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LOCA analysis of BWR-4/Mark-I nuclear power plant with TRACE

LOCA-Analyse des Kernkraftwerks BWR-4/Mark-I mit TRACE
  • J.-J. Huang EMAIL logo , S.-W. Chen , J.-R. Wang , C. Shih and H.-T. Lin
From the journal Kerntechnik


This study established an RCS-Containment coupled model that integrates the reactor coolant system (RCS) and the containment system by using the TRACE code. The coupled model was used in both short-term and long-term loss of coolant accident (LOCA) analyses. Besides, the RELAP5/CONTAN model that only contains the containment system was also developed for comparison. For short-term analysis, three kinds of LOCA scenarios were investigated: the recirculation line break (RCLB), the main steam line break (MSLB), and the feedwater line break (FWLB). For long-term analysis, the dry-well and suppression pool temperature responses of the RCLB were studied. The analysis results of RELAP5/CONTAN and TRACE models are benchmarked with those of FSAR and RELAP5/GOTHIC models, and it appears that the results of the above four models are consistent in general trends.


In dieser Studie wurde ein kombiniertes TRACE-Modell erstellt, das das Reaktorkühlsystem (RCS) und das Containment-System zusammenführt. Das gekoppelte Modell wurde sowohl für kurzfristige als auch für langfristige Kühlmittelverlust-Unfallanalysen (LOCA) verwendet. Daneben wurde zum Vergleich auch ein RELAP5/CONTAN-Modell nur für das Containment entwickelt. Für die Kurzzeitanalyse wurden drei Arten von LOCA-Szenarien untersucht: der Bruch der Rezirkulationsleitung (RCLB), der Frischdampfleitungsbruch (MSLB) und der Speisewasserleitungsbruch (FWLB). Für die Langzeitanalyse wurden die Temperaturrückwirkungen des RCLB auf das Drywell und das Suppressionsbecken untersucht. Die Analyseergebnisse der Modelle RELAP5/CONTAN und TRACE werden mit denen der Modelle FSAR und RELAP5/GOTHIC verglichen, und es zeigt sich, dass die Ergebnisse der oben genannten vier Modelle in den allgemeinen Trends übereinstimmen.



Automatic Depressurization System


Boiling Water Reactor


Core Spray


Chinshan Nuclear Power Plant


Design Basis Accident


Emergency Core Cooling Systems


Final Safety Analysis Report


Feedwater Line Break


High-Pressure Coolant Injection System


Loss of Coolant Accident


Low-Pressure Coolant Injection System


Main Steam Isolation Valves


Main Steam Lines


Main Steam Line Break


Net Positive Suction Head


Original Licensed Thermal Power


Peak Cladding Temperature


Pressure and Temperature


Reactor Core Isolation Cooling System


Recirculation Line Break


Reactor Coolant System


Residual Heat Removal


Reactor Pressure Vessel


Symbolic Nuclear Analysis Package


Safety/Relief Valves


Taiwan Power Company


TRAC/RELAP Advanced Computational Engine


interfacial unit volume force


unit volume force between wall and gas mixture


unit volume force between wall and liquid


gas mixture internal energy


liquid internal energy

gravity vector

vapor enthalpy of the bulk vapor if the vapor is condensing or the vapor saturation enthalpy if the liquid is vaporizing


total pressure


power deposited directly to the gas mixture


power deposited directly to the liquid


heat-transfer rate per unit volume between wall and gas mixture


heat-transfer rate per unit volume between wall and liquid


heat-transfer rate per unit volume between wall and saturation fluid



gas mixture velocity vector

liquid velocity vector


void fraction


gas mixture density


liquid density


interfacial mass-transfer rate (positive from liquid to gas)


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Received: 2020-09-14
Published Online: 2021-03-30
Published in Print: 2021-04-30

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