Series: Nuclear and Reactor Constants

since 1971

Русский (РФ)

ISSN 2414-1038 (online)


Usage of calculation codes in the overall safety analysis system is reduced to verifying compliance with the acceptance criteria limiting the effects of abnormal situations. The primary way of verifying compliance the acceptance criteria is the calculation of the reactor performance during an abnormal situation. Some calculation model parameters (design, operating and referring to mathematical models) are known with specific uncertainty. Variation of the values of these parameters makes certain impact on the calculation results. Evaluation of the uncertainty of calculated characteristics due to variation of calculation model parameters is the primary goal of applying the method of uncertainty and sensitivity analysis.
The issue is topical due to the fact that new reactor designs and operating VVER units are facing the requirement of unit energy extension, and to meet this objective it is necessary to evaluate (and ultimately reduce) the degree of conservatism in safety analyses.
The report outlines the problem as seen by the developers of the neutronic part of the code. It also describes an approach to selection of variable neutronic model parameters and their variation when using the uncertainty and sensitivity method, taking into account the peculiarities of typical design basis accidents whose analysis requires coupled neutronic and thermal hydraulic calculations. The experience of using the method of uncertainty and sensitivity analysis showed that feasibility of its findings essentially depends on the quality and volume of code verification. This is due to the fact that the uncertainty estimations of the calculated characteristics, obtained during the verification process, determine the parameter range of the calculation models, and, therefore, the range of calculation result uncertainty. Through the example of modeling specific emergency conditions, the paper demonstrates the application procedure of the method of uncertainty and sensitivity analysis using a coupled neutronic and thermal hydraulic model of VVER-1000, prepared on the basis of KORSAR/GP and SAPFIR_95&RC_VVER codes. Uncertainties of calculation results are estimated at variation of model parameters, corresponding to the nominal uncertainties of the codes ORSAR/GP and SAPFIR_95&RC_VVER certificated in the Russian Federal Service for Ecological, Technical and Atomic Supervision.

Key words
Monte-Carlo method, RC calculation code, KORSAR code, uncertainty and sensitivity analysis, neutronic processes, thermal-hydraulic processes, neutron kinetics, verification

Article Text (PDF, in Russian)


UDC 621.040.59:621.039.51...17

Problems of Atomic Science and Technology. Series: Nuclear and Reactor Constants", issue 3:8, 2014