Commenced in January 2007
Frequency: Monthly
Edition: International
Paper Count: 30982
ROSA/LSTF Test on Pressurized Water Reactor Steam Generator Tube Rupture Accident Induced by Main Steam Line Break with Recovery Actions

Authors: Takeshi Takeda


An experiment was performed for the OECD/NEA ROSA-2 Project employing the ROSA/LSTF (rig of safety assessment/large-scale test facility), which simulated a steam generator tube rupture (SGTR) accident induced by main steam line break (MSLB) with operator recovery actions in a pressurized water reactor (PWR). The primary pressure decreased to the pressure level nearly-equal to the intact steam generator (SG) secondary-side pressure even with coolant injection from the high-pressure injection (HPI) system of emergency core cooling system (ECCS) into cold legs. Multi-dimensional coolant behavior appeared such as thermal stratification in both hot and cold legs in intact loop. The RELAP5/MOD3.3 code indicated the insufficient predictions of the primary pressure, the SGTR break flow rate, and the HPI flow rate, and failed to predict the fluid temperatures in the intact loop hot and cold legs. Results obtained from the comparison among three LSTF SGTR-related tests clarified that the thermal stratification occurs in the horizontal legs by different mechanisms.

Keywords: RELAP5, SGTR, LSTF, thermal stratification

Digital Object Identifier (DOI):

Procedia APA BibTeX Chicago EndNote Harvard JSON MLA RIS XML ISO 690 PDF Downloads 247


