1,720,993 research outputs found
Experimental Analysis of Steam Generator Tube Rupture in CIRCE Facility for MYRRHA Configuration
No full textOne 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
Experimental qualification of the ITER pressure suppression system by means of a large scale facility: Assessment of similitude analysis
Full text linkThe paper deals with the assessment of the pressure suppression system of the international thermonuclear experimental reactor (ITER). An extensive experimental research programme performed in a small-scale test facility analysed the steam direct condensation at sub-atmospheric pressure (about 400 tests). A similitude analysis was elaborated for extrapolating the results to the ITER full scale system. The derived scale laws were applied to analyse the performances of the pressure suppression system in order to manage a large loss of coolant accident (LOCA) scenario postulated in the vacuum vessel of ITER. Experimental tests, performed in a large-scale test facility built at the University of Pisa (geometrical scale about 1 and 1/10 power scale), analysing the first part of the large LOCA event, assessed the scale laws emphasizing some discrepancies determined in the extrapolation of reduced scale results to the actual system. Introducing a justified correction factor, the presented similitude analysis assured a suitable and reliable temperature and pressure increase prediction of full-scale systems, even in scaled configuration affected by relevant thermal stratification
Pre-test Analysis of SGTR event on large scale experimental facility by SIMMER-IV code
No full textThis document reports the design of the test section planned to be implemented in the S100 vessel of CIRCE facility to perform a large scale experimental investigation of heavy liquid metal-water interaction, following the SGTR event, in a relevant configuration for MYRRHA reactor. Moreover, the detailed pre-test analysis carried out by SIMMER IV code is presented. The performed pre-tests provided information for achieving a safe and effective experimental campaign
Double wall bayonet tube steam generator investigation in hero experimental campaign
No full textThe large LBE (Lead-Bismuth Eutectic) pool integral effect CIRCE (CIRcolazione Eutettico) facility at CR ENEA Brasimone, implementing the HERO (Heavy liquid mEtal pRessurized water cOoled tubes) Test Section (TS), has been involved in an experimental campaign investigating the thermal-hydraulic behavior of a new concept of Steam Generator (SG) based on double wall bayonet tubes concept. The R&D activities aim at improving the knowledge and the experience in terms of design and operations for Lead-cooled Fast Reactor (LFR) and Accelerator Driven System (ADS), and providing a database for STH codes validation. The HERO SG represents a mock-up (1:1 in length) of the ALFRED (Advanced Lead Fast Reactor European Demonstrator) Steam Generator. An experimental campaign has been carried out in the framework of the HORIZON2020 SESAME (Simulations and Experiments for the Safety Assessment of MEtal cooled reactors) European project, in order to support the development of the ALFRED SG with a set of high secondary side pressure tests (~172 bar). In the same configuration, an experimental campaign has been carried out with low pressure secondary side (~16 bar) in the framework of the HORIZON2020 MYRTE (MYRRHA Research and Transmutation Endeavour) European project providing support to the development of MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) and acquiring thermo-dynamic feedbacks of the Primary Heat Exchanger (PHX) behavior. For these purposes, the HERO TS has been implemented in the CIRCE facility and a dedicated instrumentation has been installed. The aim of this work is to present the experimental tests performed and the main results achieved in steady-state conditions, characterizing the system behavior in terms of primary LBE mass flow rate, temperatures and pressure, both at low and high pressure on the secondary side
Analysis of feasibility of a new core catcher for the in-vessel core melt retention strategy
Full text linkThis study deals with the feasibility study of a new in-vessel core melt retention (IVCMR) strategy capable to extend the coping period in the event of adverse situations, involving the melting of the core. Since Fukushima accident, many studies have been carried out to resolve the severe accident mitigation issues related to the corium stabilization inside and outside the reactor vessel. This is in fact one of the most relevant safety issues to secure LWRs from the point of view of severe accident mitigation and containment integrity. As for the corium stabilization inside the reactor vessel, in this study it is proposed a new IVCMR concept, developed at the University of Pisa, based on the adoption of an original core catcher design made of batches of ceramic material. By profiting of its low thermal conductivity, this core catcher is capable to retard the heat-up of the lower head of the vessel during the phase of relocation of the corium. To support the feasibility of its design analytical and numerical analyses have been performed assuming homogeneous pool condition. Results show that the adoption of the proposed core catcher solution extends the severe accident coping period: after 1 h from the initiating event, the maximum temperature of the vessel wall is below the limit for which localized failure may appear
