References
- H. Shimamune et al., “Thermohydraulic Behavior in a Primary Cooling System during Loss-of-Coolant Accident of a Light-Water Reactor (Results of the Mock-up Test with ROSA-I),” JAERI-M 6318 (1975). [in Japanese]
- H. Adachi, M. Okazaki, M. Sobajima, M. Suzuki, M. Tasaka, K. Soda and M. Shiba, “ROSA-II experimental program for PWR LOCA/ECCS integral tests”, JAERI 1277, 1982
- K. Tasaka, Y. Koizumi, M. Suzuki, Y. Anoda, Y. Kukita, H. Kumamaru, H. Nakamura, T. Yonomoto, M. Kawaji and H. Murata, “ROSA-III experimental program for BWR LOCA/ECCS integral simulation tests”, JAERI 1307, 1987
- K. Tasaka, R.R. Schultz, Y. Koizumi, Y. Kukita, M. Tanaka, H. Nakamura, M. Osakabe and M. Kawaji, 'The ROSA-IV Program TMI-2 Type Scenario Experiments: A Multifaceted Investigation', Proc. Specialist Mtg. on Small Break LOCA Analysis in LWRs, Pisa, vol. 2 (1985) 11-25
- ROSA-V Group, "ROSA-V Large Scale Test Facility (LSTF) System Description for the Third and Fourth Simulated Fuel Assemblies," JAERI-Tech 2003-037 (2003)
- K. Tasaka, Y. Kukita, H. Asaka, T. Yonomoto and H. Kumamaru, “The Effects of Break Location on PWR Small Break LOCA”, 3rd Int. Topical Mtg. on Nuclear Power Plant Thermal-Hydraulics and Operations, Seoul (1988) Paper A5.B-2
- Y. Kukita, Y. Anoda and K. Tasaka, “Summary of ROSAIV LSTF First-Phase Program: Integral Simulation of PWR Small-Break LOCAs and Transients”, Nucl. Engrg. Des., 131-1 (1991) 101-111 https://doi.org/10.1016/0029-5493(91)90320-H
- H. Kumamaru and Y. Kukita, “Pressurized Water Reactor Small Break LOCA: Effect of Break Area and Intentional System Depressurization for Prevention of Excess Core Dryout”, J. Nucl. Sci. Technol. 29-12 (Dec. 1992) 1162-1172 https://doi.org/10.3327/jnst.29.1162
- Y. Kukita, H. Nakamura, T. Watanabe, H. Asaka, T. Yonomoto, M. Suzuki, H. Kumamaru and Y. Anoda, “OECD/NEA/CSNI International Standard Problem No.26 Comparison Report”, NEA/CSNI R(91)13 (Feb. 1992)
- F. Serre, T. Yonomoto, Y. Kukita and K. Tasaka, “CATHARE Analysis of Natural Circulation under Normal and Degraded Secondary Cooling Conditions in LSTF Runs ST-NC-06 and ST-NC-07”, EUROTHERM Seminar No. 16, Natural Circulation in Industrial Applications, Pisa (1990) 103-110
- T. Watanabe and Y. Kukita, “Effects of ECCS and Pressurizer Auxiliary Spray on the Experiment Simulating Mihama Unit-2 Steam Generator U-tube Rupture Incident”, Proc. 5th Int. Topical Mtg. on Nucl. Reactor Thermal-Hydraulics (NURETH5), Salt Lake (1992) 1013-1020
- Y. Anoda, H. Nakamura, T. Watanabe, M. Hirano and Y. Kukita, “Experimental and Analytical Simulations of the Mihama Unit-2 Steam Generator Tube Rupture Incident”, Proc. 1993 Simulation Multiconference (Session on Simulation Approaches to Safety Assessment and Diagnostics), Washington, D.C. (1993)
- Y. Kukita, H. Nakamura, K. Tasaka and C. Chauliac, “Non-Uniform Steam Generator U-Tube Flow Distribution during Natural Circulation Tests in ROSA-IV Large Scale Test Facility”, Nucl. Sci. Engrg. 99 (1988) 289-298 https://doi.org/10.13182/NSE99-289
