal is formulation and carrying out a series of cross test calculations of channel reactors (CANDU and RBMK) using different computer codes to understand advantages and disadvantages of some methods and codes and estimate possible inaccuracies as well.

As results of completed project the system of methodologies, verified codes, libraries of input and output data of mathematical benchmark-tests that are the most important for RBMK and CANDU reactor cells and assemblies will be obtain. This consequent 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 that has the accuracy of reference codes with calculation times like 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, design, precision analysis.

Analysis of benchmark tests has indicated that current available engineering reactor codes often need unjustified adjustments to predict correct results that have enough precision for estimation of exploitation parameters firstly it is concerned to criticality, energy distribution on the boundary of MOX-fuel assembly and uranium-fuel assembly. Also currently codes are not able to analyze some important neutron transport problems in principle.

Elaboration of new variants of nuclear reactor core design to improve its economic efficiency and safety meets ISTC goals and objectives because it is directed at environmental protection problems and energy production.

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 2880 person-days, activity of weapon scientists who possess knowledge and skills related to weapons of mass destruction is 1680 person-days.


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.

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