Water/Pb-Bi Interaction Experiments in LIFUS5/Mod2 Facility Modelled by SIMMER code
No full textThe new generation Heavy Liquid Metal Reactors (HLMRs) are characterized by a pool type configuration having high pressure steam generators set inside the reactor vessel. The
primary (hot low pressure LBE melt) and secondary (high
pressure sub-cooled water) coolant could come into contact as
consequence of Steam Generator Tube Rupture (SGTR)
phenomenon, that cannot be considered negligible. The
structural integrity of the reactor internals, steam generators in particular, could be affected by the SGTR scenario
consequences. The pressure wave propagation, cover gas
pressurization, domino effect on the surrounding tubes,
reactivity feedback due to steam dragged into the core, primary system pollution and slug formation constitute the most hazardous effects of the SGTR accident. Therefore, this
accidental scenario constitutes a safety issue in the design and
in the preliminary safety analysis. A key issue in the SGTR
analysis for HLMRs is constituted by the availability of
qualified experimental data, suitable to be extrapolated to full
scale plant and to support the computer code development and demonstrating the code reliability in the phenomena prediction (qualified code). In this paper, part of the experimental campaign performed in the LIFUS5/Mod2 facility at ENEA CR Brasimone (in the frame of the THINS project), investigating the water-LBE interaction, is reported. The experimental activity aimed to provide high-quality measurement data for supporting the development and validation phase of computer codes for SGTR numerical simulation. The reported experimental test has been
numerically simulated by SIMMER III code. The pressure,
temperature and injected water mass flow rate time trends have been computed during the water-LBE interaction in the reaction vessel. The work aims to evaluate the prediction capability of the two-dimensional SIMMER III code and to determine the suitability of the SIMMER code physical models
Numerical analysis of sub-atmospheric steam condensation in suppression tank with SIMMER IV code
No full textOne of the key safety components for nuclear fusion plants is the suppression tank, which is designed to protect the Vacuum Vessel (VV) against accidental pressurization events, e.g. Loss Of Coolant Accident (LOCA). In this framework the attention is focused on the Vacuum Vessel Pressure Suppression System (VVPSS), made of water tanks in which the pure steam, or eventually mixed with incondensable gases, is injected and consequently the overpressure is dumped profiting of Direct Contact Condensation (DCC). The design constraints of fusion reactor dictate that the pressure resulting (long-term) from any accidental or baking condition should be always kept lower than 0.15 MPa. The study of the phenomena evolving during DCC in LOCA conditions is the major novelty, especially in consideration of the lack of similar studies in the available literature. In this context, a wide series of experimental tests was carried out at Pisa University (UNIPI), Department of Civil and Industrial Engineering (DICI), in a Small Scale Test Facility (SSTF), designed and instrumented for investigating DCC at sub-atmospheric pressure, by varying water pool temperature, pressure and steam mass flow rate. The adoption and assessment of suitable numerical codes, to reliably simulate such a cutting-edge multiphase multicomponent scenario, have a crucial role for contributing to the phenomena understanding and for possible safety analysis of full-scale components. On this basis, a preliminary evaluation of the cartesian three-dimensional SIMMER IV code capabilities in simulating DCC at sub-atmospheric conditions was carried out, taking as reference one UNIPI test. SIMMER IV code was able to set up precise initial low-pressure boundary conditions and simulate superheated steam condensation in subcooled water pool, with condensation efficiency comparable to the experimental one. Moreover, SIMMER IV code predicted a longitudinal steam plume dimension and injected steam velocity consistent with experimental data
Double wall bayonet tube steam generator investigation in hero experimental campaign
No full textThe large LBE (Lead-Bismuth Eutectic) pool integral effect CIRCE (CIRcolazione Eutettico) facility at CR ENEA Brasimone, implementing the HERO (Heavy liquid mEtal pRessurized water cOoled tubes) Test Section (TS), has been involved in an experimental campaign investigating the thermal-hydraulic behavior of a new concept of Steam Generator (SG) based on double wall bayonet tubes concept. The R&D activities aim at improving the knowledge and the experience in terms of design and operations for Lead-cooled Fast Reactor (LFR) and Accelerator Driven System (ADS), and providing a database for STH codes validation. The HERO SG represents a mock-up (1:1 in length) of the ALFRED (Advanced Lead Fast Reactor European Demonstrator) Steam Generator. An experimental campaign has been carried out in the framework of the HORIZON2020 SESAME (Simulations and Experiments for the Safety Assessment of MEtal cooled reactors) European project, in order to support the development of the ALFRED SG with a set of high secondary side pressure tests (~172 bar). In the same configuration, an experimental campaign has been carried out with low pressure secondary side (~16 bar) in the framework of the HORIZON2020 MYRTE (MYRRHA Research and Transmutation Endeavour) European project providing support to the development of MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) and acquiring thermo-dynamic feedbacks of the Primary Heat Exchanger (PHX) behavior. For these purposes, the HERO TS has been implemented in the CIRCE facility and a dedicated instrumentation has been installed. The aim of this work is to present the experimental tests performed and the main results achieved in steady-state conditions, characterizing the system behavior in terms of primary LBE mass flow rate, temperatures and pressure, both at low and high pressure on the secondary side
