United States Nuclear Regulatory Commission - Protecting People and the Environment

Analysis of the VTI Test Data on the Behavior of the Heated Rod Temperatures in the Partially Uncovered VVER-440 Core Model Using RELAP5/MOD3.2.2 Gamma (NUREG/IA-0208)

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Publication Information

Date Published: July 2002

Prepared by:
V.A. Vinogradov, and A.Y. Balykin

Russian Research Center "Kurchatov Institute"
Kurchatov Square 1
123182, Moscow, Russia

Prepared as part of:
The Agreement on Research Participation and Technical Exchange
under the International Code Application and Maintenance Program (CAMP)

Office of Nuclear Regulatory Research
U.S. Nuclear Regulatory Commission
Washington, DC 20555-0001

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Abstract

This report has been prepared as a part of the Agreement on Research Participation and Technical Exchange under the International Code Application and Maintenance Program.

VTI test data on the behavior of the heated rod temperatures in the partially uncovered VVER-440 core model were simulated with RELAP5/MOD3.2.2GAMMA to assess the code, especially its heat transfer models for modeling phenomena in the partially uncovered core under Small Break LOCA conditions. This problem addresses the phenomena of high importance to VVER-440 safety.

Series of the experiments have been carried out in the VVER-440 loop model at the VTI Test Facility which are directly related to this issue. Two tests conducted in the stationary conditions with the transition mode of a steam flow in the core channel were chosen for the assessment calculations with the code.

Experimental VVER-440 loop model includes the models of all the main elements of a reactor, loop's hot leg model and cold leg simulator, and also a steam generator simulator with an active heat removal. The fuel assembly model consists of 19 electrically heated rod simulators of 9.1 mm outer diameter and 2.5 m heated height. The rod simulators are composed in the rod bundle in a hexagonal array with a pitch equal 12.2 mm (P/D=1.34).

First a study of the effect of the hydraulic nodalization to the code results was performed using different number of hydraulic volumes for the core model. After the choice of proper nodalization and maximum user-specified time step, the base case calculations were done for the tests. The differences between the code predictions for the behavior of rod's wall temperatures and test data are described and analyzed.

Sensitivity studies were carried out to investigate the influence of an increase in the calculated coefficients of heat transfer from the heated rods to a steam flow on the axial distribution of rod's wall temperatures in the uncovered part of core model.

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