DOI QR코드

DOI QR Code

Development and validation of a fast sub-channel code for LWR multi-physics analyses

  • Chaudri, Khurrum Saleem (Department of Nuclear Engineering, Pakistan Institute of Engineering and Applied Science (PIEAS)) ;
  • Kim, Jaeha (Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology (KAIST)) ;
  • Kim, Yonghee (Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology (KAIST))
  • Received : 2019.01.03
  • Accepted : 2019.02.22
  • Published : 2019.06.25

Abstract

A sub-channel solver, named ${\underline{S}}teady$ and ${\underline{T}}ransient$ ${\underline{A}}nalyzer$ for ${\underline{R}}eactor$ ${\underline{T}}hermal$ hydraulics (START), has been developed using the homogenous model for two-phase conditions of light water reactors. The code is developed as a fast and accurate TH-solver for coupled and multi-physics calculations. START has been validated against the NUPEC PWR Sub-channel and Bundle Test (PSBT) database. Tests like single-channel quality and void-fraction for steady state, outlet fluid temperature for steady state, rod-bundle quality and void-fraction for both steady state and transient conditions have been analyzed and compared with experimental values. Results reveal a good accuracy of solution for both steady state and transient scenarios. Axially different values for turbulent mixing coefficient are used based on different grid-spacer types. This provides better results as compared to using a single value of turbulent mixing coefficient. Code-to-code evaluation of PSBT results by the START code compares well with other industrial codes. The START code has been parallelized with the OpenMP algorithm and its numerical performance is evaluated with a large whole PWR core. Scaling study of START shows a good parallel performance.

Keywords

References

  1. D. Basile, M. Beghi, R. Chierici, E. Salina, E. Brega, COBRA-EN: an upgraded version of the COBRA-3C/MIT code for thermal hydraulic transient analysis of light water reactor fuel assemblies and cores, Radiat. Saf. Inf. Comput. Center, Oak Ridge Natl. Lab. (1999).
  2. U. Imke, V.H. Sanchez, Validation of the subchannel code SUBCHANFLOW using the NUPEC PWR tests (PSBT), Sci. Technol. Nucl. Install. 2012 (2012), https://doi.org/10.1155/2012/465059.
  3. D.-H. Hwang, S.-J. Kim, K.-W. Seo, H. Kwon, Accuracy and uncertainty analysis of PSBT benchmark exercises using a subchannel code MATRA, Sci. Technol. Nucl. Install. (2012) 1-13, https://doi.org/10.1155/2012/603752, 2012.
  4. J.E. Kelly, S.P. Kao, M.S. Kazimi, THERMIT-2 : a Two-Fluid Model for Light Water Reactor Subchannel Transient Analysis, Massachusetts Institute of Technology, Energy Laboratory, Cambridge, Mass, 1981, p. 1981. http://hdl.handle.net/1721.1/60460.
  5. M. Avramova, A. Velazquez-Lozada, A. Rubin, Comparative analysis of CTF and TRACE Thermal-Hydraulic codes using OECD/NRC PSBT benchmark void distribution database, Sci. Technol. Nucl. Install. 2013 (2013), https://doi.org/10.1155/2013/725687.
  6. P. Kral, J. Hyvarinen, A. Prosek, A. Guba, Sources and effect of non-condensable gases in reactor coolant system of LWR, in: 16th Int. Top. Meet. Nucl. React. Therm. NURETH-16, Chicago, 2015, pp. 5194-5208.
  7. F. Roelofs, V.R. Gopala, L. Chandra, M. Viellieber, A. Class, Simulating fuel assemblies with low resolution CFD approaches, Nucl. Eng. Des. 250 (2012) 548-559. https://doi.org/10.1016/j.nucengdes.2012.05.029.
  8. Y. Sung, R.L. Oelrich, C.C. Lee, N. Ruiz-Esquide, M. Gambetta, C.M. Mazufri, Benchmark of subchannel code VIPRE-W with PSBT void and temperature test data, Sci. Technol. Nucl. Install. (2012), 2012. Article ID 757498, 11 pages. https://doi.org/10.1155/2012/757498.
  9. K.H. Leung, D.R. Novog, Evaluation of ASSERT-PV V3R1 against the PSBT benchmark, Sci. Technol. Nucl. Install. (2012), 2012. Article ID 863503, 21 pages. https://doi.org/10.1155/2012/863503.
  10. M. Avramova, A. Rubin, H. Utsuno, Overview and discussion of the OECD/NRC benchmark based on NUPEC PWR subchannel and bundle tests, Sci. Technol. Nucl. Install. (2013) 1-20, https://doi.org/10.1155/2013/946173, 2013.
  11. J.W. Jackson, N.E. Todreas, COBRA IIIc/MIT-2: a Digital Computer Program for Steady State and Transient Thermal-Hydraulic Analysis of Rod Bundle Nuclear Fuel Elements. Final Report, Massachusetts Inst. Of Tech, Energy Lab., Cambridge (USA), 1981.
  12. A. Bennett, N. Martin, M. Avramova, K. Ivanov, Impact of Radial Void Fraction Distribution on Boiling Water Reactor Lattice Physics Calculations: Application to AREVA's Next Generation BWR Fuel As-Sembly, the $ATRIUM^{TM}$ 11 Design, 2016.
  13. N.E. Todreas, M.S. Kazimi, Nuclear Systems II: Elements of Thermal Hydraulic Design, Taylor & Francis, 1990.
  14. S.J. Yoon, S.B. Kim, G.C. Park, H.Y. Yoon, H.K. Cho, Application of CUPID for subchannel-scale thermal-hydraulic analysis of pressurized water reactor core under single-phase conditions, Nucl. Eng. Technol. 50 (2018) 54-67. https://doi.org/10.1016/j.net.2017.09.008.
  15. W. Wagner, H.-J. Kretzschmar, International Steam Tables-Properties of Water and Steam Based on the Industrial Formulation IAPWS-IF97: Tables, Algorithms, Diagrams, and CD-ROM Electronic Steam Tables-All of the Equations of IAPWS-IF97 Including a Complete Set of Supplementary Backward, Springer Science & Business Media, 2007.
  16. M.M. Awad, Y.S. Muzychka, Effective property models for homogeneous two-phase flows, Exp. Therm. Fluid Sci. 33 (2008) 106-113, https://doi.org/10.1016/j.expthermflusci.2008.07.006.
  17. K. Rehme, Pressure drop of spacer grids in smooth and roughened rod bundles, Nucl. Technol. 33 (1977) 314-317, https://doi.org/10.13182/NT77-A31793.
  18. T.S. Blyth, Development and Implementation of CFD-Informed Models for the Advanced Subchannel Code CTF, 2017.
  19. A. Rubin, M. Avramova, OECD/NRC Benchmark Based on NUPEC PWR Subchannel and Bundle Tests (PSBT). Volume II: Benchmark Results for the Void Distribution Phase, NEA/NSC/DOC, 2011.
  20. M. Bucci, P. Fillion, Analysis of the NUPEC PSBT tests with FLICA-OVAP, Sci. Technol. Nucl. Install. 2012 (2012), https://doi.org/10.1155/2012/436142.
  21. J. Kim, Y. Kim, Development of 3-D HCMFD algorithm for efficient pin-by-pin reactor analysis, Ann. Nucl. Energy 127 (2019) 87-98. https://doi.org/10.1016/j.anucene.2018.11.035.