1,720,984 research outputs found
Experimental investigation and SIMMER-III code modelling of LBE-water interaction in LIFUS5/Mod2 facility
Full text linkIn the frame of the THINS Project, an experimental campaign was performed on LIFUS5/Mod2 facility at ENEA RC Brasimone, aiming to investigate the water-LBE interaction. Such a phenomenon occurs as consequence of a postulated Steam Generator Tube Rupture event in a HLMFR system. Four tests were performed injecting sub-cooled water at 40 bar into a reaction vessel partially filled by low pressure LBE at 400°C. The post-test activity was performed by the SIMMER-III code in order to improve the understanding of the involved phenomena and to confirm the code capabilities in simulating the water-LBE interaction. The calculated data showed a qualitatively agreement with the measured values and a faster reaction kinetics due to the modelling assumptions
Fornitura Coperchio S100 Impianto CIRCE per prove a pressione elevata
No full textIl presente documento riporta la specifica tecnica di fornitura e collaudo del coperchio del serbatorio S100 dell’impianto CIRCE, specifico per prove di sicurezza dedicate all’investigazione sperimentlae su larga scala dell’ interazione metallo liquido pesante e acqua
Experimental and numerical investigations of interaction between heavy liquid metal and water for supporting the safety of LFR GEN. IV reator design
No full textGeneration IV Heavy Liquid Metal (HLM) fast reactors have steam generators inside the reactor vessel. Therefore, 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. In this framework, the separate effect facility LIFUS5/Mod2, installed at ENEA Research Center, has a new test section with a geometry representative of the SG tube bundle of ELSY design steam generator. Experiments were executed to study the interaction between LBE and water, with boundary and initial conditions relevant of the SGTR accident, the potential for tube-to-tube rupture propagation, as well as to demonstrate the reliability of computer codes in simulating the phenomena of interest. The paper presents LIFUS5/Mod2 facility and its capabilities, results from the tests' series B1 and preliminary post-test calculations by SIMMER-III code. The experiments provide pressure, temperature and strain trends versus time at a frequency up to 10 kHz, suitable for the analysis of interaction phenomena and the code validation. The experimental pressure trends highlights a remarkable damping of wave propagation generated because the presence of the tube bundle (that was not damaged). Codes are applied for supporting the experimental campaign, the design of experiments and for the experimental data analysis. RELAP5/MOD3.3 supported the analyses of the facility water injection line. The post-tests are mainly based on the comparisons of the experimental and calculated pressure trends. SIMMER-III provided an excellent simulation of the first pressure peak resulting from the rupture of the injector
Influence of Noncondensable Gas–Dust Mixture on Direct Contact Condensation of Steam at Atmospheric Pressure
No full textAt the Department of Civil and Industrial Engineering (DICI) of the University of Pisa, an experimental research program, funded by International Thermonuclear Experimental Reactor (ITER) Organization, concerning steam direct condensation in a flux containing also noncondensable gas and dust, was carried out. This mixture of fluids and dust is injected into the ITER pressure suppression tanks during a loss of coolant accident (LOCA) in the Vacuum Vessel. The aim of the research program is to determine the steam condensation efficiency in such conditions. Experimental tests were performed injecting this mixture in a tank partially filled with water. Alumina was used to simulate the actual dust present in the ITER Vacuum Vessel. Mass flow rates, temperature, and pressure of the different fluids involved were recorded during the tests. The steam condensation into the subcooled water pool at a temperature ranging between 318 and 369 K was investigated to determine the condensation regimes occurring during the mixture injection. The values of the fraction of the energy absorbed by water, dust, and metallic structures of the heat losses and of the average heat transfer coefficient were determined considering pure steam, steam-dust and steam-air-dust injection. The average heat transfer coefficient, determined calculating the steam jet surfaces by means of image elaboration, was compared with empirical correlations
Structural and thermal fluid dynamics analyses of the ITER Pressure Suppression System considering no stable steam condensation regimes
No full textAt the University of Pisa, an experimental campaign was performed with the purpose of qualifying the ITER Vacuum Vessel Pressure Suppression System (VVPSS) and studying the Direct Contact Condensation at sub-atmospheric conditions. With financial support from ITER Organization, a Large-Scale Experimental Facility was designed and built to investigate the steam condensation in the operation conditions of the ITER VVPSS. The experimental tests were performed by injecting steam, produced by an electric steam generator (1.7 MW of power), through a multi-hole sparger into a condensation tank called Experimental Test Tank (ETT). On the sparger support, strain gauges and an accelerometer were installed to analyse the structural behaviour during the experimental tests. Under certain thermal hydraulic conditions (low subcooling and low steam mass flow rate) a large steam pocket can form and collapse causing high loads on the sparger structure. The aim of