Nuclear Engineering and Technology

Korean Nuclear Society (한국원자력학회)
권호 표지
DOI QR코드

DOI QR Code

Study on the effect of flow blockage due to rod deformation in QUENCH experiment

  • Gao, Pengcheng (School of Nuclear Science and Technology, Xi'an Jiaotong University) ;
  • Zhang, Bin (School of Nuclear Science and Technology, Xi'an Jiaotong University) ;
  • Shan, Jianqiang (School of Nuclear Science and Technology, Xi'an Jiaotong University)
  • Received : 2021.11.02
  • Accepted : 2022.03.03
  • Published : 2022.08.25
Download PDF
⟨ Previous Next ⟩

Abstract

During a loss-of-coolant accident (LOCA) in the pressurized water reactor (PWR), there is a possibility that high temperature and internal pressure of the fuel rods lead to ballooning of the cladding, which causes a partial blockage of flow area in a subchannel. Such flow blockage would influence the core coolant flow, thus affecting the core heat transfer during a reflooding phase and subsequent severe accident. However, most of the system analysis codes simulate the accident process based on the assumed channel blockage ratio, resulting in the fact that the simulation results are not consistent with the actual situation. This paper integrates the developed core Fuel Rod Thermal-Mechanical Behavior analysis (FRTMB) module into the self-developed severe accident analysis code ISAA. At the same time, the existing flow blockage model is improved to make it possible to simulate the change of flow distribution due to fuel rod deformation. Finally, the ISAA-FRTMB is used to simulate the QUENCH-LOCA-0 experiment to verify the correctness and effectiveness of the improved flow blockage model, and then the effect of clad ballooning on core heat transfer and subsequent parts of core degradation is analyzed.

Keywords

Acknowledgement

This study is supported by Innovative Scientific Program of CNNC.

