Nuclear Engineering and Technology

Korean Nuclear Society (한국원자력학회)
권호 표지

Post Test Analysis to Natural Circulation Experiment on the BETHSY Facility Using the MARS 1.4 Code

  • Published : 2001.12.01
Download PDF
⟨ Previous Next ⟩

Abstract

The present study is to assess the applicability of the best-estimate thermal-hydraulic code, MARS 1.4, for the analysis of thermal-hydraulic behavior in PWRs during natural circulation conditions. The code simulates a natural circulation test, BETHSY test 4. la, which was conducted on the integral test facility of BETHSY. The test represented the cooling states of the primary cooling system under single-phase natural circulation, two-phase natural circulation and the reflux condensation mode with conditions corresponding to the residual power, 2% of the rated core power value and 6.8 MPa at the secondary system. Based on MARS 1.4 calculations, the major thermal-hydraulic behaviors during natural circulation are evaluated and the differences between the experimental data and calculated results are identified. The calculated results show generally good behavior with regard to the experimental results; the region of single-phase natural circulation is 100-92% of the initial mass inventory, two-phase natural circulation is 84-63 %, and the reflux condensation mode occurred below 58 %. U-tubes empty and the core uncovery are obtained at 39 % and 34 % of the initial mass inventory, respectively.

Keywords

References

  1. K.H. Bae, H.S. Lim, J.H. Song, S.K. Sim, and J.G. Park, 'Design options for the safety injection sytstem of Korean Next Generation Reactor', Annals of Nuclear Energy, 27, 1011 (2000) https://doi.org/10.1016/S0306-4549(99)00100-0
  2. R. B. Duffey and J. P. Sursock, 'Natural circulation phenomena relevant to small breaks and transients,' Nuclear Engineering and Design, 102, 115 (1987) https://doi.org/10.1016/0029-5493(87)90244-5
  3. K.H. Bea, Y.J. Chung, M.H. Chang and S.K. Sim, 'Assessment of SMART advanced passive safety design,' Proceedings of NTHAS98(Nuclear thermal hydraulics and safety), 169 (1998)
  4. O.B. Samoilov, V.S. Kuul, V.A. Malamud and G.I. Tarasov, 'Integral nuclear power reactor with natural coolant circulation. Investigation of passive RHR system,' Nuclear Engineering and Design, 165, 259 (1996) https://doi.org/10.1016/0029-5493(96)01188-0
  5. Y. Zvirin, P.R. Jeuck III, C.W. Sullivan and R.B. Duffey, ' Experimental and analytical investigation of a natural circulation system with parallel loops,' J. Heat Transfer, 103, 645 (1981)
  6. G. G. Loomis and K. Soda, 'Results of the Semiscale Mod-2A natural circulation experiments,' NRC Report NUREG/CR-2335, (1982)
  7. L.E. Hochreiter, E.R. Rosal, S.D. Rupprecht, B.R. Sinwell, J.T. Dederer, K.J. Kovdalski, S. Wong, and R. Quaglia, 'PWR FLECHT SEASET systems effects natural circulation and reflux condensation data evaluation and analysis report,' NRC Report NUREG/CR-3654, (1985)
  8. Eckhard Krepper and Horst-michael Prasser, 'Natural circulation experiments at the ISBVVER integral test facility and calculations using the thermal-hydraulic code ATHLET,' Nuclear Technology, 128, 75 (1999)
  9. Yong-Seok Son, Won-Seok Kim, Kyung-Doo Kim, Yong-Jong Chung and Won-Pyo Chang, 'Analsis of reflux cooling in the SG utubesunder loss of RHRS during midloop operation with primary system partly open,' J. of the Korean Nuclear Society, 30(2), 112 (1998)
  10. K. Tasaka, 'Effect of SG secondary inventory on reflux condensation,' ASME winter annual meeting, Anaheim, CA, (1986)
  11. H. C. No, Y. M. MOON and H. S. Park, 'Assessment of RELAP5/MOD3.2 for Reflux Condensation Experiment,' NUREG/IA-0181 (2000)
  12. P. Bazin, 'BETHSY: Data base,' SETh/LES/87-28, Centre D' etudes Nucleaires de Grenoble, (1987)
  13. P. Bazin, 'BETHSY test 4.1a: two-phase natural circulation-test analysis report,' SETh/LES/89-86, Centre D' etudes Nucleaires de Grenoble, (1990)
  14. W. J. Lee, 'Improved features of MARS 1.4 and verification,' KAERI/TR-1386-99, (1999)
  15. J.-J. Jeong, K. S. Ha, B. D. Chung and W. J. Lee, 'Development of a multi-dimensional thermal-hydraulic system code, MARS 1.3.1,' Annals of Nuclear Energy, 26, 1611 (1999) https://doi.org/10.1016/S0306-4549(99)00039-0
  16. RELAP5 development team, 'RELAP5/MOD3 Code manual,' INEL-95/0174/NERGE/CR-5535, (1995)
  17. C. L. Wheeler, 'An interim version of COBRA for Thermal-hydraulic analysis of rod bundle nuclear fuel elements and cores,' BNWL-1962, (1976)
  18. Juei-Tsuen Hsu, Mamoru Ishii and Takashi Hibiki, 'Experimental study on two-phase natural circulation and flow termination in a loop,' Nuclear Engineering and Design, 186, 395 (1998) https://doi.org/10.1016/S0029-5493(98)00285-4