Results 111 to 120 of about 5,558 (186)
Basis for Coupled 3-D Neutronics-Thermal-Hydraulics [PDF]
, 2008 The purpose of this seminar is first to discuss the basis of the coupling between 3-D Neutron- Kinetics and Thermal-Hydraulics codes, including the control and 3-D variables to interchange, the transform of the 3-D NK and TH core nodalizations, and the ...
Aragonés Beltrán, José María
core
Thermal hydraulic simulations of the Angra 2 PWR
EPJ Nuclear Sciences & Technologies, 2015 Angra 2, the second Brazilian nuclear power plant, began the commercial operation in 2001. The plant is a pressurized water reactor (PWR) type with electrical output of about 1350 MW.
González-Mantecón Javier +6 more
doaj +1 more source
EnergiesThe RELAP5 code is a computational tool designed for transient simulations of light water reactor coolant systems under hypothesized accident conditions.
Young Hwan Lee +2 more
doaj +1 more source
Science and Technology of Nuclear Installations, 2010 RELAP5/SCDAPSIM and MAAP4 are two widely used severe accident computer codes for the integral analysis of the core and the reactor pressure vessel behaviour following the core degradation.
Siniša Šadek +2 more
doaj +1 more source
Estudi termoenergètic i dinàmic de transitoris de Centrals Nuclears amb finalitat de seguretat. Optimització de la sistemàtica de càlcul i organització de resultats [PDF]
, 2003 L’objectiu principal d’aquest projecte és l’obtenció d’una biblioteca de casos que contingui de forma estructurada la documentació i els resultats dels estudis dels principals transitoris, tant reals com hipotètics, de les Centrals Nuclears d’Ascó i ...
Gavaldà Sobrepere, Sònia +1 more
core
Peningkatan Kinerja Sistem Keselamatan Pasif Pada Reaktor Nuklir Dengan Penambahan Komponen Rvacs [PDF]
, 2014 Completeness of passive safety systems and inherent in advanced reactors is a major prerequisite. This paper explores the results of a conceptual design of the heat removal system at the nuclear power plant (NPP) type Very High-Temperature Reactor.
Abdullah, A. G. (A) +2 more
core +1 more source
First Experience with the Consolidation of WWER Reactor Pressure Vessel Knowledge through a New Method [PDF]
, 2008 One of today¿s activities of the Joint Research Centre¿s (JRC) Institute for Energy (IE) concerns data management and dissemination in nuclear safety. An ¿Online Data & Information Network¿ (ODIN) is set-up, which maintains one document database and four
BRUMOVSKY Milan +3 more
core
Computation of Gap Conductance in Different Fuel Assemblies in VVER-1000 Type Reactors [PDF]
مجله علوم و فنون هستهای, 2011 In this paper, a calculation for fresh fuels gap conductance at different axial lengths of fuel assemblies of the VVER-1000 type reactors has been made using two models of Calza-Bini and Relap5.
M Rahgoshay, KH Shokri
doaj
A study of the dispersed flow interfacial heat transfer model of RELAP5/MOD2.5 and RELAP5/MOD3
, 1995 The model of interfacial heat transfer for the dispersed flow regime used in the RELAP5 computer codes is investigated in the present paper. Short-transient calculations of two low flooding rate tube reflooding experiments have been performed, where the hydraulic conditions and the heat input to the vapour in the post-dryout region were controlled for ...
M Andreani, G T Analytis, S N Aksan
openaire +2 more sources
Risk-Informed design process of the IRIS reactor [PDF]
, 2005 Westinghouse is currently conducting the pre-application licensing of the International Reactor Innovative and Secure (IRIS). The design philosophy of the IRIS has been based on the concept of Safety-by-DesignTM and within this framework the PSA is ...
A. MAIOLI +5 more
core