ROSA/LSTF Separate Effect Test on Natural Circulation under High Core Power Condition of Pressurized Water Reactor
Commenced in January 2007
Frequency: Monthly
Edition: International
Paper Count: 32845
ROSA/LSTF Separate Effect Test on Natural Circulation under High Core Power Condition of Pressurized Water Reactor

Authors: Takeshi Takeda


A separate effect test (SET) simulated natural circulation (NC) under high core power condition of a pressurized water reactor (PWR) utilizing the ROSA/LSTF (rig of safety assessment/large-scale test facility). The LSTF test results clarified the relationship between the primary loop mass inventory and the primary loop mass flow rate being dependent on the NC mode at a constant core power of 8% of the volumetric-scaled PWR nominal power. When the core power was 9% or more during reflux condensation, large-amplitude level oscillation in a form of slow fill and dump occurred in steam generator (SG) U-tubes. At 11% core power during reflux condensation, intermittent rise took place in the cladding surface temperature of simulated fuel rods. The RELAP5/MOD3.3 code indicated the insufficient prediction of the SG U-tube liquid level behavior during reflux condensation.

Keywords: LSTF, natural circulation, core power, RELAP5.

Digital Object Identifier (DOI):

Procedia APA BibTeX Chicago EndNote Harvard JSON MLA RIS XML ISO 690 PDF Downloads 768


[1] H. Shiroyama, “Regulatory failures of nuclear safety in Japan –the case of Fukushima accident,” in: Proc. of the Earth System Governance Tokyo Conference: Complex Architectures, Multiple Agents, Earth System Governance, Tokyo, Japan, January 2013.
[2] T. Yonomoto, Y. Anoda, Y. Kukita, and Y. Peng, “CCFL characteristicsof PWR steam generator U-tubes,” in: Proc. of the ANS International Topical Meeting on Safety of Thermal Reactors, American Nuclear Society, Portland, Ore, USA, July 1991.
[3] S. Al Issa and R. Macian-Juan, “A review of CCFL phenomenon,” Ann. Nucl. Energy, vol. 38, 2011, pp. 1795–1819.
[4] T. Kusunoki, T. Nozue, K. Hayashi, S. Hosokawa, A. Tomiyama, and M. Murase, “Condensation experiments for counter-current flow limitation in an inverted U-tube,” J. Nucl. Sci. Technol., vol. 53, 2016, pp. 486–495.
[5] 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.
[6] T. Takeda, H. Asaka, and H. Nakamura, “Analysis of the OECD/NEA ROSA project experiment simulating a PWR small break LOCA with high-power natural circulation,” Ann. Nucl. Energy, vol. 36, 2009, pp. 386–392.
[7] T. Takeda, “Uncertainty analysis of ROSA/LSTF test on pressurized water reactor cold leg small-break loss-of-coolant accident without scram,” Int. J. Nucl. Quantum Eng., vol. 13, 2019, pp. 82–90.
[8] T. Takeda, H. Asaka, and H. Nakamura, “RELAP5 analysis of OECD/NEA ROSA project experiment simulating a PWR loss-of-feedwater transient with high-power natural circulation,” Sci. Technol. Nucl. Installations, Article ID 957285, vol. 2012, 2012, pp. 1–15.
[9] G.G. Loomis and K. Soda, “Results of the Semiscale Mod-2A natural circulation experiments,” USNRC Report NUREG/CR-2335, EGG-2200, Idaho National Engineering Laboratory, 1982.
[10] F. D’Auria and G.M. Galassi, “Characterization of instabilities during two-phase natural circulation in PWR typical conditions,” Exper. Therm. Fluid Sci., vol. 3, 1990, pp. 641–650.
[11] P. Basin, R. Deruaz, T. Yonomoto, and Y. Kukita, “BETHSY/LSTF counterpart test on natural circulation in a pressurized water reactor,” in: Proc. of the 1992 National Heat Transfer Conference, San Diego, CA, USA, August 1992.
[12] Y.M. Ferng and C.H. Lee, “Numerical simulation of natural circulation experiments conducted at the IIST facility,” Nucl. Eng. Des., vol. 148, 1994, pp. 119–128.
[13] M. Cherubini, W. Giannotti, D. Araneo, and F. D’Auria, “Use of the natural circulation flow map for natural circulation systems evaluation,” Sci. Technol. Nucl. Installations, Article ID 479673, vol. 2008, 2008, pp. 1–7.
[14] A. Del Nevo, F. D’Auria, M. Mazzini, M. Bykov, I.V. Elkin, and A. Suslov, “The design of PSB-VVER experiments relevant to accident management,” J. Power Energy Syst., vol. 2, 2008, pp. 371–385.
[15] 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.
[16] V. Kouhia, V. Riikonen, O.P. Kauppinen, et al., “Benchmark exercise on SBLOCA experiment of PWR PACTEL facility,” Ann. Nucl. Energy, vol. 59, 2013, pp. 149–156.
[17] J. Kim, K.Y. Choi, K.H. Kang, Y. Park, B.U. Bae, and C.H. Song, “Experimental study for natural circulation flow regime map of ATLAS,” Transaction of the Korean Nuclear Society Spring Meeting, Jeju, Korea, May 2014.
[18] Y. Kukita, H. Nakamura, K. Tasaka, and C. Chauliac, “Nonuniform steam generator U-tube flow distribution during natural circulation tests in ROSA-IV large scale test facility,” Nucl. Sci. Eng., vol. 99, 1988, pp. 289–298.
[19] K. Tasaka, Y. Kukita, Y. Koizumi, M. Osakabe, and H. Nakamura, “The results of 5% small-break LOCA tests and natural circulation tests at the ROSA-IV LSTF,” Nucl. Eng. Des., vol. 108, 1988, pp. 37–44.
[20] F. D’Auria and M. Frogheri, “Use of a natural circulation map for assessing PWR performance,” Nucl. Eng. Des., vol. 215, 2002, pp. 111–126.
[21] N. Aksan, F. D’Auria, H. Glaeser, R. Pochard, C. Richards, and A. Sjoberg, “Separate Effects Test Matrix for Thermal-Hydraulic Code Validation – Volume I: Phenomena Characterisation and Selection of Facilities and Tests,” CSNI report OECD/GD(94)82, 1994.
[22] N. Aksan, F. D’Auria, H. Glaeser, J. Lillington, R. Pochard, and A. Sjoberg, “Evaluation of the CSNI Separate Effects Tests (SET) Validation Matrix,” CSNI report OECD/GD(97)9, 1996.
[23] USNRC Nuclear Safety Analysis Division, “RELAP5/MOD3.3 Code Manual,” NUREG/CR-5535/Rev 1, Information Systems Laboratories, Inc., 2001.
[24] D. Bestion, F. D’Auria, P. Lien, and H. Nakamura, “A state-of-the-art report on scaling in system thermal-hydraulics applications to nuclear reactor safety,” NEA/CSNI/R(2016)14, 2017.
[25] N. Zuber, “Problems in Modeling Small Break LOCA,” NUREG-0724, USNRC, Washington, DC, 1980.
[26] G.B. Wallis, “One-Dimensional Two-Phase Flow,” McGraw-Hill Book, New York, USA, 1969.