Araştırma Makalesi
BibTex RIS Kaynak Göster

Yıl 2026, Cilt: 11 Sayı: 1, 143 - 160, 17.03.2026
https://doi.org/10.58559/ijes.1879510
https://izlik.org/JA47MU93GA

Öz

Kaynakça

  • [1] Nath PD, Rahman KM, Bari MAA. Thermal hydraulic analysis of a nuclear reactor due to loss of coolant accident with and without emergency core cooling system. Journal of Engineering Advancements 2020; 1(2): 53–60.
  • [2] Sofrany EA, Widodo S. Analysis of steam generator tube rupture (SGTR) based on Mihama Unit 2 scenario; Analisis kejadian steam generator tube rupture (SGTR) berdasarkan skenario Mihama Unit 2. Indonesia, 2010.
  • [3] Saleh W, Kim J. Assessment of source terms and potential doses due to steam generator tube rupture of VVER-1200 at the El Dabaa nuclear power plant in Egypt. Journal of Radiation Protection and Research 2025; 50(S1): S30–S42.
  • [4] Mondal, J, Afsar SM Anan. Analysis of Nuclear Reactor Parameters in the VVER-1200 during a Loss of Coolant Accident and Steam Generator Tube Rupture with Emergency Core Cooling System Failure. SciEn Conference Series: Engineering 2025; Vol. 3.
  • [5] Chenshi Y, Yaru L. Analysis and research on long-term safety and stability state of VVER-1200 after FLB accident. Proceedings of the 32nd International Conference on Nuclear Engineering (ICONE 2025) Weihai, China, 2025.
  • [6] Bagheri, AA, Zohuri, B. Study on Severe Accidents of the New Generation WWER-1200 Reactor. Journal of Material Sciences & Manufacturing Research 2022; (3), 134.
  • [7] Khan AH, Ghosh AK, Rahman MS, Ahmed SMT, Karmaker CL. Investigation of possible radioactive contamination of environment during a steam-line break accident in a VVER-1200 nuclear power plant. Current World Environment 2019; 14(2): 299–311.
  • [8] Fyza N, Hossain A, Sarkar R. Analysis of thermal-hydraulic parameters of VVER-1200 due to loss of coolant accident concurrent with loss of offsite power. Energy Procedia 2019; 160: 155–161.
  • [9] Khan AH, Imran MIA, Fyza N, Sarkar MAR. Numerical study on transient response of VVER-1200 plant parameters during a large-break loss of coolant accident. Indian Journal of Science and Technology 2019; 12(27).
  • [10] Akter S, Joarder MSA, Zakir MG, Hossain A, Razzak MA, Islam MS. 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.
  • [11] Bui, TH, Tran, CT. Evaluation of VVER-1200/V-491 Reactor Pressure Vessel integrity during large break LOCA along with SBO using MELCOR 1.8. 6. Journal of Nuclear Science and Technology 2015; 5(4), 54-63.
  • [12] Omar S, Hasan MN. 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 2022; 5(2): 31–34.
  • [13] Hossen MM, Ahmed S, Hossain M. Analyzing thermal-hydraulic parameters during cold-leg loss of coolant accident without reactor SCRAM using PCTRAN model of VVER-1200. Acta Mechanica Malaysia 2024; 7(1): 16–22.
  • [14] Barua S, Tabassum M, Joarder MSA, Zakir MG, Abdur M. A comparison of simulation analysis of safety systems and severe accident progression for VVER-1200 and AP-1000 following LOCA in hot leg by PCTRAN. International Conference on Mechanical Engineering and Renewable Energy Chattogram, Bangladesh, 2021.
  • [15] Hassan MS, Sharif I, Khan AH, Sikder D. Evaluation of safety for VVER-1200-based nuclear power plants during main steam-line break accidents. International Journal of Nuclear Energy Science and Technology 2025; 18(2).
  • [16] Hossen MM. Analysis of thermal-hydraulic parameters during steam generator tube rupture event of VVER-1200 NPP using PCTRAN simulator. Applications of Modelling and Simulation 2022; 6: 28–35.
  • [17] Hossen MM, Khalaquzzaman M, Rahman SKA. Analysis of steam line break accident using PCTRAN model of VVER-1200 NPP. Applications of Modelling and Simulation 2023; 7: 8–15.
  • [18] Tanim MMH, Ali MF, Shobug MA, Abedin S. Analysis of various types of possible fault and consequences in VVER-1200 using PCTRAN. 2020 International Conference for Emerging Technology (INCET) Belgaum, India, 2020.
  • [19] Saha A, Fyza N, Hossain A, Sarkar MAR. Simulation of tube rupture in steam generator and transient analysis of VVER-1200 using PCTRAN. Energy Procedia 2019; 160: 162-169.
  • [20] Karim MSS, Datta D, Hossain A. Safety and security analysis of VVER-1200 reactor pressure vessel under pressurized thermal shock. International Journal of Nuclear Security 2024; 9(2).
  • [21] Korotkikh AG, Odii CJ. Analysis of VVER-1200 thermal characteristics using analytical calculation validated by PCTRAN simulator. Discover Mechanical Engineering 2025; 4(1): 79.
  • [22] Microsimulation Technology Inc. PCTRAN VVER-1200 user documentation. Available at: http://www.microsimtech.com/VVER1200/VVER1200d.html. Accessed 10 December 2025.