- Y. Kukita, H. Nakamura, Y. Anoda and K. Tasaka, “Hot Leg Flow Characteristics during Two-Phase Natural Circulation in pressurized Water Reactor”, Proc. NURETH4, Karlsruhe (1989) 465-470
- Y. Kukita, T. Yonomoto, H. Asaka, H. Nakamura, H. Kumamaru, Y. Anoda, T.J. Boucher, M.G. Ortiz, R.A. Shaw and R.R. Schultz, “ROSA/AP600 Testing: Facility Modifications and Initial Test Results,” J. Nucl. Sci.Technol. 33(3) (1996) 259-265 https://doi.org/10.3327/jnst.33.259
- M. Kondo, H. Nakamura, Y. Anoda, S. Saishu, H. Obata, R. Shimada and S. Kawamura, “Roll Wave Effects on Annular Condensing Heat Transfer in Horizontal PCCS Condenser Tube”, Proc. 10th Intnl. Conf. on Nucl. Engng. (ICONE10), Arlington (2002) ICONE10-22403 https://doi.org/10.1115/ICONE10-22403
- M. Suzuki, T. Takeda, H. Asaka and H. Nakamura, ”Effects of Secondary Depressurization on Core Cooling in PWR Vessel Bottom Small Break LOCA Experiments with HPI Failure and Gas Inflow,” J. Nucl. Sci. Technol., 43 (2006) 55 https://doi.org/10.3327/jnst.43.55
- H. Nakamura and Y. Kukita, “PWR thermal-hydraulic phenomena following loss of residual heat removal (RHR) during mid-loop operation,” Proc. of Int. Conf. on New Trends in Nucl. System Thermal-hydraulics, Pisa, Vol. 1 (1994) 77
- M. EricksonKirk et al., “Technical Basis for Revision of the Pressurized Thermal Shock (PTS) Screening Limit in the PTS Rule (10 CFR 50.61): Summary Report,” NUREG-1806 (2006)
- P. Muehlbauer, “Review of experimental database on mixing in primary loop and future needs,” EVOL-ECORAD03 (2003)
- FLUENT 6.2 User’s Guide, Fluent Inc.(2005)
- K. Kawanishi et al., “Experimental Study on Suddenly Steam Condensation in Horizontal Tubes with Vertical Inlet Tubes,” J. of Atomic Energy Society of Japan 39-5 (1997) (in Japanese)
- The RELAP5 Code Development Team, “RELAP5/MOD3 Code Manual,” NUREG/CR-5535, INEL-95/0174 (1995)
- V.G. Zimin et al., “Verification of the J-TRAC Code with 3D Neutron Kinetics Model SKETCH-N for PWR Rod Ejection Analysis,” Proc. NURETH-9, USA (1999)
- M. Suzuki, T. Takeda and H. Nakamura, “Performance of Core Exit Thermocouple for PWR Accident Management Action in Vessel Top Break LOCA Simulation Experiment at OECD/NEA ROSA Project,” Proc. ICONE-16, Orlando (2008) ICONE16-48754 https://doi.org/10.1299/jpes.3.146
- H. Nakamura, T. Watanabe, T. Takeda, H. Asaka, M. Kondo, Y. Maruyama, I. Ohtsu and M. Suzuki, “RELAP5/MOD3 de Verification through PWR Pressure Vessel Small Break LOCA Tests on OECD/NEA ROSA Project,” ibid. ICONE16-48615
Cited by
- Condensation induced water hammer (CIWH) – relevance in the nuclear industry and state of science and technology vol.78, pp.1, 2013, https://doi.org/10.3139/124.110303
- RELAP5 Analyses of ROSA/LSTF Experiments on AM Measures during PWR Vessel Bottom Small-Break LOCAs with Gas Inflow vol.2014, pp.2314-6060, 2014, https://doi.org/10.1155/2014/803470