Heavy liquid metal natural circulation in a one-dimensional loop (original) (raw)
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Natural circulation in a liquid metal one-dimensional loop
Journal of Nuclear Materials, 2008
a r t i c l e i n f o a b s t r a c t A wide use of pure lead, as well as its alloys (such as lead-bismuth, lead-lithium), is foreseen in several nuclear-related fields: it is studied as coolant in critical and sub-critical nuclear reactors, as spallation target for neutron generation in several applications and for tritium generation in fusion systems. In this framework, a new facility named NAtural CIrculation Experiment (NACIE), has been designed at ENEA-Brasimone Research Centre. NACIE is a rectangular loop, made by stainless steel pipes. It consists mainly of a cold and hot leg and an expansion tank installed on the top of the loop. A fuel bundle simulator, made by three electrical heaters placed in a triangular lattice, is located in the lower part of the cold leg, while a tube in tube heat exchanger is installed in the upper part of the hot leg. The adopted secondary fluid is THT oil, while the foreseen primary fluid for the tests is lead-bismuth in eutectic composition (LBE). The aim of the facility is to carry out experimental tests of natural circulation and collect data on the heat transfer coefficient (HTC) for heavy liquid metal flowing through rod bundles. The paper is focused on the preliminary estimation of the LBE flow rate along the loop. An analytical methodology has been applied, solving the continuity, momentum and energy transport equations under appropriate hypothesis. Moreover numerical simulations have been performed. The FLUENT 6.2 CFD code has been utilized for the numerical simulations. The main results carried out from the pre-tests simulations are illustrated in the paper, and a comparison with the theoretical estimations is done.
Natural circulation studies in a lead bismuth eutectic loop
Lead Bismuth Eutectic (LBE) is increasingly getting more attraction as the coolant for advanced reactor systems. It is also the primary coolant of the Compact High Temperature Reactor (CHTR), being designed at BARC. A loop has been set up for thermal hydraulics, instrument development and material related studies relevant to CHTR. Steady state natural circulation experimental studies were carried out for different power levels. Transient studies for start-up of natural circulation in the loop, loss of heat sink and step power change have also been carried out. An 1D code named LeBENC has been developed at BARC to simulate the natural circulation characteristics in closed loops. The salient features of the code include ability to handle non-uniform diameter components, axial thermal conduction in fluid and heat losses from the piping to the environment. This paper deals with the experimental studies carried out in the loop. Detailed validation of the LeBENC code with the experimental data is also discussed in the paper.
Study of the Natural Circulation Phenomenon for Nuclear Reactors
2011
The objective of this paper is to study the natural circulation phenomenon in one and two-phase regime. There has been a crescent interest in the scientific community in the study of the natural circulation. New generation of compact nuclear reactors uses the natural circulation for residual heat removal in case of accident or shutdown. For this study, the modeling and the simulation of the experimental circuit is performed with the RELAP5 code. The theoretical results showed to be satisfactory when compared with the experimental ones.
Numerical simulation of natural circulation experiments conducted at the IIST facility
Nuclear Engineering and Design, 1994
Natural circulation plays an important role in long-term cooling of pressurized water reactors (PWRs) under small break loss-of-coolant accidents. Recently, natural circulation experiments have been conducted at the Institute of Nuclear Energy Research integral system test (IIST) facility, which is used to simulate the Westinghouse three-loop Maanshan PWR. A numerical simulation is presented to investigate the natural circulation phenomena of the IIST facility with the RELAP5/MOD3 code. The calculated results are in good agreement with the experimental data of the single-phase natural circulation both quantitatively and qualitatively. The influences of power level and system pressure on natural circulation can also be predicted by the current model. Based on the two-phase natural circulation data, the calculated flow rate history is similar to that obtained from the experiment.
Natural circulation data and methods for advanced water cooled nuclear power plants designs
2002
The use of particular designations of countries or territories does not imply any judgement by the publisher, the IAEA, as to the legal status of such countries or territories, of their authorities and institutions or of the delimitation of their boundaries. The mention of names of specific companies or products (whether or not indicated as registered) does not imply any intention to infringe proprietary rights, nor should it be construed as an endorsement or recommendation on the part of the IAEA.
Preliminary numerical analysis of the flow distribution in the core of a research reactor
2020
The thermal-hydraulic safety analysis of research reactors establishes the safety criteria to ensure the integrity of the fuel elements in the reactor core. It assures that all core components are being adequately cooled during operation. It is necessary to know if the average mass flow rate (and their standard deviation) among the fuel assemblies are enough to cool the power generated during operation. Once satisfied such condition, it allows the calculation of the maximum heat flux transferred from fuel assemblies to the coolant, and if the maximum cladding temperatures are below the limits set by the safety criteria. Among the objectives, this study presents a methodology for a preliminary threedimensional numerical analysis of the flow distribution in the core of the IEA-R1 research reactor, under steady state condition. For this, the ANSYS-CFX commercial code was used to analyze the flow dynamics in the core, and to visualize the velocity field. It was possible to conclude that...
Progress in Nuclear Energy, 2015
Helium experimental loop High temperature reactor Gas fast reactor High temperature materials a b s t r a c t This paper focuses on the research infrastructure of advanced gas cooled reactors in the Czech Republic, particularly on the high-temperature helium loop HTHL, which is a unique facility of its kind. HTHL is intended mainly for testing structural materials. It also can be used to research technologies relating to helium coolant. The maximum temperature and pressure that can be used within the specimen testing space are 900 C and 7 MPa, respectively, and the maximum gas flow rate in the main loop is 38 kg/hr. Originally, the equipment was envisaged as a device for corrosion tests of materials in the reactor LVR-15 but, according to current plans, a different equipment will be built for this purpose within the frame of the SUSEN project. At the same time, an additional helium loop (S-Allegro) will be built to test selected components of advanced gas-cooled reactors.