Early phase behavior of steam line break accidents in the VVER-1200 reactor: Effect of break location, AC power loss, and break size

Yıl 2026, Cilt: 11 Sayı: 1, 143 - 160, 17.03.2026
https://doi.org/10.58559/ijes.1879510
https://izlik.org/JA47MU93GA

Öz

This study investigates the early-phase (0–300 s) behavior of steam line break (SLB) accidents in a VVER-1200 pressurized water reactor, with and without AC power availability, using the PCTRAN simulation code. Break location and power supply status are treated as the primary governing parameters. Eight comparative accident scenarios are developed to represent different break configurations. Reactor building pressure and peak fuel cladding temperature (PCT), which are critical safety indicators, are evaluated as time-dependent responses. The results demonstrate that break location and AC power availability exert the dominant influence on containment pressure evolution. SLB events occurring inside the reactor building lead to pronounced pressure increases, particularly under AC power-loss conditions, whereas breaks outside the building maintain pressure close to atmospheric levels. Across all investigated scenarios, peak fuel cladding temperatures remain below safety limits, indicating that natural circulation and passive safety systems effectively sustain core cooling even during power loss. Overall, the findings confirm the high early-phase safety margin of the VVER-1200 design against SLB accidents and highlight the decisive role of system configuration in transient safety assessment.

Kaynakça

  • [1] Nath PD, Rahman KM, Bari MAA. Thermal hydraulic analysis of a nuclear reactor due to loss of coolant accident with and without emergency core cooling system. Journal of Engineering Advancements 2020; 1(2): 53–60.
  • [2] Sofrany EA, Widodo S. Analysis of steam generator tube rupture (SGTR) based on Mihama Unit 2 scenario; Analisis kejadian steam generator tube rupture (SGTR) berdasarkan skenario Mihama Unit 2. Indonesia, 2010.
  • [3] Saleh W, Kim J. Assessment of source terms and potential doses due to steam generator tube rupture of VVER-1200 at the El Dabaa nuclear power plant in Egypt. Journal of Radiation Protection and Research 2025; 50(S1): S30–S42.
  • [4] Mondal, J, Afsar SM Anan. Analysis of Nuclear Reactor Parameters in the VVER-1200 during a Loss of Coolant Accident and Steam Generator Tube Rupture with Emergency Core Cooling System Failure. SciEn Conference Series: Engineering 2025; Vol. 3.
  • [5] Chenshi Y, Yaru L. Analysis and research on long-term safety and stability state of VVER-1200 after FLB accident. Proceedings of the 32nd International Conference on Nuclear Engineering (ICONE 2025) Weihai, China, 2025.
  • [6] Bagheri, AA, Zohuri, B. Study on Severe Accidents of the New Generation WWER-1200 Reactor. Journal of Material Sciences & Manufacturing Research 2022; (3), 134.
  • [7] Khan AH, Ghosh AK, Rahman MS, Ahmed SMT, Karmaker CL. Investigation of possible radioactive contamination of environment during a steam-line break accident in a VVER-1200 nuclear power plant. Current World Environment 2019; 14(2): 299–311.
  • [8] Fyza N, Hossain A, Sarkar R. Analysis of thermal-hydraulic parameters of VVER-1200 due to loss of coolant accident concurrent with loss of offsite power. Energy Procedia 2019; 160: 155–161.
  • [9] Khan AH, Imran MIA, Fyza N, Sarkar MAR. Numerical study on transient response of VVER-1200 plant parameters during a large-break loss of coolant accident. Indian Journal of Science and Technology 2019; 12(27).
  • [10] Akter S, Joarder MSA, Zakir MG, Hossain A, Razzak MA, Islam MS. 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.
  • [11] Bui, TH, Tran, CT. Evaluation of VVER-1200/V-491 Reactor Pressure Vessel integrity during large break LOCA along with SBO using MELCOR 1.8. 6. Journal of Nuclear Science and Technology 2015; 5(4), 54-63.
  • [12] Omar S, Hasan MN. 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 2022; 5(2): 31–34.
  • [13] Hossen MM, Ahmed S, Hossain M. Analyzing thermal-hydraulic parameters during cold-leg loss of coolant accident without reactor SCRAM using PCTRAN model of VVER-1200. Acta Mechanica Malaysia 2024; 7(1): 16–22.
  • [14] Barua S, Tabassum M, Joarder MSA, Zakir MG, Abdur M. A comparison of simulation analysis of safety systems and severe accident progression for VVER-1200 and AP-1000 following LOCA in hot leg by PCTRAN. International Conference on Mechanical Engineering and Renewable Energy Chattogram, Bangladesh, 2021.
  • [15] Hassan MS, Sharif I, Khan AH, Sikder D. Evaluation of safety for VVER-1200-based nuclear power plants during main steam-line break accidents. International Journal of Nuclear Energy Science and Technology 2025; 18(2).
  • [16] Hossen MM. Analysis of thermal-hydraulic parameters during steam generator tube rupture event of VVER-1200 NPP using PCTRAN simulator. Applications of Modelling and Simulation 2022; 6: 28–35.
  • [17] Hossen MM, Khalaquzzaman M, Rahman SKA. Analysis of steam line break accident using PCTRAN model of VVER-1200 NPP. Applications of Modelling and Simulation 2023; 7: 8–15.
  • [18] Tanim MMH, Ali MF, Shobug MA, Abedin S. Analysis of various types of possible fault and consequences in VVER-1200 using PCTRAN. 2020 International Conference for Emerging Technology (INCET) Belgaum, India, 2020.
  • [19] Saha A, Fyza N, Hossain A, Sarkar MAR. Simulation of tube rupture in steam generator and transient analysis of VVER-1200 using PCTRAN. Energy Procedia 2019; 160: 162-169.
  • [20] Karim MSS, Datta D, Hossain A. Safety and security analysis of VVER-1200 reactor pressure vessel under pressurized thermal shock. International Journal of Nuclear Security 2024; 9(2).
  • [21] Korotkikh AG, Odii CJ. Analysis of VVER-1200 thermal characteristics using analytical calculation validated by PCTRAN simulator. Discover Mechanical Engineering 2025; 4(1): 79.
  • [22] Microsimulation Technology Inc. PCTRAN VVER-1200 user documentation. Available at: http://www.microsimtech.com/VVER1200/VVER1200d.html. Accessed 10 December 2025.
Toplam 22 adet kaynakça vardır.

