Strength evaluation of a steam distribution device in the Ignalina NPP accident localisation system |
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Authors: | Gintautas Dundulis Renatas Karalevicius Sigitas Rimkevicius Ronald F Kulak Algirdas H Marchertas |
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Affiliation: | aLaboratory of Nuclear Installation Safety, Lithuanian Energy Institute, 3 Breslaujos, 44403 Kaunas 35, Lithuania;bRFK Engineering Mechanics Consultants, USA;cNorthern Illinois University, USA |
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Abstract: | The Ignalina NPP has a pressure suppression type of confinement, which is referred to as the accident localization system (ALS). The ALS prevents the release of the radioactive material from the NPP to the environment during a loss-of-coolant accident (LOCA). Ten water pools are located in the two ALS towers (five pools in each tower), which separate the dry well from the wet well. These water pools condense the accident-generated steam and prevent high overpressures in the compartments.The steam distribution device (SDD), with the vertical vent pipes (nozzles) that are inserted under the water of the condensing pools, connects the dry well and the wet well. In case of an accident, these components must be capable of withstanding the dynamic loads generated by a LOCA for successful pressure suppression function.This paper presents the transient analysis of the SDD and their connections to the vertical steam corridors following a LOCA. A thermo-hydraulic analysis of the SDD was performed using the state-of-the-art COCOSYS code to determine pressure and temperature histories resulting from a LOCA. The finite element code NEPTUNE was used to evaluate the structural integrity of the SDD and its supporting reinforced concrete wall. Results show that, although portions of the SDD undergo plastic response and the outside surface of the vertical steam corridor reinforced concrete wall cracks, the structural integrity of the SDD and wall are maintained during a LOCA. |
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Keywords: | ALS accident localization system BRU-B valve for steam discharge to ALS tower (Russian abbreviation) BSRC bottom steam reception chamber CTCS condenser tray cooling circuit ECCS emergency core cooling system GDH group distribution header HCC hot condensate chamber LOCA loss-of-coolant accident MCC main circulation circuit MDBA maximum design basis accident MSV main safety valve NPP nuclear power plant RBMK Russian acronym for ‘channelized large power reactor’ SDD steam distribution device SDH steam distribution header VATESI Lithuanian state atomic energy safety inspection |
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