DOI QR코드

DOI QR Code

Transient full core analysis of PWR with multi-scale and multi-physics approach

  • Received : 2023.07.05
  • Accepted : 2023.12.01
  • Published : 2024.03.25

Abstract

Steam line break accident (SLB) in the nuclear reactor is one of the representative Non-LOCA accidents in which thermal-hydraulics and neutron kinetics are strongly coupled each other. Thus, the multi-scale and multi-physics approach is applied in this study in order to examine a realistic safety margin. An entire reactor coolant system is modelled by system scale node, whereas sub-channel scale resolution is applied for the region of interest such as the reactor core. Fuel performance code is extended to consider full core pin-wise fuel behaviour. The MARU platform is developed for easy integration of the codes to be coupled. An initial stage of the steam line break accident is simulated on the MARU platform. As cold coolant is injected from the cold leg into the reactor pressure vessel, the power increases due to the moderator feedback. Three-dimensional coolant and fuel behaviour are qualitatively visualized for easy comprehension. Moreover, quantitative investigation is added by focusing on the enhancement of safety margin by means of comparing the minimum departure from nucleate boiling ratio (MDNBR). Three factors contributing to the increase of the MDNBR are proposed: Various geometric parameters, realistic power distribution by neutron kinetics code, Radial coolant mixing including sub-channel physics model.

Keywords

Acknowledgement

This study was partly supported by the Nuclear Safety Research Program through the Korea Foundation of Nuclear Safety (KOFONS) under a grant from the Nuclear Safety and Security Commission (NSSC) (Grant No. 210621) and Korea Atomic Energy Research Institute (KAERI, 524510-23).

References

  1. R. Salko, M. Avramova, CTF Theory Manual, Pennsylvania State University, 2014.
  2. H. Kwon, et al., "Validation of a Subchannel Analysis Code MATRA Version 1.1," KAERI/TR-5581/2014, Korea Atomic Energy Research Institute, 2014.
  3. J.P. Turinsky, D.B. Kothe, Modeling and simulation challenges pursued by the consortium for advanced simulation of light water reactors (CASL), J. Comput. Phys. 313 (2016) 367.
  4. B. Kochunas, T. Downar, D. Jabaay, Validation and application of the 3D neutron transport MPACT code within CASL VERA-CS, Proc. NURETH-16 (2015). Chicago, IL, August 30-September 4.
  5. Y.S. Jung, C.B. Shim, C.H. Lim, H.G. Joo, Practical numerical reactor employing direct whole core neutron transport and subchannel thermal/hydraulic solvers, Ann. Nucl. Energy 62 (2013) 357-374. https://doi.org/10.1016/j.anucene.2013.06.031
  6. J. Lee, A. Facchini, H.G. Joo, Development of a drift-flux model based core thermal-hydraulics code for efficient high-fidelity multiphysics calculation, Nucl. Eng. Technol. 51 (2019) 1487-1503. https://doi.org/10.1016/j.net.2019.04.002
  7. B.O. Cho, et al., MASTER-2.0: Multi-Purpose Analyzer for Static and Transient Effects of Reactors, KAERI/TR-1211/99, Korea Atomic Energy Research Institute, 1999.
  8. L. Holt, et al., Investigation of feedback on neutron kinetics and thermal hydraulicsfrom detailed online fuel behavior modeling during a boron dilutiontransient in a PWR with the two-way coupled code system DYN3DTRANSURANUS, Nucl. Eng. Des. 297 (2016) 32-43. https://doi.org/10.1016/j.nucengdes.2015.11.005
  9. M. Garcia, et al., A Serpent2-SUBCHANFLOW-TRANSURANUS coupling for pin-bypin depletion calculations in Light Water Reactors, Ann. Nucl. Energy 139 (2020), 107213.
  10. J. Magedanz, al el, High-fidelity multi-physics system TORT-TD/CTF/FRAPTRAN for light water reactor analysis, Ann. Nucl. Energy 84 (2015) 234-243. https://doi.org/10.1016/j.anucene.2015.01.033
  11. J. Yu, et al., MCS based neutronics/thermal-hydraulics/fuel-performance coupling with CTF and FRAPCON, Comput. Phys. Commun. 238 (2019) 1-18. https://doi.org/10.1016/j.cpc.2019.01.001
  12. H. Yoon, et al., A multiscale and multiphysics PWR safety analysis at a subchannel scale, Nucl. Sci. Eng. 194 (2020) 633-649. https://doi.org/10.1080/00295639.2020.1727698
  13. H.Y. Yoon, et al., Recent improvements in the CUPID code for a multi-dimensional two-phase flow analysis of nuclear reactor components, Nucl. Eng. Technol. 46 (5) (2014) 655-666. https://doi.org/10.5516/NET.02.2014.023
  14. Jeong, et al., Development of a multi-dimensional thermal-hydraulic system code, MARS 1.3.1, Ann. Nucl. Energy 26 (18) (1999) 1611-1642. https://doi.org/10.1016/S0306-4549(99)00039-0
  15. K.J. Geelhood, et al., FRAPTRAN-2.0: AComputer code for the transient analysis of oxide fuel rods, PNNL 1 (2016), 19400. Rev.2.
  16. Y.J. Cho, et al., Development of a three-dimensional two-phase flow analysis code for nuclear reactor thermal hydraulics: Part II. Application and validation, Proc. PHYTRA4. Marrakech, Morocco (2018). Sepp. 17-19.
  17. I.K. Park, et al., An implicit code coupling of 1-D system code and 3-D in-house CFD code for multi-scaled simulations of nuclear reactor transients, Ann. Nucl. Energy 59 (2013) 80-91. https://doi.org/10.1016/j.anucene.2013.03.048
  18. I.K. Park, et al., Multi-scale analysis of an ATLAS-MSLB test using the coupled CUPID/MARS code, Ann. Nucl. Energy 113 (2018) 332-343. https://doi.org/10.1016/j.anucene.2017.11.036
  19. J.R. Lee, H.Y. Yoon, Multi-physics simulation of nuclear reactor core by coupled simulation using CUPID/MASTER, Int. J. Heat Mass Tran. 115 (2017) 1020-1032. https://doi.org/10.1016/j.ijheatmasstransfer.2017.07.124
  20. H.K. Cho, Heat structure coupling of CUPID and MARS for the multi-scale simulation of the passive auxiliary feedwater system, Nucl. Eng. Des. 273 (2014) 459-468. https://doi.org/10.1016/j.nucengdes.2014.03.017
  21. S.W. Lee, et al., Coupled Calculation of SAPCE and FRAPTRAN, Proceedings of Korean Nuclear Society Spring Meeting, Jeju, Korea, 2018. May 17-18.
  22. H.C. Kim, et al., Development of fully coupled FRAPTRAN with MARS-KS code system for calculation of fuel behavior during LOCA, Proc. Topfuel (2018), 2018, A0015, Prague, Czech Republic, Sep. 30 - 04. Oct.
  23. D. Groeneveld, et al., AECL-UO critical heat flux lookup table, Heat Tran. Eng. 7 (1-2) (1986) 46-62, 1986. https://doi.org/10.1080/01457638608939644