1,721,260 research outputs found
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. Copyright © 2016 by ASME
Thermal-hydraulic code-to-code benchmark in a simplified EBR-II geometry
No full text[No abstract available
Caratterizzazione sperimentale dell'interazione metallo liquido pesante - acqua e qualifica rottura generatore di vapore LFR
Full text linkIl presente lavoro di ricerca è relativo allo studio dell'interazione metallo liquido pesante-acqua, d'interesse per le analisi di sicurezza dei reattori veloci di quarta generazione refrigerati con piombo liquido. La prima parte del lavoro ha riguardato il supporto fornito all'ENEA per la scelta delle modifiche da apportare all'apparecchiatura LIFUS 5. Successivamente, è stata effettuata una serie di simulazioni di pre-test per valutare l'influenza di alcuni parametri operativi sul fenomeno dell'interazione, in modo da dare supporto alla scelte delle modifiche da effettuare e alle condizioni al contorno da definire per le future prove sperimentali. Le modifiche effettuate per aggiornare l'apparecchiatura sperimentale LIFUS5 sono descritte distinguendo tra modifiche al "layout" dell'apparecchiatura (incluse le strutture di supporto), alla strumentazione, al sistema di controllo e ai sistemi di interfaccia
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. Copyright © 2016 by ASME
Simulation study of pressure trends in the case of loss of coolant accident in Water Cooled Lithium Lead blanket module
No full textThe water-lithium lead interaction implies a direct energy release, which leads to temperature and pressure increase, due to a combined thermal and chemical reaction, and an indirect form of energy release, the hydrogen production, due to secondary chemical reaction involving the initial reaction products. Review and understanding of the knowledge acquired in past studies, experimental works and numerical activities are needed in view of the renewed interest in the Water Cooled Lithium Lead blanket concept and safety issues connected with the fusion reactor design. This paper presents a review of the studies carried out in the past to characterize the potential safety concerns associated with the use of water and lithium-lead eutectic alloy, the main experimental campaigns, and numerical simulations of BLAST Test No. 5 performed by SIMMER-III code. As results, no code was found able to perform a satisfactory post-test analysis of separate effect experiments, without engineering assumptions. Therefore, a code model for the exothermic reaction and hydrogen production, and experimental data are needed for solving the WCLL blanket safety issues associated with the water-PbLi interaction. © 2015 Elsevier B.V. All rights reserved
Qualification of a RELAP5-3D © system code nodalization of EBR-II
No full textENEA is setting up, applying and validating numerical models and an integrated multi-physics approach, based on existing codes and aimed at supporting safety analysis of liquid metal Gen. IV fast reactors. The paper provides an outline of the activity and it is focused on qualification of a three-dimensional thermal- hydraulic model of EBR-II primary and secondary systems using the system code RELAP5-3DO. The nodalization models one by one the fuel assemblies of the core and of the extended core of the reactor for an easy and efficient coupling with a 3D neutron kinetic code. The paper presents the qualification of the nodalization based on the EBR-II test SHRT-17. The analysis of the experimental data, the identification of the thermal-hydraulics phenomena observed in the tests are the basis for assessing the code performances and for discussing its limitations. © Copyright (2015) by American Nuclcar Society All rights reserved
Preliminary system modeling for the EUROfusion water cooled lithium lead blanket
No full textThe Water Cooled Lithium Lead (WCLL) blanket is one of the four breeder blanket technologies under consideration within the framework of the EUROfusion Consortium activities. The aim of this work is to develop a preliminary model that can track tritium concentration and tritium fluxes along each part of the WCLL blanket and its ancillary systems at any time. Because of tritium's nature, the phenomena of diffusion, dissociation, recombination and solubilization have been taken into account when describing the tritium behavior inside the lead-lithium channels, the structural materials and the water coolant circuits. The simulations have been performed using the object oriented modeling software EcosimPro. Results have been obtained for the pulsed generation scenario of the European demonstration power plant (DEMO). The tritium inventory in every part of the blanket has been computed. Permeation rates have been calculated as well allowing to know how much tritium ends up in the coolant system and how much remains in the liquid metal. The amount of tritium extracted from the lead-lithium loop has been also obtained. All this information allows having a global perspective of tritium behavior all over the blanket at any time. The model provides valuable information for the design of the WCLL blanket. More complex upgrades are planned to be implemented based on this model in future stages of the EUROfusion project. © American Nuclear Society
IAEA's coordinated research project on EBR-II Shutdown heat removal tests
No full textA Coordinated Research Project (CRP) on “Benchmark Analysis of EBR-II Shutdown Heat Removal Tests (SHRT)” was launched by the International Atomic Energy Agency (IAEA) in 2012. A series of transient tests were conducted on the EBR-II reactor at Argonne National Laboratory (ANL) to improve the understanding of thermal hydraulics and neutronics of fast reactors. Shutdown heat removal tests conducted in 1984 and 1986 demonstrated mechanisms by which fast reactors can survive severe accident initiators with no core damage. Two SHRT tests, SHRT-17 representing Protected Loss of Flow (PLOF) transients and SHRT-45R representing Unprotected Loss of Flow (ULOF) transients, were studied in the IAEA CRP. The objectives of the CRP were to improve design and simulation capabilities in fast reactor thermal hydraulics, neutronics and safety analyses through benchmark analysis of these two important tests. At the first stage of the benchmark, ANL provided the input data on EBR-II geometry, as well as initial and boundary conditions for the SHRT-17 and SHRT-45R tests to perform “blind” calculations. At the second stage, ANL released the experimental observations and participants had the chance to analyze the difference and refine the models. At the third stage, a methodology to systematically analyze and compare the models and the results of each participant was applied. Nineteen organizations from eleven countries participated in the CRP, making it one of the largest CRP coordinated by the IAEA fast reactor team. The paper provides a general CRP overview, gives the basics of the EBR-II reactor design, describes the shutdown heat removal tests, the benchmark setup, and results of numerical simulations, followed by a detailed discussion on the EBR-II CRP. © 2016 Association for Computing Machinery Inc. All Rights Reserved
Qualification of TRACE V5.0 code against fast cooldown transient in the pkl-iii integral test facility
No full textThe present paper deals with the analytical study of the PKL experiment G3.1 performed using the TRACE code (version 5.0 patch1). The test G3.1 simulates a fast cooldown transient, namely, a main steam line break. This leads to a strong asymmetry caused by an increase of the heat transfer from the primary to the secondary side that induces a fast cooldown transient on the primary side-affected loop. The asymmetric overcooling effect requires an assessment of the reactor pressure vessel integrity considering PTS (pressurized thermal shock) and an assessment of potential recriticality following entrainment of colder water into the core area. The aim of this work is the qualification of the heat transfer capabilities of the TRACE code from primary to secondary side in the intact and affected steam generators (SGs) during the rapid depressurization and the boiloff in the affected SG against experimental data. © 2013 Eugenio Coscarelli et al
Experimental and numerical investigation of double wall bayonet tubes performances in pool type integral test facility
No full textENEA has designed and manufactured an experimental mock-up (called Heavy liquid metal - pressurized water cooled tube - HERO) with seven instrumented full scale double walled bayonet tubes installed in CIRCE facility in order to provide integral test experiments which are representative of the ALFRED Steam Generator. The present paper summarizes the activities performed in support to the experimental campaigns in the HERO-CIRCE facility. In particular, preliminary calculations by means of RELAP-5 and pretest-experimental campaigns are given in this document
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