Series: Nuclear and Reactor Constants

since 1971

Русский (РФ)

ISSN 2414-1038 (online)

Authors & Affiliations

Artemov V.G., Artemova L.M., Korotaev V.G., Mikheev P.A., Shemaev Yu.P.
A.P. Aleksandrov Research Technological Institute” (NITI), Sosnovy Bor, Russia


In order to evaluate the conservatism of VVER reactor design calculations with regard to fuel element thermal conditions, the model of pin-by-pin fuel assembly calculation was developed and tested with the use of capabilities of thermal hydraulic calculation code KORSAR and a software package for neutronic calculations SAPFIR_95&RC_VVER. The proposed approach is based on the results of pin-to-pin power density simulation by the method of superposiotion of micro- and macro-distributions of neutron fluxes, proven for steady-state neutronic calculations of VVER reactors in the course of SAPFIR_95&RC_VVER verification and cell-by-cell calculation of thermal hydraulic characteristics of certain fuel assemblies, performed with the use of the KORSAR code. The following operations are performed sequentially within the framework of the approach developed in the course of calculation of the specific emergency condition: calculation of the condition by the KORSAR/GP code with the coupled neutronic and thermal hydraulic core models in the channel-by-channel approximation, with determination of the most heat-stressed fuel assemblies, based on the method of uncertainty and sensitivity analysis, and recording of thermo-hydraulic and neutronic parameters varied in time; recovery of pin-to-pin power density distribution for the chosen mostly stressed fuel assemblies, based on the neutron macro-flux values recorded in the course of channel-by-channel dynamic analysis and calculated in advance relative power density micro-distributions in fuel pins of the relevant fuel assemblies; calculation of pin-to-pin characteristics by the KORSAR/GP code with the use of cell-by-cell fuel assembly model, based on the values of boundary thermal hydraulic conditions recorded in the course of channel-by-channel core analysis and recovered power density field in fuel pins. Correctness of the pin-to-pin power density recovery technique has been justified in comparison with the results of test calculation obtained with the software that used Monte Carlo method. Cell-by-cell thermal hydraulic fuel assembly model was verified on the basis of the experiments with the bundles of electrically heated fuel pins. The effectiveness of the proposed approach is demonstrated in the paper by the example of simulating an emergency condition with VVER-1000 steam generator steam line break. The 1-st and 3-rd loadings of the Rostov NPP core were chosen as a prototype, i.e. the source of data for neutronic model.

Key words
VVER-type reactor, core, multi-rod thermohydraulic model, pin-to-pin power density, channel-to-channel approximation, full-circuit calculation, multi-group constant library

Article Text (PDF, in Russian)


UDC 621.039.54:621.039.51...17

Problems of Atomic Science and Technology. Series: Nuclear and Reactor Constants", issue 4:3, 2014