The Steam Generator Tube Rupture postulated event in a pool type Gen IV Heavy Liquid Metal cooled Fast Reactor system needs to by deeply investigated due to the possible hazardous consequences affecting the structural integrity of internals. These reactor designs have steam generators inside the reactor vessel, thus, the interaction between the secondary side coolant (water) and the HLM (e.g. steam generator tube rupture) has to be considered as challenging safety issue in the design and also in the preliminary safety analysis of these reactor types. The LIFUS5/Mod2 facility experiments series B were executed in the framework of the FP7 EC Lead-cooled European Advanced Demonstration Reactor – LEADER – project. The test section was representative of the Spiral Tube Steam Generator of European Lead Fast Reactor. Water at about 180 bar and 270°C was injected into Lead Bismuth Eutectic alloy at about 2 bar and 400°C. The injection was performed in the center of a tube bundle composed of 188 tubes (a representative portion of SG bundle). This analysis is aimed at providing engineering feedbacks to SG designers (e.g. domino effect and mechanical loadings on components). Moreover, the acquired experimental data constituted a wide database for development and validation of numerical models and calculation codes. © 2019 American Nuclear Society.

Experimental campaign in support of the safety studies of the STGR in LFR

Del Nevo, A.;Eboli, M.;
2019

Abstract

The Steam Generator Tube Rupture postulated event in a pool type Gen IV Heavy Liquid Metal cooled Fast Reactor system needs to by deeply investigated due to the possible hazardous consequences affecting the structural integrity of internals. These reactor designs have steam generators inside the reactor vessel, thus, the interaction between the secondary side coolant (water) and the HLM (e.g. steam generator tube rupture) has to be considered as challenging safety issue in the design and also in the preliminary safety analysis of these reactor types. The LIFUS5/Mod2 facility experiments series B were executed in the framework of the FP7 EC Lead-cooled European Advanced Demonstration Reactor – LEADER – project. The test section was representative of the Spiral Tube Steam Generator of European Lead Fast Reactor. Water at about 180 bar and 270°C was injected into Lead Bismuth Eutectic alloy at about 2 bar and 400°C. The injection was performed in the center of a tube bundle composed of 188 tubes (a representative portion of SG bundle). This analysis is aimed at providing engineering feedbacks to SG designers (e.g. domino effect and mechanical loadings on components). Moreover, the acquired experimental data constituted a wide database for development and validation of numerical models and calculation codes. © 2019 American Nuclear Society.
Generation IV, LIFUS5/Mod2, Safety, SGTR
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Utilizza questo identificativo per citare o creare un link a questo documento: http://hdl.handle.net/20.500.12079/54683
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