Analysis of Thermal-Hydraulics Parameters During Steam Generator Tube Rupture Event of VVER-1200 NPP Using PCTRAN Simulator

Muhammed Mufazzal Hossen


The overall performance of steam generators plays a significant role in ensuring the safety of a nuclear power plant (NPP) operation. The analysis of thermal-hydraulic parameters during a steam generator tube rupture (SGTR) event of VVER-1200 NPP is conducted by applying the personal computer transient analyzer (PCTRAN) simulator. Four cases, namely, 25%, 50%, 75% and 100% of one tube rupture in two steam generators with the concurrent loss of AC power have been performed. Among the four cases, major variation in time was not observed for the occurrence of the reactor scram, reactor coolant pump trip, main feed-water pump trip, and turbine trip. The pressure and the temperature of the reactor coolant system (RCS) increase rapidly to a peak value due to event initiation, and drop promptly after the reactor scram. The stabilized pressure and temperature of the RCS are higher for the smaller break size of the SGTR. The secondary pressure of the steam generator is also increased to a peak value, followed by an increasing and decreasing trend, in turn, due to the repeated opening and closing of safety relief valves of the steam generators. The liquid level of the pressurizer is increased rapidly due to the liquid surge towards the pressurizer after the event and it is stabilized after the opening of the safety relief valve. The stabilized liquid level of the pressurizer and the steam generator is higher for the smaller break size of the SGTR. The earlier emergency core coolant injection to the reactor was required for the larger break size of the SGTR. There is no increase in the peak cladding temperature and the peak fuel temperature during the calculation period for all these cases. The results of this study provide a valuable understanding of SGTR events with the concurrent loss of AC power for the PCTRAN model of the VVER-1200 NPP.


Nuclear power plant; PCTRAN; Steam generator tube rupture; Thermal-hydraulics; VVER-1200.

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M. R. Nematollahi and A. Zare, A simulation of a steam generator tube rupture in a VVER-1000 plant, Energy Conversion and Management, 49, 2008, 1972-1980.

Zbigniew Koszela and Łukasz Sokołowski, Thermal-hydraulic analysis of single and multiple steam generator tube ruptures in a typical 3-loop PWR, Journal of Power Technologies, 95 (3), 2015, 175-182.

J. P. Adams and M. B. Sattison, Frequency and consequences associated with a steam generator tube rupture event, Nuclear Technology, 90(2), 1990, 168-185.

C. S. Lin, A. T. Wassel, S.P. Kalra and A. Singh, The thermal-hydraulics of a simulated PWR facility during steam generator tube rupture transients, Nuclear Engineering and Design, 98(1), 1986, 15-38.

I. Parzer, S. Petelin and B. Mavko, Feed-and-bleed procedure mitigating the consequences of a steam generator tube rupture accident, Nuclear Engineering and Design, 154, 1995, 51-59.

Micro-Simulation Technology (MST) Inc., Personal Computer Transient Analyzer,, 2019 (Accessed 20.12.2021).

Pronob Deb Nath, Kazi Mostafijur Rahman and Md. Abdullah Al Bari, Thermal hydraulic analysis of a nuclear reactor due to loss of coolant accident with and without emergency core cooling system, Journal of Engineering Advancements, 01(02), 2020, 53-60.

Abid Hossain Khan, Md. Ibrahim Al Imran, Nashiyat Fyza and M. A. R. Sarkar, A numerical study on the transient response of VVER-1200 plant parameters during a large-break loss of coolant accident, Indian Journal of Science and Technology, 12(27), 2019, 1-12.

Nashiat Fyza, Altab Hossain and Rashid Sarkar, Analysis of the thermal-hydraulic parameters of VVER-1200 due to loss of coolant accident concurrent with loss of offsite power, Energy Procedia, 160, 2019, 155-161.

M. M. Hasan Tanim, M. Feroz Ali, M. A. Shobug and S. Abedin, Analysis of various types of possible fault and consequences in VVER-1200 using PCTRAN, 2020 International Conference for Emerging Technology (INCET), Belgaum, India, 2020, 1-4.

Abid Hossain Khan and Md Shafiqul Islam, A PCTRAN-based investigation on the effect of inadvertent control rod withdrawal on the thermal -hydraulic parameters of a VVER-1200 nuclear power reactor, Acta Mechanica Malaysia, 2(2), 2019, 32-38.

Abid Hossain Khan, Angkush Kumar Ghosh, Md Sumon Rahman, S. M. Tazim Ahmed and C. L. Karmaker, An investigation on the possible radioactive contamination of environment during a steam-line break accident in a VVER-1200 nuclear power plant, Current World Environment, 14(2), 2019, 299-311.

Arnob Saha, Nashiyat Fyza, Altab Hossain and M. A. Rashid Sarkar, Simulation of tube rupture in steam generator and transient analysis of VVER-1200 using PCTRAN, 2nd International Conference on Energy and Power (ICEP2018), Sydney, Australia, 2018.

M. G. Zakir, A. S. M. Nasim, A. Islam, M. A. Hossain, Md. Rosaidul Mawla and M. A. R. Sarkar, Transient analysis of VVER-1200 nuclear power reactor in the event of AC Power failure, International Conference on Mechanical Engineering and Renewable Energy 2019 (ICMERE2019), Chittagong, Bangladesh, 2019.

S. Akter, M. S. A. Joarder, M. G. Zakir, A. Hossain, M. A. Razzak and M. S. Islam, Comparative analysis of thermal hydraulic parameters of AP-1000 and VVER-1200 nuclear reactor for turbine trip concurrent with anticipated transient without SCRAM (ATWS), 2021 International Conference on Automation, Control and Mechatronics for Industry 4.0 (ACMI), Rajashi, Bangladesh, 2021, 1-6.

Le Da Dien and Do Ngoc Diep, Verification of VVER-1200 NPP simulator in normal operation and reactor coolant pump coast-down transient, World Journal of Engineering and Technology, 5, 2017, 507-519.

Advanced Reactors Information System (ARIS), Status report 108 - VVER-1200 (V-491) (VVER-1200 (V-491)),, 2011.

ROSATOM, The VVER today: Evolution, Design, Safety, (Accessed 22.12.2021).

Microsimulation Technology Inc., PCTRAN VVER 1200, (Accessed 15.12.2021).

Yi-Hsiang Cheng, Chunkuan Shih, Show-Chyuan Chiang and Tung-Li Weng, Introducing PCTRAN as an evaluation tool for nuclear power plant emergency responses, Annals of Nuclear Energy, 40, 2012, 122-129.

International Atomic Energy Agency, Accident Analysis for Nuclear Power Plants with Pressurized Water Reactors, IAEA Safety Report Series, 3, Vienna, 2003.


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