D4 Published development or research report or study

PACTEL Small Break LOCA Experiment SBL-30 Calculation with TRACE Code


Open Access publication

Publication Details
Authors: Vihavainen Juhani
Publisher: U.S. Nuclear Regulatory Commission
Publishing place: Washington, DC
Publication year: 2018
Language: English
JUFO-Level of this publication: 0
Open Access: Open Access publication

Abstract

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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Last updated on 2019-13-03 at 12:00