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Codes for Nuclear Safety Evaluation

#0815.2


Development of Computerized Technology for Critically Safety Uncertainty Evaluation based on the Analysis of Data for the International Bank for Critical Experiments

Tech Area / Field

  • FIR-MOD/Modelling/Fission Reactors

Status
8 Project completed

Registration date
25.07.2001

Completion date
06.05.2009

Senior Project Manager
Tocheny L V

Leading Institute
FEI (IPPE), Russia, Kaluga reg., Obninsk

Collaborators

  • Oak Ridge National Laboratory, USA, TN, Oak Ridge\nIdaho National Engineering and Environmental Laboratory, USA, ID, Idaho Falls

Project summary

Almost all stages of the external reactor fuel cycle require criticality calculations and criticality safety evaluations to be performed, i.e., determination of conditions under which uncontrolled development of chain reaction is possible, and proving the practical impossibility of originating of such conditions.

The necessity to evaluate realistically the uncertainties of criticality calculations arises when criticality safety analysis is performed for nuclear fuel storage facilities (especially of spent fuel), casks and containers for fuel transportation to chemical reprocessing and refabrication plants, technological equipment of those plants, and waste disposal locations containing nuclear fuel.

Evaluating criticality calculation uncertainty for an emergency reactor is a very complex task, especially when the emergency situation leads to changes in fissile system characteristics such as nuclide composition, neutron spectrum, etc.

Nowadays there is no generally accepted engineering methodology for evaluating criticality calculation uncertainty. The objective of the present project is to develop such a method that could be proposed to the world community, at least as a basis for further development.

The foundation assumed to be used for constructing the methodology for evaluating calculation criticality prediction is the International Handbook of Evaluated Criticality Safety Benchmark Experiments (hereinafter ICSBEP Handbook) created and continuously updated with participation of Russian specialists within the frame of the project on criticality safety benchmark evaluation. The ICSBEP Handbook contains reliably evaluated results of numerous critical experiments conducted in the context of defense activities in various countries (mainly in the United States, Russia, France, Great Britain, and Japan), as well as of experiments performed in these countries for substantiation of criticality safety at external fuel cycle enterprises. Critical systems investigated are persified in terms of fuel and moderators, enrichments, neutron spectra, and geometries, and they substantially complement the narrow set of critical experiments simulating traditional reactors. Contemporary calculation analysis of all these experiments allows expected realistic reception of evaluations of uncertainties, even when the criticality of extremely exotic fissile systems is calculated.

Thus, the Project aims to ensure the use of experimental results obtained over decades in the context of defense activities for increasing the level of criticality safety of civil nuclear power objects. In the course of work on the project, activities were undertaken that complement the Handbook and contribute special chapters containing description and grounding of uncertainty covariance matrixes for inpidual groups of the most informative experiments. The team of specialists involved in the Project have many years of cooperative experience with foreign colleagues in the frame of the ICSBEP, and their work provides a strong prerequisite for international coordination.

The project objective is to develop a calculation tool for realistic evaluation of the uncertainty of calculation predictions for criticality parameters of fissile systems used in nuclear power engineering and in enterprises involved with the external fuel cycle.

The calculation tool will include:

• Codes for calculating critical parameters and the sensitivity of these results in terms of initial data (nuclear cross-sections and technological parameters)


• Uncertainty covariance matrix bank for nuclear cross-sections;
• Uncertainty covariance matrix bank for the most informative evaluated critical experiments;
• Codes for critical parameters uncertainty evaluation based on the analysis experience of combination of reliably evaluated critical experiments.


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