Search results for "Anti nuclear"
showing 10 items of 415 documents
Analyses of single- and two-phase flow pressure drops in helical pipes using a modified RELAP5 code
2012
Abstract Thermal fluid dynamics analyses on single- and two-phase flows in helical pipes of steam generators to be used in Generation III and IV nuclear reactors have been performed. The study concerned with experimental activities as well code simulations carried out in the framework of a collaboration between the Department of Energetics of the Polytechnic of Torino and the Department of Energy of the University of Palermo. The goal was the validation of models implemented in Relap5/Mod3.2β thermal–hydraulic advanced code to simulate the hydrodynamic behaviour of helical pipe components in spite of the one-dimensional nature of the code. It is shown that much of the experimental data obta…
A semi-empirical approach for predicting two-phase flow discharge through branches of various orientations connected to a horizontal main pipe
2010
Abstract The subdivision of two-phase flow in branching conduits consisting of a large horizontal main pipe with upward, downward, or lateral branches of reduced diameter is of great interest in various technological fields. For example, these conduits are important in light-water nuclear reactors (LWRs) in the case of a small break loss-of-coolant accident (SBLOCA) in a leg of the reactor's primary coolant loops, as well as for breaks or valve malfunctions in a large pipeline. In these kinds of circumstances, the relevant phenomenology often involves phase stratification coupled with possible liquid entrainment or gas pool-through phenomena. Therefore, these phenomena were studied in depth…
Sensitivity Analysis of the MASLWR Helical Coil Steam Generator Using TRACE
2011
Accurate simulation of transient system behavior of a nuclear power plant is the goal of systems code calculations, and the evaluation of a code's calculation accuracy is accomplished by assessment and validation against appropriate system data. These system data may be developed either from a running system prototype or from a scaled model test facility, and characterize the thermal hydraulic phenomena during both steady state and transient conditions. The identification and characterization of the relevant thermal hydraulic phenomena, and the assessment and validation of thermal hydraulic systems codes, has been the objective of multiple international research programs. The validation and…
Thermo-mechanical testing of Li-ceramic for the helium cooled pebble bed (HCPB) breeding blanket
2004
The helium cooled pebble bed (HCPB) Test blanket module (TBM) for the DEMO Reactor foresees the utilization of lithiate ceramics as breeder in form of pebble beds. The pebbles are organized in several layers alternatively stacked among couples of cooling plates (CP). ENEA has launched an experimental programme for the out-of-pile thermo-mechanical testing of mock-ups simulating a portion of the HCPB-TBM. The programme foresees the fabrication and testing of different mock-ups, to be tested in the HE-FUS3 facility at ENEA Brasimone. The paper describes the HELICHETTA III campaign carried-out in 2003. In particular, the test section layout, the pebble filling procedure, the experimental set-u…
Analytical and Numerical Assessment of Thermally Induced Pressure Waves in the IFMIF-DONES Liquid-Lithium Target
2020
The intended steady-state operation conditions of the International Fusion Materials Irradiation Facility-DEMO Oriented Neutron Source (IFMIF-DONES) target system are based on the D+ beam stationary running at full nominal power (5 MW). Nevertheless, critical situations can occur in the case of unavoidable sudden events like beam trips. The instantaneous variation in the heating power deposited in lithium when the beam is rapidly switched between ON-and OFF-states leads to thermal expansion, which is compensated by the compression of the target material, resulting in locally high pressures and a pressure wave propagating through the target toward the back wall. Besides the tensile stress of…
Nuclear Analysis of an ITER Blanket Module
2013
ITER blanket system is the reactor’s plasma-facing component, it is mainly devoted to provide the thermal and nuclear shielding of the Vacuum Vessel and external ITER components, being intended also to act as plasma limiter. It consists of 440 individual modules which are located in the inboard, upper and outboard regions of the reactor. In this paper attention has been focused on to a single outboard blanket module located in the equatorial zone, whose nuclear response under irradiation has been investigated following a numerical approach based on the Monte Carlo method and adopting the MCNP5 code. The main features of this blanket module nuclear behaviour have been determined, paying part…
Overview of the HCPB Research Activities in EUROfusion
2018
In the framework of the EUROfusion’s Power Plant Physics and Technology, the working package breeding blanket (BB) aims at investigating four different BB concepts for an EU demonstration fusion reactor (DEMO). One of these concepts is the helium-cooled pebble bed (HCPB) BB, which is based on the use of pebble beds of lithiated ternary compounds and Be or beryllides as tritium breeder and multiplier materials, respectively, EUROFER97 as structural steel and He as coolant. This paper aims at giving an overview of the EU HCPB BB Research and Development (R&D) being developed at KIT, in collaboration with Wigner-RCP, BUTE-INT, and CIEMAT. The paper gives an outline of the HCPB BB design evolut…
Fuzzy modelling of HEART methodology: application in safety analyses of accidental exposure in irradiation plants
2009
The present paper refers to the obtained results by using Fuzzy Fault Tree analyses of accidental scenarios which entail the potential exposure of operators working in irradiation industrial plants. For these analyses the HEART methodology, a first generation of the Human Reliability Analysis method, has been employed to evaluate the probability of human erroneous actions. This technique has been modified by us on the basis of fuzzy set concept to more directly take into account the uncertainties of the so called error-promoting factors, on which the method is grounded. The results allow also to provide some recommendations on procedures and safety equipments to reduce the radiological expo…
On the effects of the supporting frame on the radiation-induced damage of HCLL-TBM structural material
2007
Within the European Fusion Technology Programme, research activities have been conducted on the Helium-Cooled Lithium Lead (HCLL) breeding blanket concept with the aim of manufacturing a Test Blanket Module (TBM) to be irradiated in ITER. HCLL-TBM is planned to be located in an ITER equatorial port, housed inside an AISI 316 stainless steel-supporting frame. Since that frame has been designed to provide two positions separated by a dividing plate and the HCLL-TBM is expected to fill one of them, its nuclear response could vary depending on the filling status of the other position and on the plate thickness. A parametric study has been carried out to investigate the potential effects on the …
Hydraulic analysis of EU-DEMO divertor plasma facing components cooling circuit under nominal operating scenarios
2019
Within the framework of the Work Package DIV 1 – “Divertor Cassette Design and Integration” of the EUROfusion action, a research campaign has been jointly carried out by University of Palermo and ENEA to investigate the steady state thermal-hydraulic behaviour of the DEMO divertor cassette cooling circuit, focussing the attention on its Plasma Facing Components (PFCs). The research campaign has been carried out following a theoretical-computational approach based on the Finite Volume Method and adopting the commercial Computational Fluid-Dynamic code ANSYS-CFX. A realistic model of the PFCs cooling circuit has been analysed, specifically embedding each Plasma Facing Unit (PFU) cooling chann…