Modeling of DECL accident in the reactor containment by the CONTAIN 2.0 code
-
M. Abbasi
Abstract
In this paper, a specific type of the Loss of Coolant Accident (LOCA), the DECL (Double Ended Cold Leg) break, that means totally guillotine type of break in the cold leg pipe, has been modeled. The accident is simulated with the CONTAIN 2.0 code. In the event of a LOCA accident, coolant mass and energy are released to the containment through the break. This causes an increase of pressure and temperature in the containment. The modeling is performed in the VVER-1000 reactor containment. The analysis includes average pressure in the containment and temperature distribution in sample cells in the long-time. The results are compared with the existing reports on studies that used the ANGAR code. Results show that the CONTAIN 2.0 code is an adaptable tool for the analysis of nuclear events such as DECL accident.
References
1 Tiselj, I.; Kljenak, I.; Prosek, A.: simulation of containment response during a Large- Break Loss-of-Coolant accident scenario with the CONTAIN computer code, 4th meeting Nuclear Energy in Central Europe, 613–620, 1997.Search in Google Scholar
2 Kljenak, I.; Parzer, I.; Tiselj, I.: containment phenomena during a severe accident with reactor vessel failure at low pressure in a tow-loop pressurized water reactor, Nuclear Energy in Central Europe, 287–294, 1998.Search in Google Scholar
3 Murata, K. K.; Williams, D. C.; Tills, J.; Griffith, R. O.; Gido, R. G.; Tadios, E. L.; Davis, F. J.; Martinze, G. M.; Washington, K. E.: code manual for CONTAIN 2.0: a computer code for nuclear reactor containment analysis, NUREG/CR-653, SAND97-1735, Sandia National Laboratories, 1997.Search in Google Scholar
4 Bergeron, K. D.; Williams, D. C.: CONTAIN calculations of containment loading of DRY PWRs, Nuclear Engineering and Design, 90, 153–159, 1985.10.1016/0029-5493(85)90005-6Search in Google Scholar
5 Williams, D. C.; Griffith, R. O.; Tadios, E. L.; Washington, K. E.: Assessment of the CONTAIN Direct Containment Heating (DCH) Model: Analysis of DCH Integral Experiments, SAND94-1174, Sandia National Laboratories, 1996.Search in Google Scholar
6 Notafrancesco, A.; Esmaili, H.; Lee, B.; Tills, J.: Application of the MELCOR code to design basis BWR containment analysis, RES/FSTB 2011-01, United States Nuclear Regulatory Commission, 2011.Search in Google Scholar
7 Kljenak, I.; Mavko, B.: Simulation of containment thermal-hydraulics in the Marviken blowdown 16 experiment with ASTEC and CONTAIN codes, Nuclear Engineering and Design, 2010.10.1115/ICONE17-75097Search in Google Scholar
8 Final Safety Analysis Report (FSAR), Engineering Safety Features, chapter 6, (2005) 6.2.1.9–35Search in Google Scholar
9 Final Safety Analysis Report (FSAR), Accident Analysis, chapter 15, (2005) 15.2.3.4.2.Search in Google Scholar
© 2013, Carl Hanser Verlag, München
Articles in the same Issue
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Technical Contributions/Fachbeiträge
- Development of inflatable seals for the rotatable plugs of sodium cooled fast breeder reactors: a review – Part I: key areas
- Numerical simulation of turbulent flow mixing inside a square chimney structure of a research reactor
- Thermal analysis of VWSB-IP1 at Tarapur
- Tracer transport modeling with the Alliances platform in the presence of evapotranspiration
- Seasonally gross alpha and beta activity concentration in surface water and sediments in Sır Dam Pond
- Evaluation of radioactive emissions of lignite-fired power plants in Turkey using the Analytic Hierarchy Process
- Technical Notes/Technische Mitteilungen
- Modeling of DECL accident in the reactor containment by the CONTAIN 2.0 code
- Application of the Henyey-Greenstein and Anlı-Güngör phase functions for the solution of the neutron transport equation with Legendre polynomials: Reflected critical slab problem
- Effect of gamma-radiation on sorption and precipitation of radionuclides
Articles in the same Issue
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Technical Contributions/Fachbeiträge
- Development of inflatable seals for the rotatable plugs of sodium cooled fast breeder reactors: a review – Part I: key areas
- Numerical simulation of turbulent flow mixing inside a square chimney structure of a research reactor
- Thermal analysis of VWSB-IP1 at Tarapur
- Tracer transport modeling with the Alliances platform in the presence of evapotranspiration
- Seasonally gross alpha and beta activity concentration in surface water and sediments in Sır Dam Pond
- Evaluation of radioactive emissions of lignite-fired power plants in Turkey using the Analytic Hierarchy Process
- Technical Notes/Technische Mitteilungen
- Modeling of DECL accident in the reactor containment by the CONTAIN 2.0 code
- Application of the Henyey-Greenstein and Anlı-Güngör phase functions for the solution of the neutron transport equation with Legendre polynomials: Reflected critical slab problem
- Effect of gamma-radiation on sorption and precipitation of radionuclides