[1] G.G. Loomis, “Steam generator tube rupture in an experimental facility scaled from a pressurized water reactor,” in: Proc. of the 5th International Meeting on Thermal Nuclear Reactor Safety, Karlsruhe, Germany, October 1984.
[2] J.M. Rogers, “An analysis of Semiscale Mod-2C S-FS-1 steam line break test using RELAP5/MOD2,” NUREG/IA-0052, USNRC, Washington, DC, 1992.
[3] G. Srikantiah, “Methods for PWR transient analysis,” Nucl. Eng. Des., vol. 83, 1984, pp. 189–198.
[4] G.F. De Santi, “Analysis of steam generator U-tube rupture and intentional depressurization in LOBI-MOD2 facility,” Nucl. Eng. Des., vol. 126, 1991. pp. 113–125.
[5] C. Addabbo and A. Annunziato, “The LOBI integral system test facility experimental programme,” Sci. Technol. Nucl. Installations, vol. 2012, 2012, Article 238019, pp. 1–16.
[6] 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,” in: Proc. of the International Conference on New Trends in Nuclear System Thermalhydrulics, Pisa, Italy, May-June 1994.
[7] T.J. Liu, C.H. Lee, C.C. Yao, and S.C. Chiang, “An evaluation of emergency operator actions by an experimental SGTR event at the IIST facility and a comparison of Mihama-2 SGTR event record,” Nucl. Technol., vol. 129, 2000, pp. 36–50.
[8] K. Umminger, T. Mull, and B. Brand, “Integral effect tests in the PKL facility with international participation,” Nucl. Eng. Technol., vol. 41, 2009, pp. 765–774.
[9] K. Umminger, L. Dennhardt, S. Schollenberger, and B. Schoen, “Integral Test Facility PKL: Experimental PWR Accident Investigation,” Sci. Technol. Nucl. Installations, Article ID 891056, vol. 2012, 2012, pp. 1–16.
[10] K.Y. Choi, Y.S. Kim, C.H. Song, and W.P. Baek, “Major achievements and prospect of the ATLAS integral effect tests,” Sci. Technol. Nucl. Installations, vol. 2012, 2012, Article 375070, pp. 1–18.
[11] K.H. Kang, Y.S. Park, B.U. Bae, J.R. Kim, N.H. Choi, and K.Y. Choi, “Code assessment of ATLAS integral effect test simulating main steam-line break accident of an advanced pressurized water reactor,” J. Nucl. Sci. Technol., vol. 55, 2018, pp. 104–112.
[12] The ROSA-V Group, “ROSA-V Large Scale Test Facility (LSTF) System Description for the Third and Fourth Simulated Fuel Assemblies,” JAERI-Tech 2003-037, Japan Atomic Energy Research Institute, Ibaraki, Japan, 2003.
[13] T. Takeda, “ROSA/LSTF test and RELAP5 code analyses on PWR steam generator tube rupture accident with recovery actions,” Nucl. Eng. Technol., vol. 50, 2018, pp. 981–988.
[14] NEA, “Final Integration Report of Rig-of-safety Assessment (ROSA-2) Project - 2009–2012,” NEA/CSNI/R(2016)10, 2017.
[15] T. Watanabe, “Effects of ECCS on the cold-leg fluid temperature during SGTR accidents,” Int. J. Mech. Mechatronics Eng., vol. 9, 2015, pp. 1618–1622.
[16] G. Jimenez, C. Queral, M.J. Rebollo-Mena, J.C. Martínez-Murillo, and E. Lopez-Alonso, “Analysis of the operator action and the single failure criteria in a SGTR sequence using best estimate assumptions with TRACE 5.0,” Ann. Nucl. Energy, vol. 58, 2013, pp. 161–177.
[17] K.Y. Choi, K.H. Kang, and C.H. Song, “Recent achievement and future prospects of the ATLAS program,” Nucl. Eng. Des., vol. 354, 2019, Article 110168, pp. 1–10.
[18] D. Lucas, D. Bestion, E. Bodèle, et al., “An overview of the pressurized thermal shock issue in the context of the NURESIM Project,” Sci. Technol. Nucl. Installations, vol. 2009, 2009, Article 583259, pp. 1–13.
[19] USNRC Nuclear Safety Analysis Division, “RELAP5/MOD3.3 Code Manual,” NUREG/CR-5535/Rev 1, Information Systems Laboratories, Inc., 2001.
[20] S. Gallardo, A. Querol, M. Lorduy, and G. Verdu, “Assessment of TRACE 5.0 against ROSA-2 Test 5, Main Steam Line Break with Steam Generator Tube Rupture,” NUREG/IA-0505, Rev. 1, USNRC, Washington, DC, 2019.
[21] K.W. Seul, Y.S. Bang, I.G. Kim, T. Yonomoto, and Y. Anoda, “Simulation of multiple steam generator tube rupture (SGTR) event scenario,” J. Korean Nucl. Soc., vol. 35, 2003, pp. 179–190.
[22] E.F. Hicken, “Important thermohydraulic aspects during refilling and reflooding of an uncovered LWR core,” in: Proc. of a Seminar on the Results of the European Communities’ Indirect Action Research Programme on Safety of Thermal Water Reactors, Brussels, Belgium, October 1984.
[23] N. Zuber, “Problems in Modeling Small Break LOCA,” NUREG-0724, USNRC, Washington, DC, 1980.
[24] H. Kumamaru and K. Tasaka, “Recalculation of Simulated Post-scram Core Power Decay Curve for Use in ROSA-IV/LSTF Experiments on PWR Small-break LOCAs and Transients,” JAERI-M 90-142, Japan Atomic Energy Research Institute, Ibaraki, Japan, 1990.
[25] NEA, “Final Integration Report of OECD/NEA ROSA Project 2005–2009,” NEA/CSNI/R(2013)1, 2013.
[26] V.H. Ransom and J.A. Trapp, “The RELAP5 choked flow model and application to a large scale flow test,” in: Proc. of the ANS/ASME/NRC International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Saratoga Springs, New York, USA, October 1980.
[27] D.W. Sallet, “Thermal hydraulics of valves for nuclear applications,” Nucl. Sci. Eng., vol. 88, 1984, pp. 220–244.
[28] NEA, “Best Practice Guidelines for the Use of CFD in Nuclear Reactor Safety Applications - Revision,” NEA/CSNI/R(2014)11, 2015.