Analysis of Steam Line Break Accident Using PCTRAN Model of VVER-1200 NPP

Muhammed Mufazzal Hossen, Mohammad Khalaquzzaman, S. K. Anisur Rahman

Abstract

The investigation of thermal-hydraulic parameters during steam-line break (SLB) accidents is performed by applying the personal computer transient analyzer (PCTRAN) simulator model of the VVER-1200 nuclear power plant (NPP). Five cases, namely, 0.005 m2 (Case-1), 0.01 m2 break (Case-2), 0.02 m2 break (Case-3), 0.04 m2 (Case-4), and 0.08 m2 (case-5) of SLB accident inside containment with the concurrent loss of AC power have been simulated. There was no variation in the timing of the trip of the reactor coolant pumps, the main feedwater pumps, or the turbine in any of the five SLB accidents. However, the reactor scram's onset time varies slightly between the five scenarios.  Pressure and temperature in the reactor coolant system (RCS) quickly reached a peak following the start of the SLB accident, fell shortly after the reactor scram, and eventually stabilized in all cases. In comparison to the larger breaks in the SLB accident, the smaller breaks result in a higher RCS temperature and pressure. After the SLB accident, the pressurizer's liquid level rises and then quickly drops in all cases. The break mass flow rate from the steam line rapidly increases until the occurrence of the reactor scram and then decreases to a stabilized value. Steam generator A has a faster rate of heat removal rate than steam generator B, and its pressure and liquid level decrease more quickly than those of steam generator B. The thermal power of the reactor, peak cladding temperature, and fuel temperature showed a rapid drop after the initiation of the SLB accident. There was no increase in these parameters from the initial state of the simulation. The radiation in the air of the reactor building and steam line was very low during the simulation period. Therefore, there was no violation of the safety aspects of the SLB accident of the PCTRAN simulation of the VVER-1200 NPP model.

Keywords

PCTRAN; PWR; Steam line break accident; Steam generator, Thermal-hydraulics; VVER-1200.

Article Metrics

Abstract view : 191 times
PDF - 130 times

Full Text:

PDF

References

Chon-Kwo Tsai, Mujid S. Kazimi and Allan F. Henry, Three dimensional effects in analysis of PWR steam line break accident, Energy Laboratory Report, MIT-EL 85-004, 1985, 1-206.

A. S. Ekariansyah, Deswandri and Geni R. Sunaryo, Main steam line break accident simulation of APR1400 using the model of ATLAS facility, Journal of Physics: Conference Series, 962, 2017, 012037.

Y. Alzaben, V. H. Sanchez-Espinoza and R. Stieglitz, Analysis of a steam line break accident of a generic SMART-plant with a boron-free core using the coupled code TRACE/PARCS, Nuclear Engineering and Design, 350, 2019, 33-42.

L. D. Dien and D. N. 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)), https://aris.iaea.org/PDF/VVER-1200(V-491).pdf, 2011.

ROSATOM, The VVER today: Evolution, Design, Safety. https://www.rosatom.ru/upload/iblock/0be/0be1220af25741375138ecd1afb18743.pdf (Accessed on 30/12/2022).

Micro-Simulation Technology (MST) Inc., Personal Computer Transient Analyzer, http://microsimtech.com/, 2019 (Accessed on 30/12/2022).

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.

Md. Mehedi Hasan Tanim, Md. Feroz Ali, Md Asaduzzaman Shobug and Shamsul 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.

Salauddin Omar and Mohammad Nasim Hasan, A PCTRAN based analysis on the effect of break size and comparative study between hot and cold leg loss of coolant accidents in VVER 1200 power reactor, Acta Mechanica Malaysia, 5(2), 2022, 31-34.

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.

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), Rajshahi, Bangladesh, 2021, 1-6.

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, Energy Procedia, 160, 2019, 162-169.

Muhammed Mufazzal Hossen, Analysis of thermal-hydraulics parameters during steam generator tube rupture event of VVER-1200 NPP Using PCTRAN simulator, Applications of Modelling and Simulation, 6, 2022, 28-35.

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.

Microsimulation Technology Inc., PCTRAN VVER 1200, http://www.microsimtech.com/VVER1200/VVER1200d.html (Accessed on 02/01/2023).

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.

J. J. Carbajo, G. L. Yoder, E. Popov and V. K. Ivanov, Main-steam-line-break accident analyses in a VVER-1000 reactor, Oak Ridge National Laboratory, Oak Ridge, Tennessee, 3783, and Kurchatov Institute, Moscow, Russia, https://technicalreports.ornl.gov/cppr/y2001/pres/111689.pdf.

M. Pavlova, M. Andreeva and P. Groudev, Steam line break investigation at full power reactor for VVER-1000/V320M. Nuclear Engineering and Design, 285, 2015, 65-74.

International Atomic Energy Agency, Accident analysis for nuclear power plants with pressurized water reactors, IAEA Safety Report Series, Vienna, 3, 2003.

Refbacks

  • There are currently no refbacks.