Nuclear Engineering and Technology

  • 제53권6호
  • /
  • Pages.1769-1785
  • /
  • 2021
  • /
  • 1738-5733(pISSN)
  • /
  • 2234-358X(eISSN)
한국원자력학회 (Korean Nuclear Society)
권호 표지
DOI QR코드

DOI QR Code

Hot and average fuel sub-channel thermal hydraulic study in a generation III+ IPWR based on neutronic simulation

  • 투고 : 2019.12.19
  • 심사 : 2020.11.27
  • 발행 : 2021.06.25
PDF 다운로드
⟨ 이전 논문 다음 논문 ⟩

초록

The Integral Pressurized Water Reactors (IPWRs) as the innovative advanced and generation-III + reactors are under study and developments in a lot of countries. This paper is aimed at the thermal hydraulic study of the hot and average fuel sub-channel in a Generation III + IPWR by loose external coupling to the neutronic simulation. The power produced in fuel pins is calculated by the neutronic simulation via MCNPX2.6 then fuel and coolant temperature changes along fuel sub-channels evaluated by computational fluid dynamic thermal hydraulic calculation through an iterative coupling. The relative power densities along the fuel pin in hot and average fuel sub-channel are calculated in sixteen equal divisions. The highest centerline temperature of the hottest and the average fuel pin are calculated as 633 K (359.85 ℃) and 596 K (322.85 ℃), respectively. The coolant enters the sub-channel with a temperature of 557.15 K (284 ℃) and leaves the hot sub-channel and the average sub-channel with a temperature of 596 K (322.85 ℃) and 579 K (305.85 ℃), respectively. It is shown that the spacer grids result in the enhancement of turbulence kinetic energy, convection heat transfer coefficient along the fuel sub-channels so that there is an increase in heat transfer coefficient about 40%. The local fuel pin temperature reduction in the place and downstream the space grids due to heat transfer coefficient enhancement is depicted via a graph through six iterations of neutronic and thermal hydraulic coupling calculations. Working in a low fuel temperature and keeping a significant gap below the melting point of fuel, make the IPWR as a safe type of generation -III + nuclear reactor.

키워드

참고문헌

  1. International Atomic Energy Agency (IAEA), Design Features to Achieve Defence in Depth in Small and Medium Sized Reactors, IAEA Nuclear Energy Series., 2009. No. NP-T-2.2.
  2. D.T. Ingersoll, Deliberately small reactors and the second nuclear era, Prog. Nucl. Energy 51 (2009) 589-603. https://doi.org/10.1016/j.pnucene.2009.01.003
  3. NEA, Current Status, Technical Feasibility, and Economics of Small Nuclear Reactors, Nuclear Energy Agency, OECD, Paris, 2011.
  4. J. Vujic, et al., Small modular reactors: simpler, safer, cheaper, Energy 45 (2012) 288-295. https://doi.org/10.1016/j.energy.2012.01.078
  5. X. Yan, et al., Study of a nuclear energy supplied steelmaking system for near-term application, Energy 39 (2012) 154-165. https://doi.org/10.1016/j.energy.2012.01.047
  6. W.K. In, et al., Assessment of core protection and monitoring systems for an advanced reactor SMART, Ann. Nucl. Energy 29 (2002) 609-662. https://doi.org/10.1016/S0306-4549(01)00058-5
  7. G.C. Lee, W.P. Baek, S.H. Chang, Improved methodology for generation of axial flux shapes in digital core protection systems, Ann. Nucl. Energy 29 (2002) 805-819. https://doi.org/10.1016/S0306-4549(01)00076-7
  8. M.G. Na, et al., Estimation of the nuclear power peaking factor using in-core sensor signals, Nuclear Engineering and Technology 36 (2004) 420-429.
  9. ANSYS Version 15 User Guide, 2014.
  10. T. Hohne, et al., Application of CFD codes in nuclear reactor safety analysis, Science and Technology of Nuclear Installations 2010 (2010) 1-8.
  11. H. Sharifian, et al., Study of PWR transients by coupling of ANSYS-CFX with a kinetic model, Radiation Physics and Engineering 2 (2018) 3-21.
  12. D.B. Pelowitz, et al., MCNPXTM User's Manual Version 2.6.0, LOS ALAMOS NATIONAL LABORATORY, 2008.
  13. J.F. Briesmeister, et al., MCNP- A General Monte Carlo N-Particle Transport Code Version 4C, LA-13709-M, Los Alamos National Laboratory, April 2000.
  14. A. Sood, The Monte Carlo method and MCNP - a brief review of our 40 Year history, in: Topical Meeting on Industrial Radiation and Radioisotope Measurement - Applications Conference, Chicago, July, 2017.
  15. Y.P. Mahlers, VVER-1000 neutronics calculation with ENDF/B-VII data, Ann. Nucl. Energy 36 (2009) 1224-1229. https://doi.org/10.1016/j.anucene.2009.04.002
  16. B. Verboomen, et al., Monte Carlo calculation of the effective neutron generation time, Ann. Nucl. Energy 33 (2006) 911-916. https://doi.org/10.1016/j.anucene.2006.05.001
  17. Monte Carlo N-Particle Transport System User's Manual:C700, Los Alamos National laboratory, New Mexico, 2000.
  18. D. Delmastro, Thermal-hydraulic aspects of CAREM reactor, in: Proceedings of the IAEA TCM on Natural Circulation Data and Methods for Innovative Nuclear Power Plant Design, Vienna, Austria, 2000.
  19. S. Gomez, Development activities on advanced LWR designs in Argentina, in: Proceedings of the Technical Committee Meeting on Performance of Operating and Advanced Light Water Reactor Designs, Munich, Germany, 2000.
  20. M. Ishida, Development of new nuclear power plant in Argentina, in: Proceedings of the Advisory Group Meeting on Optimizing Technology, Safety and Economics of Water Cooled Reactors, Vienna, Austria, 2000.
  21. M. Ishida, Carem project development activities, in: Proceedings of the International Seminar on Status and Prospects for Small and Medium Size Reactors, Cairo, Egypt, 2001.
  22. H. Boado Magan, et al., CAREM project status, Science and Technology of Nuclear Installations 140373 (2011) 1-6.
  23. E. Villarino, D. Hergenreder, S. Matzkin, Neutronic core performance of CAREM-25 reactor, in: 27 Annual Meeting of the Argentine Association of Nuclear Technology, AANT), Argentina, 2000.
  24. International Atomic Energy Agency (IAEA), Design Features to Achieve Defence in Depth in Small and Medium Sized Reactors, IAEA Nuclear Energy Series., 2009. No. NP-T-2.2.
  25. Internationa AtomicEnergy Agency (IAEA), Thermo Physical Properties Database of Materials for Light Water Reactors and Heavy Water Reactors, IAEA., 2006. TECDOC-1496.
  26. W. Jens, P. lottes, Analysis of Heat Transfer, Burnout, Pressure Drop and Density Data for High Pressure Water, Argone National Laboratory, Chicago, 1951.
  27. F. Faghihi, et al., Neutronics and sub-channel thermal-hydraulics analysis of the IranianVVER-1000 fuel bundle, Prog. Nucl. Energy 87 (2016) 39-46. https://doi.org/10.1016/j.pnucene.2015.10.020