Analyses of KS Test Data on the Heated Rod Bundle Temperature Behavior in RBMK-1500 Core Model Under Stop and Recovery Flow Using RELAP5/MOD3.2 and RELAP5/MOD3.2.2 GAMMA (NUREG/IA-0202)

On this page:

• Publication Information
• Abstract

Download complete document

• NUREG/IA-0202 (PDF 3.57 MB)

Publication Information

Date Published: April 2001

Prepared by:
V.A. Vinogradov, A.Y. Balykin
Nuclear Reactor Institute, 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

Availability Notice

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.

KS experimental data on the behavior of the heated rod bundle temperature in the RBMK-1500 core model under stop and recovery flow conditions were simulated with RELAP5/MOD3.2 and RELAP5/MOD3.2.2GAMMA to assess the codes suitability. Especially calculations were performed to estimate differences in the code version predictions for processes/phenomena, which could occur under specific conditions of RBMK type reactor during LOCA with MCP's pressure header rupture and subsequent ECCS water injection.

This problem addresses phenomena of high importance to RBMK-1500 safety including water discharge, critical heat flux, post dryout heat transfer, reflood and propagation of rewetting front in the fuel channel during an accident.

The test have been carried out at KS semi-integral test facility (Russian Research Center Kurchatov Institute) represented RBMK primary circuit. The main purpose of the experiment was investigation of temperature conditions in the fuel assembly under high power value, inlet flow rate being decreased drastically up to complete stop of the flow followed by pressure header rupture. A subsequent flow rate resumption was also assumed, the fuel assembly power high level being unchanged.

First a study of the effect of the hydraulic nodalization to the code calculations was performed using different number of hydraulic volumes for the fuel channel model. After the choice of proper nodalization and maximum user-specified time step, base case calculations were done for the test.

Sensitivity studies were carried out to investigate the effects of modeling on the behavior of the rod simulator temperatures along the height of the fuel assembly model.