References

  1. J. Conti, International Energy Outlook 2016, U.S. Energy Information Administration, 2016.
  2. H.K. Park, S.H. Kim, B.-J. Chung, Natural convection of melted core at the bottom of nuclear reactor vessel in a severe accident, Int. J. Energy Res. 42 (2018) 303-313, https://doi.org/10.1002/er.3893.
  3. Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors, 10 Code of Federal Register, Part 50.
  4. G. Repetto, C. Dominguez, B. Durville, et al., The R&D PERFROI project on Thermal mechanical and thermal hydraulics behaviors of a fuel rod assembly during a Loss of Coolant Accident, 2015. NURETH 1 (2015) 1-14, 2015.
  5. C. Grandjean, Coolability of blocked regions in a rod bundle after ballooning under LOCA conditions, Nucl. Eng. Des. 237 (2007) 15-17, 1872-1886. https://doi.org/10.1016/j.nucengdes.2007.02.022
  6. B. Jdpca, et al., Experimental thermal hydraulics study of the blockage ratio effect during the cooling of a vertical tube with an internal steam-droplets flow, Int. J. Heat Mass Tran. 140 (2019) 648-659, https://doi.org/10.1016/j.ijheatmasstransfer.2019.06.012.
  7. E. Williams, R. Martin, P. Gandrille, R. Meireles, R. Prior, C. Henry, Q. Zhou, Recent Revisions to MAAP4 for US EPR Severe Accident Applications, Conference ICAPP 08, Anaheim (California, USA), June 2008.
  8. R.O. Gauntt, J.E. Cash, R.K. Cole, et al., MELCOR Computer Code Manuals 1-2, Version 1.8.6, Sandia National Laboratories, U.S. Nuclear Regulatory Commission, 2005.
  9. P. Chatelard, N. Reinke, S. Arndt, S. Belon, L. Cantrel, L. Carenini, K. Chevalier-Jabet, F. Cousin, J. Eckel, F. Jacq, C. Marchetto, C. Mun, L. Piar, ASTEC V2 severe accident integral code main features, current V2.0 modelling status, perspectives, Nucl. Eng. Des. 272 (2014) 119-135, https://doi.org/10.1016/j.nucengdes.2013.06.040.
  10. RELAP5 Code Development Team, RELAP5/MOD3 Code Manual. NUREG/CR-5535. Idaho National Engineering Lab, US Nuclear Regulatory Commission, 1995.
  11. Y. Guo, G. Wang, D. Qian, et al., Thermal hydraulic analysis of loss of flow accident in the JRR-3M research reactor under the flow blockage transient, Ann. Nucl. Energy 118 (AUG) (2018) 147-153, https://doi.org/10.1016/j.anucene.2018.04.014.
  12. T. Lu, J. Shan, J. Gou, et al., Preliminary safety analysis on loss of flow accidents and external source transients for LBE cooled ADSR core, Prog. Nucl. Energy 88 (Apr) (2016) 134-146, https://doi.org/10.1016/j.pnucene.2016.01.001.
  13. Pengcheng Gao, Bin Zhang, et al., Development of mechanistic cladding rupture model for integrated severe accident code ISAA. Part I: module verification and application in CAP1400, Ann. Nucl. Energy 158 (2021) 108305, https://doi.org/10.1016/j.anucene.2021.108305, 2-3.
  14. B. Zhang, J. Deng, M. Jing, et al., A newly developed suppression pool model based on the ISAA code, Nucl. Sci. Eng.: J.Am. Nucl. Soc. (2021) 1-11, https://doi.org/10.1080/00295639.2020.1861862.
  15. Kenneth Geelhood, Walter Luscher, Ian Porter, Material Property Correlations: Comparisons between FRAPCON-4.0, FRAPTRAN-2.0, and MATPRO, 2015, https://doi.org/10.2172/1030897.
  16. E. Rolstad, et al., Measurements of the Length Changes of UO2 Fuel Pellets during Irradiation, Enlarged HPG Meeting on Computer Control and Fuel Research, 1974. June 4-7, 1974.
  17. K. Pettersson, M. Billone, et al., Nuclear Fuel Behaviour in Loss-Of-Coolant Accident (LOCA) Conditions, State-of-the-art Report, 2009, ISBN 978-92-64-99091-3.
  18. C. Grandjean, A State-Of-The-Art Review of Past Programs Devoted to Fuel Behaviour under LOCA Conditions - Part One: Clad Swelling and Rupture Assembly Flow Blockage, Technical Report, IRSN/DPAM/SEMCA, 2005.
  19. I.E. Idelchik, Handbook of Hydraulic Resistance, Fourth fourth ed., BHB, 2008. -13: 978-1567002515.
  20. J. Stuckert, M. Grosse, C. Rossger, et al., Experimental Results of the Commissioning Bundle Test QUENCH-L0 Performed in the Framework of the QUENCH-LOCA Program, 2015.
  21. Fernandez-Moguel, L. Preliminary Analysis of Air Ingress Experiment QUENCH-16 Using RELAP/SCDAPSim3.5 and MELCOR 1.8.6.
  22. Leticia Fernandez-Moguel, Jonathan Birchley, Analysis of QUENCH-10 and -16 air ingress, experiments with SCDAPSim3.5 53 (2013) 202-212, https://doi.org/10.1016/j.anucene.2012.08.030.
  23. A.D. Vasiliev, et al., Application of Thermal Hydraulic and Severe Accident Code SOCRAT/V3 to Bottom Water Reflood Experiment QUENCH-LOCA-0, Nuclear Engineering & Design, 2013, https://doi.org/10.1016/j.nucengdes.2012.11.021.
  24. T. Kaliatka, A. Kaliatka, V. Vileiniskis, Application of best estimate approach for modelling of QUENCH-03 and QUENCH-06 experiments, Nucl. Eng. Technol. 48 (2) (2016) 419-433, https://doi.org/10.1016/j.net.2015.12.011.
  25. I. Gomez-Garcia-Torano, V.H. Sanchez-Espinoza, R. Stieglitz, et al., Validation of ASTEC V2.1 based on the QUENCH-08 experiment, Nucl. Eng. Des. 314 (APR) (2017) 29-43, https://doi.org/10.1016/j.nucengdes.2016.12.039.