Ayrıntılar

Birincil Dil İngilizce
Konular Enerji, Nükleer Enerji Sistemleri
Bölüm Araştırma Makalesi
Yazarlar

Güven Tunç 0000-0001-7038-8168

Tuna Han Ulu 0009-0002-6761-9193

Ömer Faruk Yılmaz 0009-0007-7166-7848

Gönderilme Tarihi 1 Şubat 2026
Kabul Tarihi 25 Şubat 2026
Yayımlanma Tarihi 17 Mart 2026
DOI https://doi.org/10.58559/ijes.1879510
IZ https://izlik.org/JA47MU93GA
Yayımlandığı Sayı Yıl 2026 Cilt: 11 Sayı: 1

Kaynak Göster

APA Tunç, G., Ulu, T. H., & Yılmaz, Ö. F. (2026). Early phase behavior of steam line break accidents in the VVER-1200 reactor: Effect of break location, AC power loss, and break size. International Journal of Energy Studies, 11(1), 143-160. https://doi.org/10.58559/ijes.1879510
AMA 1.Tunç G, Ulu TH, Yılmaz ÖF. Early phase behavior of steam line break accidents in the VVER-1200 reactor: Effect of break location, AC power loss, and break size. International Journal of Energy Studies. 2026;11(1):143-160. doi:10.58559/ijes.1879510
Chicago Tunç, Güven, Tuna Han Ulu, ve Ömer Faruk Yılmaz. 2026. “Early phase behavior of steam line break accidents in the VVER-1200 reactor: Effect of break location, AC power loss, and break size”. International Journal of Energy Studies 11 (1): 143-60. https://doi.org/10.58559/ijes.1879510.
EndNote Tunç G, Ulu TH, Yılmaz ÖF (01 Mart 2026) Early phase behavior of steam line break accidents in the VVER-1200 reactor: Effect of break location, AC power loss, and break size. International Journal of Energy Studies 11 1 143–160.
IEEE [1]G. Tunç, T. H. Ulu, ve Ö. F. Yılmaz, “Early phase behavior of steam line break accidents in the VVER-1200 reactor: Effect of break location, AC power loss, and break size”, International Journal of Energy Studies, c. 11, sy 1, ss. 143–160, Mar. 2026, doi: 10.58559/ijes.1879510.
ISNAD Tunç, Güven - Ulu, Tuna Han - Yılmaz, Ömer Faruk. “Early phase behavior of steam line break accidents in the VVER-1200 reactor: Effect of break location, AC power loss, and break size”. International Journal of Energy Studies 11/1 (01 Mart 2026): 143-160. https://doi.org/10.58559/ijes.1879510.
JAMA 1.Tunç G, Ulu TH, Yılmaz ÖF. Early phase behavior of steam line break accidents in the VVER-1200 reactor: Effect of break location, AC power loss, and break size. International Journal of Energy Studies. 2026;11:143–160.
MLA Tunç, Güven, vd. “Early phase behavior of steam line break accidents in the VVER-1200 reactor: Effect of break location, AC power loss, and break size”. International Journal of Energy Studies, c. 11, sy 1, Mart 2026, ss. 143-60, doi:10.58559/ijes.1879510.
Vancouver 1.Güven Tunç, Tuna Han Ulu, Ömer Faruk Yılmaz. Early phase behavior of steam line break accidents in the VVER-1200 reactor: Effect of break location, AC power loss, and break size. International Journal of Energy Studies. 01 Mart 2026;11(1):143-60. doi:10.58559/ijes.1879510