Ara es mostren els items 1-12 de 12

    • Application of a BEPU-based code assessment to the ATLAS upper head SB-LOCA test 

      Al Awad, Abdulrahman; Freixa Terradas, Jordi; Pérez-Ferragut, Marina (2021-12-15)
      Article
      Accés obert
      The prevailing state of knowledge of two-phase flow in complex systems, as in the nuclear field, usually leads a certain degree of ad hoc calibration of computational models and the consequent inevitable subjectivity in ...
    • Aproximación multiescala a la simulación termohidráulica de reactores de fusión en la Universidad Politécnica de Cataluña 

      Batet Miracle, Lluís; Mas de les Valls Ortiz, Elisabet; Osychenko, Oleg; Martínez Quiroga, Víctor Manuel; Pérez-Ferragut, Marina; Freixa Terradas, Jordi; Reventós Puigjaner, Francesc Josep (2014)
      Text en actes de congrés
      Accés restringit per política de l'editorial
      El Grupo de Estudios Termohidráulicos (GET) de la UPC posee gran experiencia en estudios de seguridad nuclear mediante códigos de simulación de plantas nucleares. Desde 1987, el grupo ha colaborado con las CN de Ascó y ...
    • Assessment of SBO Fukushima likewise scenario for an IPWR design with RELAP5MOD33 and RELAPSCDAPSIMMOD3. 5 codes 

      Martínez Quiroga, Víctor Manuel; Pérez-Ferragut, Marina; Pericas Casals, Raimon; Freixa Terradas, Jordi (2022)
      Text en actes de congrés
      Accés obert
      In recent years Small Modular Reactors (SMR) have become very popular within the nuclear industry. These designs allow to reduce costs as well as to enhance the safety due to passive nuclear safety features. Within these ...
    • Development of flow regime maps for lead lithium eutectic–helium flows 

      Mas de les Valls Ortiz, Elisabet; Cegielski, Andrzej; Jaros, Mark; Pérez-Ferragut, Marina; Batet Miracle, Lluís; Freixa Terradas, Jordi (2020-09-01)
      Article
      Accés obert
      nstitute for Plasma Research (IPR), Gandhinagar (India) is currently involved in the design and development of its Lead-Lithium Ceramic Breeder (LLCB) module for testing in the International Thermo-nuclear Experimental ...
    • Linden code validation against NUPEC / PSBT experimental data for void fraction and temperature benchmarks 

      Pérez-Ferragut, Marina; Freixa Terradas, Jordi; Martínez Quiroga, Víctor Manuel; Casamor Vidal, Max (2022)
      Text en actes de congrés
      Accés restringit per política de l'editorial
      The subchannel analysis code LINDEN is being developed by China Nuclear Power Technology Research Institute Co. Ltd (CNPRI). The LINDEN code is used in thermal-hydraulics design and safety analysis of pressurized water ...
    • Methodology for phenomenological code assessment with integral test data 

      Pérez-Ferragut, Marina; Martínez Quiroga, Víctor Manuel; Casamor Vidal, Max; Freixa Terradas, Jordi (2022-02-28)
      Article
      Accés obert
      The use of codes in the licensing process requires a rigorous validation process that can be accomplished by means of qualitative and quantitative assessment. In thermal hydraulics, this validation has to be performed at ...
    • Modelling of a supercritical CO2 power cycle for nuclear fusion reactors using RELAP5-3D 

      Batet Miracle, Lluís; Álvarez Fernández, Josep Maria; Mas de les Valls Ortiz, Elisabet; Martínez Quiroga, Víctor Manuel; Pérez-Ferragut, Marina; Reventós Puigjaner, Francesc Josep; Sedano Miguel, Luis Ángel (2014-04-01)
      Article
      Accés restringit per política de l'editorial
      A supercritical recompression CO2 power cycle has been simulated using the system code RELAP5-3D. This code is being developed by INL and has traditionally been used in the simulation of operational and accidental transients ...
    • RELAP/SCDAPSIM/MOD4.0 modification for transient accident scenario of test blanket modules involving helium flows into heavy liquid metal 

      Pérez-Ferragut, Marina; Freixa Terradas, Jordi; Mas de les Valls Ortiz, Elisabet (2015)
      Comunicació de congrés
      Accés restringit per política de l'editorial
      The Institute for Plasma Research (IPR), India, is currently involved in the design and development of its Test Blanket Module (TBM) for testing in ITER (International Thermo nuclear Experimental Reactor). The Indian TBM ...
    • Revisiting ISP-13 with RELAP/SCDAPSIM/MOD3.5 using core SCDAP components 

      Freixa Terradas, Jordi; Pérez-Ferragut, Marina; Reventós Puigjaner, Francesc Josep; Allison, Chris M. (2015)
      Comunicació de congrés
      Accés restringit per política de l'editorial
      The recent accident in the Fukushima Daiichi nuclear power plant opened a discussion on severe accident management that includes the analysis of the accident by means of computational tools that can predict the core behavior ...
    • Significance of the input parameters selection and the nodalization qualification in the final results of an IBLOCA BEPU calculation 

      Martínez Quiroga, Víctor Manuel; Freixa Terradas, Jordi; Pérez-Ferragut, Marina; Reventós Puigjaner, Francesc Josep (2016)
      Text en actes de congrés
      Accés restringit per política de l'editorial
      In the framework of Design Basis Accidents (DBA) for Pressurized Water Reactor (PWR), and based on recent studies on pipe integrity combined with the Risk-Informed Decision Making (RIDM), the USNRC proposed the Intermediate ...
    • Spanish contribution to the development and application of best estimate plus uncertainty methodologies: past, present and future 

      Freixa Terradas, Jordi; Barrachina Celda, Teresa María; Berna Escriche, Cesar; Bocanegra Melian, