PACTEL Small Break LOCA Experiment SBL-30 Calculation with TRACE Code (NUREG/IA-0484)

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Publication Information

Manuscript Completed: May 2017
Date Published: November 2018

Prepared by:
J. Vihavainen

Lappeenranta University of Technology
P.O. Box 20
FI-53851 Lappeenranta

K. Tien, NRC Project Manager

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

Prepared as part of:
The Agreement on Research Participation and Technical Exchange
Under the Thermal-Hydraulic Code Applications and Maintenance Program (CAMP)

Published by:
U.S. Nuclear Regulatory Commission
Washington, DC 20555-0001

Availability Notice


TRACE is one of the main codes used for performing nuclear power plant thermal-hydraulic safety analysis at present. Therefore, the importance of assessing the TRACE code capability to predict various thermal-hydraulic transients in reactor systems becomes evident. One such transient that can occur small break loss-of-coolant-accident. The natural circulation is of particular interest for code assessment as it requires the system code to accurately predict temperature and density distributions throughout the system. Specific modeling capabilities are required for heat transfer and two-phase flow phenomena.

This research presents the assessment of the PACTEL small break LOCA experiment SBL-30 with the TRACE V5.0 Patch 4. The PACTEL facility is volumetrically scaled full-height model of a six-loop Russian design VVER-440 PWR. This reactor type has specific features like horizontal steam generators and hot leg loop seals. Although the TRACE code has not been originally developed for the special geometry of the VVER-440 reactor type, it was proven that the code is capable for relatively accurate reproducing the natural circulation phenomena at a satisfactory level.

However, some discrepancies between the predicted variables and the experimental data suggests that further investigation of the TRACE modeling is necessary.

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