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Implementation of dynamic start-up test experimental data as a main part of the nuclear code validation procedure: Developed RELAP5 model for VVER-1000

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초록

The main purposes of start-up tests in nuclear power plants (NPPs) are to ensure safe and reliable operation, verify system functionality, comply with regulatory requirements, optimize performance, and establish a foundation for ongoing plant operation and maintenance. However, the start-up tests of NPPs also could be used as a main part of the nuclear code validation procedure for several reasons including: realistic simulation, comprehensive evaluation, detection of code limitations, validation of safety margins and confidence in code predictions. The main purpose of the current study is to define and assess the validation procedure based on actual start-up test data. In this regard, the developed RELAP5 model has been validated against the actual data of VVER-1000 plant during a dynamic start-up test. The results of this full-scale validation show a good agreement between the developed RELAP5 model results and actual plant data. Finally, by defining a step by step validation procedure, it has been recommended to use the start-up phase test data as a more robust validation process which allow for full-scale validation of the nuclear code by comparing its predictions with actual plant measurements and also other advantages which have been demonstrated in the current study.

Abstract

The main purposes of start-up tests in nuclear power plants (NPPs) are to ensure safe and reliable operation, verify system functionality, comply with regulatory requirements, optimize performance, and establish a foundation for ongoing plant operation and maintenance. However, the start-up tests of NPPs also could be used as a main part of the nuclear code validation procedure for several reasons including: realistic simulation, comprehensive evaluation, detection of code limitations, validation of safety margins and confidence in code predictions. The main purpose of the current study is to define and assess the validation procedure based on actual start-up test data. In this regard, the developed RELAP5 model has been validated against the actual data of VVER-1000 plant during a dynamic start-up test. The results of this full-scale validation show a good agreement between the developed RELAP5 model results and actual plant data. Finally, by defining a step by step validation procedure, it has been recommended to use the start-up phase test data as a more robust validation process which allow for full-scale validation of the nuclear code by comparing its predictions with actual plant measurements and also other advantages which have been demonstrated in the current study.

참고문헌 (33)

