Experimental Analysis of Steam Generator Tube Rupture in CIRCE Facility for MYRRHA Configuration

One of the main safety issues of Generation IV (Gen IV) heavy liquid-metal fast reactors is the postulated steam generator tube rupture (SGTR) accident. This event is characterized by primary and secondary coolant interaction, referred to in the literature as a coolant-coolant interaction event having a nonzero probability to occur. This accident scenario could affect the safety of a pool-type reactor, as a consequence of water secondary coolant flashing into the primary coolant liquid metal. The SGTR event needs to be experimentally characterized to evaluate the pressure waves effect, tube rupture propagation (domino effect), oxide precipitation and slug and plug formation, cover gas pressurization of the reactor, and steam flow paths through the pool and eventually the core, entailing the risk of positive reactivity insertion (due to positive local void coefficient). The design phase of the Gen IV MYRRHA plant has dealt with postulated SGTR safety issues in the framework of the MAXSIMA project, which is supported by the European Commission. A relevant contribution to this research activity was provided by the Italian Agency for New Technologies, Energy and Sustainable Economic Development Research Center Brasimone, where a new test section has been designed, assembled, instrumented, and implemented in the large-scale integral-effects pool facility CIRCE for investigating the SGTR event in a relevant configuration for the heat removal system of MYRRHA. This research reactor is not oriented to steam production for running a turbogenerator (no electric production), thus the heat removal system is referred to as the primary heat exchanger (PHX) and not as a steam generator. Four full-scale portions (four bundles of 31 tubes) of the MYRRHA PHX were adopted to carry out four independent SGTR experiments. Water flowed upward in the central tube of the bundle and two rupture positions were investigated at the lower and upper levels, named the bottom and middle scenarios, respectively. After the rupture, water was injected at 16 bar and 200°C into lead bismuth eutectic alloy at 350°C. The experimental results showed a remarkable repeatability and were presented in terms of (1) CIRCE vessel pressurization up to 2.7 and 4 bar absolute for the middle and bottom scenarios, respectively; (2) vapor flow paths through the bundle and its cooling effect up to 120°C and 140°C for the middle and bottom tests, respectively; and (3) strain measurements on tubes and bundle shells up to 2800 μm/m. The integrity of the tubes surrounding the ruptured one and the effectiveness of implemented safety device (rupture disks) pressure relief were significant engineering feedback for MYRRHA designers. The acquired high-quality data also constitute a database increase for future code verification and validation and possible new model development. The performed experimental analysis provided the awareness that a suitable design of a depressurization system (e.g., rupture disks) could allow for addressing postulated SGTR events in the MYRRHA configuration with confidence and safety.