Analyses of loads on reactor pressure vessel internals in a pressurized water reactor due to a loss-of-coolant accident considering fluid-structure interaction
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P. Akimov
Abstract
During a postulated loss-of-coolant accident, significant structural loadings on the reactor pressure vessel and its internals can be exerted due to the non-uniform and time-dependent pressure differences. The integrity of the internals must be secured, so as to ensure reactor shutdown and post-accident heat removal. Two developed computer codes to be discussed in this paper are used for the analysis of loads on the reactor pressure vessel internals. In the analyses the fluid-structure interaction effects are taken into account in the annular space between the pressure vessel and the core support barrel as well as in the upper plenum. This is a more realistic approach with respect to peak loadings and structural response frequencies.
Kurzfassung
Während eines Kühlmittelverluststörfalles können signifikante Strukturbelastungen auf den Reaktordruckbehälter und seine Einbauten durch zeitabhängige und ungleichförmig verteilte Druckdifferenzen angeregt werden. Die Integrität der Einbauten muss sichergestellt werden, um eine Reaktorabschaltung und die Wärmeabfuhr nach dem Störfall gewährleisten zu können. In dieser Arbeit werden zwei Rechenprogramme präsentiert und zur Ermittlung der Belastungen auf die Einbauten des Reaktordruckbehälters angewendet. In den Analysen wurde die Fluid-Struktur-Wechselwirkung im Ringraum zwischen dem Reaktordruck- und Kernbehälter sowie im oberen Plenum berücksichtigt. Dieses Verfahren führt zu realistischeren maximalen Belastungen und Strukturfrequenzen.
References
1 Dynamic loads on the RPV internals and the Loop of a Pressurized Water Reactor in case of Blowdown (calculation with DAISY for a 0.1 A-leak in the cold leg). Report GRS-A-1025, 1984.Suche in Google Scholar
2 Dynamic loads on the RPV internals and the Loop of a Pressurized Water Reactor in case of Blowdown (calculation with DAISY for a 2A-break in the cold leg). Report GRS-A-1065, 1985.Suche in Google Scholar
3 Rivard, W. C.: HAUPT: A computer program for hydroelastic analysis of upper plenum transients. Flow Science Inc., Los Alamos, USA, 1984.Suche in Google Scholar
4 TRAC-PF1/MOD: An advanced best-estimate computer program for pressurized water reactor thermal-hydraulic analysis, Los Alamos National Laboratory Report NUREG/CR-3858, 1987.Suche in Google Scholar
5 Obereisenbuchner, L.: Analysis of loads on upper plenum internals in a PWR due to pressure waves. NURETH-11, 11th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Avignon (France), October 2–6, 2005, p. 300.Suche in Google Scholar
© 2010, Carl Hanser Verlag, München
Artikel in diesem Heft
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries/Kurzfassungen
- Technical Contributions/Fachbeiträge
- Analyses of loads on reactor pressure vessel internals in a pressurized water reactor due to a loss-of-coolant accident considering fluid-structure interaction
- Preliminary evaluation of a radioactive waste repository safety performance by a Monte Carlo simulation-based reliability model
- Determination of effects of burn up and reflector material on the kinetic parameters for open pool reactor using MCNP code
- Analysis of fuel rod behaviour during limiting RIA in RBMK plants
- Reflection on the ductility of irradiated zircaloy-4 fuel rod cladding
- Prediction, analysis and solution of flow inversion phenomenon in a typical MTR reactor with upward core cooling
- Steam drum process dynamics and level control of a pressure tube BWR
- Nuclear data for cyclotron production of 114mIn/114In and 140Nd/140Pr used in gamma camera monitoring, RIT, ERT and PET
- Effect of thermal gap conductance for MoO3 ampoules irradiated in a high neutron flux
- U1 approximation to the neutron transport equation and calculation of the asymptotic relaxation length
- Application of the TN method to critical slab problem for one-speed neutrons with forward and backward scattering and efficiency of reflection coefficient
- Technical Notes/Technische Mitteilungen
- Rapid preparation of Uranium and Thorium alpha sources by electroplating technique
Artikel in diesem Heft
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries/Kurzfassungen
- Technical Contributions/Fachbeiträge
- Analyses of loads on reactor pressure vessel internals in a pressurized water reactor due to a loss-of-coolant accident considering fluid-structure interaction
- Preliminary evaluation of a radioactive waste repository safety performance by a Monte Carlo simulation-based reliability model
- Determination of effects of burn up and reflector material on the kinetic parameters for open pool reactor using MCNP code
- Analysis of fuel rod behaviour during limiting RIA in RBMK plants
- Reflection on the ductility of irradiated zircaloy-4 fuel rod cladding
- Prediction, analysis and solution of flow inversion phenomenon in a typical MTR reactor with upward core cooling
- Steam drum process dynamics and level control of a pressure tube BWR
- Nuclear data for cyclotron production of 114mIn/114In and 140Nd/140Pr used in gamma camera monitoring, RIT, ERT and PET
- Effect of thermal gap conductance for MoO3 ampoules irradiated in a high neutron flux
- U1 approximation to the neutron transport equation and calculation of the asymptotic relaxation length
- Application of the TN method to critical slab problem for one-speed neutrons with forward and backward scattering and efficiency of reflection coefficient
- Technical Notes/Technische Mitteilungen
- Rapid preparation of Uranium and Thorium alpha sources by electroplating technique
