Experimental and Numerical Analysis of Steam Generator Tube Rupture Event for MYRRHA Reactor in CIRCE Facility With SIMMER-IV Code

作者：A. Pesetti, M. Tarantino, N. Forgione · 发表于：Volume 5: Advanced Reactors and Fusion Technologies; Codes, Standards, Licensing, and Regulatory Issues · 年份：2018 · DOI：10.1115/icone26-82503 · 被引用次数：9 · 研究领域：Materials Science

The Steam Generator Tube Rupture (SGTR) postulated event constitutes one of the most hazardous safety issues for Gen IV pool reactors, cooled by heavy liquid metals. This accidental scenario is characterized by quick water flashing when in contact with primary coolant liquid metal, causing pressure wave propagation, cover gas pressurization in the reactor main vessel as well as possible tube rupture propagation, vapour dragged through the core, oxides precipitation and consequent slugs and plugs formation. The design phase of Gen IV MYRRHA reactor addressed the SGTR scenario issues in the framework of MAXSIMA project, supported by the European Commission. This research activity was fully executed at ENEA CR Brasimone, where a new test section was designed, assembled, instrumented and implemented in the large scale pool facility CIRCE. It was supported by the execution of preliminary and detailed pre-tests analysis performed adopting SIMME-III and -IV code, respectively. This paper details the test section main features, able to host four full scale portions (each one constituted by 31 tubes) of the MYRRHA Primary Heat eXchanger (PHX), for carrying out four independent SGTR experiments. A couple of tests investigated the tube rupture at middle position between two spacer grids of the bundle. The other two tests analysed instead the rupture near the bottom tube plate. Auxiliary systems were adopted for reaching primary (Lead Bismuth Eutectic alloy, LBE) and secondary (water) ...