• Lee, Yeon-Gun (Institute for Nuclear Science and Technology, Jeju National University) ;
  • Park, Il-Woong (Department of Nuclear Engineering, Seoul National University) ;
  • Park, Goon-Cherl (Department of Nuclear Engineering, Seoul National University)
  • Received : 2013.03.12
  • Accepted : 2013.05.27
  • Published : 2013.06.25


This paper presents an experimental investigation of the small-break loss-of-coolant accident (SBLOCA) and the loss-of-feedwater accident (LOFW) in a scaled integral test facility of REX-10. REX-10 is a small integral-type PWR in which the coolant flow is driven by natural circulation, and the RCS is pressurized by the steam-gas pressurizer. The postulated accidents of REX-10 include the system depressurization initiated by the break of a nitrogen injection line connected to the steam-gas pressurizer and the complete loss of normal feedwater flow by the malfunction of control systems. The integral effect tests on SBLOCA and LOFW are conducted at the REX-10 Test Facility (RTF), a full-height full-pressure facility with reduced power by 1/50. The SBLOCA experiment is initiated by opening a flow passage out of the pressurizer vessel, and the LOFW experiment begins with the termination of the feedwater supply into the helical-coil steam generator. The experimental results reveal that the RTF can assure sufficient cooldown capability with the simulated PRHRS flow during these DBAs. In particular, the RTF exhibits faster pressurization during the LOFW test when employing the steam-gas pressurizer than the steam pressurizer. This experimental study can provide unique data to validate the thermal-hydraulic analysis code for REX-10.


Supported by : National Research Foundation of Korea (NRF)


  1. H. S. Park et al., "Experiments on the Performance Sensitivity of the Passive Residual Heat Removal System of an Advanced Integral Type Reactor," Nucl. Eng. Technol., 41, 53, 2009.
  2. Y. J. Chung, S. W. Lee, S. H. Kim, and K. K. Kim, "Passive Cooldown Performance of a 65MW Integral Reactor," Nucl. Eng. Des., 238, 1681 (2008)
  3. Y. G. Lee, J. W. Kim, and G. C. Park, "Experimental Study on Wall Condensation with Noncondensables in Steam-gas pressurizer," Proc. 13th Int. Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-13), Kanazawa city, Japan, Sep. 27 - Oct. 2 (2009)
  4. Y. G. Lee, J. W. Kim, and G. C. Park, "Development of a Thermal-hydraulic System Code, TAPINS, for 10 MW Regional Energy Reactor," Nucl. Eng. Des., 249, 364 (2012)
  5. S. M. Modro, P. North, and T. H. Chen, "LOFT Small Break Experiments," Nucl. Eng. Des., 102, 143 (1987)
  6. D. Juhel and G. Briday, "BETHSY Test 6.2 TC 6-inch Cold Leg Side Break Comparative Test," Note SETh/LES/90-112, CEA (1990)
  7. G. G. Loomis and J. E. Streit, "Quick Look Report for Semiscale Mod-2C Experiments S-LH-1 and S-LH-2," EGG-SEMI-6884, EG&G Idaho, Inc. (1985)
  8. K. Umminger and A. D. Nevo, "Integral Test Facilities and Thermal-Hydraulic System Codes in Nuclear Safety Analysis," Science and Technology of Nuclear Installations, 2012, Article ID 826732 (2012)
  9. K. Y. Choi et al., "Parametric Studies on Thermal Hydraulic Characteristics for Transient Operations of an Integral Reactor," Nucl. Eng. Technol., 38, 185 (2006)
  10. J. N. Reyes et al., "Testing of the Multi-application Small Light Water Reactor (MASLWR) Passive Safety Systems," Nucl. Eng. Des., 237, 1999 (2007)
  11. H. S. Park et al., "An Integral Effect Test Facility of the SMART, SMART-ITL," Trans. Korean Nuclear Society Autumn Meeting, Gyeingju, Korea, Oct. 25 - 16 (2012)
  12. B. I. Jang, M. H. Kim, and G. D. Jeun, "Transient Analysis of Natural Circulation Nuclear Reactor REX-10," J. Nucl. Sci. Technol., 48, 1046 (2011)
  13. B. I. Jang, M. H. Kim, and G. D. Jeun, "Experimental and Computational Investigation of a Natural Circulation System in Regional Energy Reactor-10MWth," Nucl. Eng. Des., 241, 2214 (2011)
  14. G. Kocamustafaogullari and M. Ishii, "Scaling of Two-phase Flow Transients Using Reduced Pressure System and Simulant Fluid," Nucl. Eng. Des., 104, 121 (1987)
  15. M. Ishii et al., "The three-level scaling approach with application to the Purdue University Multi-dimensional Integral Test Assembly (PUMA)," Nucl. Eng. Des., 186, 177 (1998)
  16. American Society of Mechanical Engineers, "Rules for Construction of Nuclear Power Plant Components," ASME Boilers and Pressure Vessel Code: An American National Standard, Section III, ASME, New York (1986)
  17. Y. G. Lee, "Development of TAPINS Code for Thermal-hydraulic Analysis of Integral Pressurized Water Reactor, REX-10," Ph. D. thesis, Seoul National University (2013)

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