this study was comparing numerical fluid dynamics and structural analyses with the experimental results in order to estimate the pressure impulse caused by bubble collapse. The bubble collapse was simulated by means of ANSYS Fluent code in order to analyse the bubble dynamics and to determine the pressure impulse. The shape, the dimension and the collapse time of steam pocket were experimentally determined by means of an image analysis of video recorded during the tests. The numerical and experimental maximum steam bubble volume at the final stage of the transient before the detachment shows a good agreement with an error of 6 %. The pressure load was estimated at 5 different points of the simulation domain; the maximum pressure impulse on the sparger is 1.15 bar and has a duration of 40 ms. A dynamic structural FEM analysis was performed applying the pressure impulses on the sparger in order to evaluate the accelerations. The average value of RMS, maximum, minimum and duration of the experimental acceleration signals was obtained and compared to the numerical one. The errors of the RMS, the maximum and the minimum acceleration of the sparger are 16 %, 35 % and 51 %, respectively. The time duration of the oscillations is in good agreement (the error is only 2.54 %)
Assessment of SIMMER-III code based on steam generator tube rupture experiments in LIFUS5/Mod2 facility
No full textAn experimental campaign investigating the postulated Steam Generator Tube Rupture (SGTR) event, in relevant configurations for Heavy Liquid Metal Reactors (HLMRs), was carried out in the separate-effect facility LIFU5/Mod2, at ENEA CR Brasimone. Ten tests were performed injecting pressurized subcooled water into the reaction tank partially filled by Lead-Bismuth Eutectic alloy (LBE) at 400° C with a cover gas of argon at about 2 bar. Fast pressure transducers, thermocouples and strain gages provided high-quality measurement data for improving the phenomena understanding and supporting the development and validation phase of computer codes for SGTR numerical simulation. The experimental campaign is composed by two series of tests, characterized by different water pressure: 40 and 16 bar. The first two tests belonging to the low pressure experiments are presented, highlighting the pressurization time trends of the water injection tank, injection line and reaction vessel. The injected water mass flow rate and temperature trends in the reaction vessel were measured. The former test is the reference one and the latter was carried out for investigating the injection of water with higher sub-cooling. A post-test analysis of the two mentioned tests was carried out by SIMMER-III code. The pressure profile in the water injection tank was set as boundary condition of the calculation. The numerical analysis provided injection line and reaction tank pressurization in agreement with the experimental data. The lower water temperature test provided a better accordance with the measured data, due to the lower evaporation along the injection line. The SIMMER-III analysis also studied the water-LBE interaction from the volume fraction point of view and the energy released in the total reaction tank and in its cover gas
Experimental and numerical analysis of steam generator tube rupture event for myrrha reactor in circe facility with SIMMER-IV code
No full textThe 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) coolant initial conditions in accordance with MYRRHA design. Water was injected at 16 bar and 200°C in LBE at 350°C under an argon cover gas at about atmospheric pressure. The experimental results of the first test (middle rupture), in terms of CIRCE vessel pressurization, vapour flow path through tube bundle and tubes deformation, are presented. The post-test analysis was performed by SIMMER-IV code adopting the 3D Cartesian code version. The whole main vessel of CIRCE facility and implemented test section were modelled conserving heights and flowing areas. The experimental initial conditions were successfully matched by numerical results as well as the vessel pressurization and temperature time trends in the tube bundle following the SGTR. An important engineering feedback, for MYRRHA designer, was the evidence of rupture propagation absence. Moreover, the effectiveness of implemented safety devices, rupture disks, was evaluated and characterized for pressure relief feedbacks. A wide series of high quality measured data (pressure, temperature, strain and mass flow rate) was acquired and constitutes a database enlargement for future codes validation and possible new model development
Experimental investigation of spiral tubes steam generator rupture scenarios in LIFUS5/MOD2 facility for ELFR
No full textIn the framework of the EC FP7 LEADER project, an experimental campaign was performed in the LIFUS5/Mod2 facility, at ENEA CR Brasimone, for investigating the postulated Steam Generator Tube Rupture (SGTR) event in a relevant configuration for the spiral tube Steam Generator (SG) of the European Lead Fast Reactor (ELFR). Two tests are analysed. The LIFUS5/Mod2 facility implemented a test section composed by 188 tubes, vertically disposed with triangular pitch, in a shell closed by top and bottom flanges and having a perforated cylindrical wall. The central tube injected water at about 180 bar and 270°C, at middle height of the tube bundle, in the reaction tank partially filled by Lead-Bismuth Eutectic alloy (LBE) at 400°C with an argon cover gas at about 2 bar. It was connected to a 2 m3 dump tank, due to the high injection pressure. In the reaction tank fast instrumentation was set: 