Assessment of TRACE V5.0 Patch 7 Using OECD-ATLAS2 B3.2 Test (NUREG/IA-0548)

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Publication Information

Manuscript Completed: February 2024
Date Published: April 2024

Prepared by:
Seung Hun Yoo*, Kyung-Won Lee*, Dong Gu Kang*, Andong Shin*

*Korea Institute of Nuclear Safety (KINS)
62 Gwahak-ro, Yuseong-gu,
Daejeon 34142, Republic of Korea

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

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Abstract

The assessment of TRACE V5.0 Patch 7 was performed using the OECD-ATLAS2 B3.2 test which is a 100% Direct Vessel Injection line break in the ATLAS referring to the APR1400. TRACE showed a generally good agreement with most of the sequence of events of the experiment including the injection timing of the Emergency Core Cooling System (ECCS), the minimum core level, and the occurrence of Loop Seal Clearing (LSC) but delayed in the core quenching time. The predicted integrated discharge flow, the mass flow rate of cold legs and hot legs, and the ECCS flow were well-matched with the experiment. TRACE predicted the same Maximum Cladding Temperature (MCT) at the same timing as the experiment. However, the predicted position of the MCT was different from the experiment due to the different predicted local behaviors and the predicted core quenching time was delayed. Through sensitivity studies, it was found that the cladding temperature generally increased in proportion to the break size and the most decisive factor in determining the cladding temperature is the behavior of the minimum core level. In conclusion, TRACE showed an excellent capability to predict the same MCT at the same timing as the experiment. However, it showed a reasonable capability to predict the pressure drop in the reactor vessel for the region of the steam-dominant two-phase flow, the recovery of the core level, and the core quenching time.