Protected loss of flow accident simulation in circe-hero facility: Experimental test and system code assessment
No full textIn the framework of the Generation IV ALFRED (Advanced Lead Fast Reactor European Demonstrator) design, a dedicated Test Section (TS) named HERO (Heavy liquid mEtal pRessurized water cOoled tubes) has been implemented in the large LBE (Lead-Bismuth Eutectic) pool integral effect CIRCE (CIRColazione Eutettico) facility at CR ENEA Brasimone. HERO is a mock-up (1:1 in length) of the ALFRED Steam Generator (SG), consisting of seven double wall bayonet tubes arranged in triangular pitch inside a hexagonal shell, with an active length of ~6 m. An integral test experiment has been designed and carried out in the framework of the HORIZON2020 SESAME (Simulations and Experiments for the Safety Assessment of MEtal cooled reactors) European project. The test consists in a Protected Loss of Flow Accident (PLOFA) occurring with the facility operated in nominal steady-state conditions for both primary side (LBE) and secondary side (high pressure water). In steady-state conditions, the LBE mass flow rate is promoted by the injection of argon simulating the behavior of the primary pump, while the thermal power is supplied with an electrically heated Fuel Pin Simulator (FPS). The transient is obtained reducing the FPS power according to a characteristic heat decay curve, while the loss of the primary pump is simulated by the reduction of the gas injection. The loss of the heat sink is simulated managing the HERO feedwater in the secondary loop. A simulation activity has been carried out with RELAP5-3D© v. 4.3.4 system thermal-hydraulic code. A numerical model of the primary loop and HERO SG have been set up and employed to reproduce the main phenomena involved in the system during the test. The present paper aims to present and to describe the experimental results achieved during the transient and to compare the code results with the experimental data
Effect of non-condensable gas in steam condensation at sub-atmospheric pressure condition
No full textIn the International Thermonuclear Experimental Reactor (ITER), a postulated Loss Of Coolant Accident (LOCA) in the Vacuum Vessel (VV) has to be managed with a pressure suppression system working at sub-atmospheric pressure. The operating conditions considerably differ from those experienced in the fission nuclear power plants such as BWR, since the ITER Tokamak works at very low pressure conditions and can withstand a maximum pressure of 0.15 MPa. For this reason, the pressure value must not exceed 10 kPa for a water temperature of 30 °C inside the Vapour Suppression Tanks (VSTs) that are the fundamental components of the Vacuum Vessel Pressure Suppression System (VVPSS). During a LOCA some non-condensable gases (mainly hydrogen and oxygen gases due to the water radiolysis or thermolysis) may be mixed in the steam and this could impair the condensation efficiency. In order to investigate the effects of non-condensable gas on DCC, we conducted a research program funded by ITER Organization at the laboratory of the University of Pisa: we designed and built a small-scale experimental rig to study the steam Direct Contact Condensation (DCC) with the presence of non-condensable gas and simulate the behaviour of a VST. Since DCC can occur with different characteristics, we ran 12 closed mode tests exploring all condensation regimes injecting a certain mass of air with the steam discharged in the subcooled water. The tests started at the saturation pressures corresponding to water temperatures ranging from 40 °C to 80 °C and ended when the free space volume reached the atmospheric pressure. From the analysis of the data acquired during the tests we observed that the condensation efficiency remained higher than 95%. We observed that despite this, the presence of a certain quantity of non-condensable gas has negative aspects on condensation: the condensation regime never reaches stability (the regimes quickly shift towards instability). Furthermore, the presence of air triggers a turbulence of the flow which interferes with the transfer of heat from the steam to the water. Not only the introduction of air into the flow increases linearly the pressure above the water head but its high temperature further contributes to the pressure increase. The air mixed with the flow also forms eddies that can trap the steam and transport it out of the water, preventing it from condensing. Apart from condensation, the most noticeable problem is the rapid increase in pressure inside the condensation tank
- …