6 fast Pressure Transducers (PTs) acquiring data at 10 kHz for precisely characterize the first injection peaks; 70 low constant time Thermocouples (TCs) to understand the vapour evolution path; and 13 strain gages (SGGs) to measure the strain of the bundle and main vessel. The first test analysed showed a first pressure peak of about 25 bar, due to pressure wave propagation at the cap rupture instant. It did not appear in the second test as consequence of a leakage from the cap before the complete rupture. The following pressurization caused by the entering of water into the reaction vessel was of an analogues magnitude for both the tests (about 30 bar). The water/LBE interaction lower temperature was reached on the inner ranks of tubes, about 160°C. The outer rank was cooled down to 340°C. The strain gage measurements showed a decreasing deformation on the tubes toward the outer positions. No ruptures were observed on tubes surrounding the injector. The amount of LBE transported into the dump tank was strongly dependent on the LBE level in the reaction tank at the start of the tests
Test section design for SGTR experimental investigation in CIRCE facility for HLMRS supported by SIMMER-III code
No full textIn the framework of MAXSIMA project, the design of a large-scale Test Section (TS), aiming to experimentally investigate the Steam Generator Tube Rupture (SGTR) postulated event in a relevant configuration for Gen IV MYRRHA reactor, was carried out. The TS will be implemented in the large pool CIRCE facility, at ENEA CR Brasimone. The TS is composed of four tube bundles representing a full scale portion of the Primary Heat eXchanger (PHX) of MYRRHA plant. They allow the execution of four SGTR tests, one at a time, excluding the necessity to extract the TS from the facility after each test. Water is foreseen to be injected at 16 bar and 200° C in the pool, partially filled by LBE at 350° C with a cover gas of argon at about 1 bar. The pressurization transients of CIRCE vessel and the sizing of the discharge lines and relative rupture disks were numerically predicted by SIMMER-III code on the base of a preliminary simplified configuration of the TS. The obtained results showed that the design pressure of CIRCE main vessel was not reached during more than 10 s of water injection, implementing a singular rupture disk having a diameter of 2 inch activated at 6.5 bar. It appears more than enough to notice, in a real reactor, the occurrence of the SGTR event and stop the water supply, interrupting the accidental scenario. These numerical results were adopted to support the design of the presented TS
Large-Scale Experimental Analysis of Steam Sub-Atmospheric Condensation for ITER Vacuum Vessel Pressure Suppression System During LOCA Event
No full textThe loss of coolant accidents (LOCA) postulated to occur in the vacuum vessel (VV) of the International Thermonuclear
Experimental Reactor (ITER) entails dangerous VV pressurization, which is foreseen to be mitigated by steam
discharge into the Vacuum Vessel Pressure Suppression System (VVPSS). This safety system realizes steam direct contact
condensation (DCC) in four Vapour Suppression Tanks (VSTs, 100 m3 each) at a very characteristic initial condition: subatmospheric cover gas pressure, for receiving steam from the VV at low pressure (0.2 ÷ 1.5 bar abs). This upper limit was defined to preserve the VV structural integrity.
The scenario of steam DCC at vacuum condition in a VVPSS
tank was extensively experimentally investigated at University of Pisa (UNIPI) Laboratory Guerrini, in a suitable Large-Scale
Test Facility (LSTF), assuming a wide range of water pool temperature, injected steam mass flow rate and water pool
pressure, characteristic for ITER VV postulated LOCA event. In addition, both a scaled and mock-up configuration of the
sparger adopted in the VVPSS were implemented and tested in the LSTF. This activity continues and builds up the UNIPI
research commitment, started with Small-Scale Test Facility projects (SSTF, condensation vessel of 4.5 m3), increasing the
condensation tank volume and steam mass flow rate to 92 m3 and 500 g/s, respectively. This effort aimed to contribute to fill the existing literature gap regarding the steam DCC at subatmospheric pressure.
An overall number of 174 steam DCC experimental tests, subdivided in four campaigns, was carried out in the LSTF for
water pool temperature ranging between 30 and 100°C, steam mass flow rate 50 ÷ 500 g/s and water pool cover gas pressure 20 ÷ 100 kPa abs. Steady state pure steam and steam plus noncondensable (air) tests were carried out with both sparger system A (scaled) and B (mock-up). In addition, pure steam transient tests were also performed adopting both spargers. The experimental results were plotted in the available condensation map, previously elaborated during SSTF activities at subatmospheric condition, verifying the instability regime domains with the support of the LSTF video acquisition system. In addition, the analysis of pressure drops across both spargers was carried out providing a relation with the injected steam mass flow rate. Finally, the condensation efficiency was evaluated for the whole test campaign showing a value next to 100% (complete condensation), with the exception of tests performed with water pool at saturation conditions (92%). This research activity permitted an extensive characterization of steam DCC regimes in vacuum conditions, occurring in ITER VVPSS during postulated LOCA scenarios, and contributed to evaluate and confirm the VVPSS safe functioning
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