Nuclear Engineering and Technology

Korean Nuclear Society (한국원자력학회)
권호 표지

Simulation of Multiple Steam Generator Tube Rupture (SGTR) Event Scenario

  • Published : 2003.06.01
Download PDF
⟨ Previous Next ⟩

Abstract

The multiple steam generator tube rupture (SGTR) event scenario with available safety systems was experimentally and analytically evaluated. The experiment was conducted on the large scaled test facility to simulate the multiple SGTR event and investigate the effectiveness of operator actions. As a result, it indicated that the opening of pressurizer power operated relief valve was significantly effective in quickly terminating the primary-to-secondary break flow even for the 6.5 tubes rupture. In the analysis, the recent version of RELAP5 code was assessed with the test data. It indicated that the calculations agreed well with the measured data and that the plant responses such as the water level and relief valve cycling in the damaged steam generator were reasonably predicted. Finally, sensitivity study on the number of ruptured tubes up to 10 tubes was performed to investigate the coolant release into atmosphere. It indicated that the integrated steam mass released was not significantly varied with the number of ruptured tubes although the damaged steam generator was overfilled for more than 3 tubes rupture. These findings are expected to provide useful information in understanding and evaluating the plant ability to mitigate the consequence of multiple SGTR event.

Keywords

References

  1. G.G. Loomis, 'Steam Generator Tube Rupture in an Experimental Facility Scaled from a Pressurized Water Reactor,' Proc. Int. Topl. Mtg., Thermal Nuclear Reactor Safety, Karlsruhe, Germany, September 10-13, (1984)
  2. B. Faydide, A. Messie, and E. Allen, 'Multiple Steam Generator Tube Rupture on the BETHSY Integral Test Facility,' Proc. Int. Topl. Mtg., Safety of Thermal Reactors, pp 186, Portland, Oregon, U.S.A, July 21-25, (1991)
  3. Y. Anoda, H. Nakamura, T. Watanabe, M. Hirano and Y. Kukita, 'Steam Generator Tube Rupture Simulations,' Proc. Int. Conf. New Trends in Nuclear System Thermalhydraulics, pp 539-545, Pisa, Italy, May 30-June 2, (1994)
  4. K.W. Seul, Y.S. Bang, and H.J. Kim, 'Evaluation of Thermal-Hydraulic Behavior on Break Simulation of Steam Generator U-Tube,' 2nd European Thermal Sciences and 14th UIT National Heat Transfer Conf., pp 975-982, Rome, Italy, May 29-31, (1996)
  5. T. Liu and C. Lee, 'An Evaluation of Emergency Operator Actions by an Experimental SGTR Event at the IIST Facility and a Comparison of Mihama-2 SGTR Event Record,' Nuclear Technology, Vol. 129, pp 36-50, January (2000)
  6. G.F. De Santi, 'Analysis of Steam Generator U-tube Rupture and Intentional Depressurization in LOBI-MOD2 Facility,' Nuclear Engineering and Design, Vol. 126, pp 113-125, February (1991)
  7. H. Nakamura, Y. Anoda and Y. Kukita, 'Steam Generator Multiple U-tube Rupture Experiments on ROSA-IV/LSTF,' Proc. 6th Int. Topl. Mtg., Nuclear Reactor Thermal Hydraulics (NURETH-6), pp 993-1001, Grenoble, France, October 5-8, (1993)
  8. J.C. Barbier, P. Clement, and R. Deruaz, 'A Single SGTR with Unavailability of Both the High Pressure Safety Injection System and the Steam Generator Auxiliary Feedwater System on BETHSY Integral Test Facility,' Proc. Int. Conf., New Trends in Nuclear System Thermalhydraulics, pp 553-559, Pisa, Italy, May 30-June 2, (1994)
  9. ROSA-IV Group - ROSA-IV Large Scale Test Facility (LSTF) System Description for Second Simulated Fuel Assembly, JAERI-M 90-176, Japan Atomic Energy Research Institute, October (1990)
  10. Agency of Natural Sources and Energy, 'Status of Investigation on the Steam Generator Tube Break at Mihama Unit No.2 of the Kansai Electric Power Co.,' Japan, (1991)
  11. Information Systems Laboratories, 'RELAP5/MOD3.3 Beta Code Manual: User's Guide and Input Requirements,' NUREG/CR-5535/Rev. 1, U. S. Nuclear Regulatory Commission, Vol. 1-5, May (2001)