  1. R. Akbari, "The longevity of light water-cooled large-scale nuclear reactors : a holistic perspective", Nucl. Eng. Des, 417, pp.112830~ (2024)
  2. IAEA Nuclear Energy Series, Commissioning of Nuclear Power Plants: Training and Human Resource Considerations, , IAEA, ~, pp.NG-T-2.2 (2008)
  3. M. A. Nasr, "Neutronic and fuel cycle performance analysis of fluoride and chloride fuels in Molten Salt Fast Reactor(MSFR)", Nucl. Eng. Des, 413, pp.112506~ (2023)
  4. R. Akbari, "Small modular reactor full scope core optimization using Cuckoo Optimization Algorithm", Prog. Nucl. Energy, 122, pp.103271~ (2020)
  5. R. Akbari, "Analysis of thorium-transuranic fuel deployment in a LW-SMR : A solution toward sustainable fuel supply for the future plants", Nuclear Eng. Desig, 421, pp.113090~ (2024)
  6. N. Vahman, "An assessment of a VVER-1000 core during Turbo-Generator load reduction test using RELAP5", MOD3.2 and WIMSD-5B, Prog. Nucl. Energy, pp.2016, 93[DOI]
  7. S. A. Hosseini, "Small modular reactors licensing process based on BEPU approach : status and perspective", Sustainability, 15(8), pp.6636~ (2023)
  8. R. Akbari-Jeyhouni, "Assessment of an integral small modular reactor during rod ejection accident by using DRAGON", PARCS codes, 2018, pp.108, 136~143
  9. A. K. Trivedi, "RELAP5", SCDAPSIM model development for AP1000 and verification for large break LOCA, 2016, pp.305, 222~229
  10. A. K. Trivedi, "AP1000 station blackout study with and without depressurization using RELAP5", SCDAPSIM, 2016, pp.307, 299~308
  11. L. Fernandez-Moguel, "Analysis of the accident in the Fukushima Daiichi nuclear power station Unit 3 with MELCOR_2. 1", Ann. Nucl. Energy, 83, pp.193~215 (2015)
  12. O. Noori-Kalkhoran, "Simulation of rod ejection accident in a WWER-1000 Nuclear Reactor by using PARCS code", Ann. Nucl. Energy, 65, pp.132~140 (2014)
  13. P. Tusheva, "Assessment of accident management measures on early in-vessel station blackout sequence at VVER-1000 pressurized water reactors", Nucl. Eng. Des, 277, pp.106~116 (2014)
  14. R. Berar, "RELAP5 and TRACE assessment of the Achilles natural reflood experiment", Nucl. Eng. Des, 261, pp.306~316 (2013)
  15. S. Chatzidakis, "Evaluation of RELAP5", MOD3behavior against loss of flow experimental results from two research reactor facilities, 2013, pp.255, 321~329
  16. D. Grishchenko, "The TALL-3D facility design and commissioning tests for validation of coupled STH and CFD codes", Nucl. Eng. Des, 290, pp.144~153 (2015)
  17. A. Agung, "Validation of PARCS", RELAP5coupled codes against a load rejection transient at the Ringhals-3 NPP, 2013, pp.257, 31~44
  18. J. R. Wang, "TRACE modeling and its verification using Maanshan PWR start-up tests", Ann. Nucl. Energy, 36, pp.527~536 (2009)
  19. S. F. Corzo, "Loss of flow accident analysis in Atucha II nuclear power plant using RELAP5 model", Nucl. Eng. Des, 402, pp.112108~ (2023)
  20. S. A. Hosseini, "Analysis of the natural circulation flow map uncertainties in an integral small modular reactor", Nucl. Eng. Des, 378, pp.111156~ (2021)
  21. R. V. Gencheva, "Investigation of station blackout scenario in VVER440", v230 with RELAP5 computer code, 2015, pp.295, 441~456
  22. P. Groudev, "Investigation of main coolant pump trip problem in case of SB LOCA for Kozloduy Nuclear Power Plant, WWER-440", V230, 2015, pp.76, 137~145
  23. M. Pavlova, "Steam Line Break investigation at full power reactor for VVER-1000", V320, 2015, pp.285, 65~74
  24. A. Hadjam, "Simulation of a LBLOCA in the CALLISTO test facility using the best estimate computer code RELAP5", SCDAP3. 2, 2013, pp.262, 153~167
  25. J. R. Jonnet, "RELAP5 analysis of PKL, main steam line break test", Nucl. Eng. Des, 265, pp.755~764 (2013)
  26. R. Vaghetto, "Study of debris-generated core blockage scenarios during loss of coolant accidents using RELAP5-3D", Nucl. Eng. Des, 261, pp.144~145 (2013)
  27. D. Lioce, "AP1000passive core cooling system pre-operational tests procedure definition and simulation by means of Relap5 Mod. 3. 3 computer code", Nucl. Eng. Des, 250, pp.538~547 (2012)
  28. M. Kristof, "RELAP5 simulation of surge line break accident using combined and best estimate plus uncertainty approaches", Nucl. Eng. Des, 239, pp.2500~2513 (2009)
  29. Final Safety Analysis Report (FSAR) for Bushehr VVER-1000 Reactor, "AEOI", ~ (2007)
  30. S. Kordalivand, "Quantifying the impact of risk mitigation measures using SPAR-H and RCM Approaches : Case study based on VVER-1000systems", Nuclear Eng. Desig, 423, pp.113174~ (2024)
  31. AEOI, "Album of neutron and physical characteristics of the initial fuel inventory", Organization of Operation on BNPP-1, ~ (2006)
  32. RELAP5, "MOD3 code manual, user’s guidelines", C.D. Fletcher, 1995, NUREG[DOI]
  33. N. Vahman, "Investigation of small break LOCA with failure of the emergency core cooling system accident management in VVER-1000 reactor type", Ann. Nucl. Energy, 183, pp.109632~ (2023)

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