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A Pctran Based Analysis On The Effect Of Breaksize And Comparative Study Between Hot And Cold Leg Loss Of Coolant Accidents In Vver 1200 Power Reactor

Author

Listed:
  • Salauddin Omar

    (Department of Mechanical and Production Engineering, Ahsanullah University of Science and Technology, Dhaka-1208, Bangladesh)

  • Mohammad Nasim Hasan

    (Department of Mechanical Engineering, Bangladesh University of Engineering and Technology, Dhaka-1000, Bangladesh)

Abstract

In this paper, a comparative analysis of loss of coolant accident (LOCA) in hot leg and cold leg of primary circuit in a VVER 1200 nuclear power plant is investigated. The effect of break size on the severity of the accident is observed. The break size was varied in the range 200-11350 cm2. For all the accident scenarios, station blackout (SBO) condition is set up. Additionally, it is assumed that no ECCS (Emergency Core Cooling System) is available due to system malfunction. The whole scenario is simulated in PCTRAN (Personal Computer Transient Analyzer) software. Results reveal that with the increase in the size of the break area, the core uncovering time decreases sharply. However, for a break size of 2800 cm2 or smaller, the water level in the core doesn’t drop to zero, indicating that the core is partially uncovered throughout the accident scenario. In case of hot leg LOCA, the draining of the reactor vessel is observed to be more rapid compared to cold leg LOCA, while the core melting started earlier in case of cold leg.

Suggested Citation

  • Salauddin Omar & Mohammad Nasim Hasan, 2022. "A Pctran Based Analysis On The Effect Of Breaksize And Comparative Study Between Hot And Cold Leg Loss Of Coolant Accidents In Vver 1200 Power Reactor," Acta Mechanica Malaysia (AMM), Zibeline International Publishing, vol. 5(2), pages 31-34, July.
    • Handle: RePEc:zib:zbnamm:v:5:y:2022:i:2:p:31-34
    DOI: 10.26480/amm.02.2022.31.34
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    References listed on IDEAS

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    1. Abid Hossain Khan & Md Shafiqul Islam, 2019. "A Pctran-Based Investigation On The Effect Of Inadvertent Control Rod Withdrawal On The ThermalHydraulic Parameters Of A Vver-1200 Nuclear Power Reactor," Acta Mechanica Malaysia (AMM), Zibeline International Publishing, vol. 2(2), pages 32-38, October.
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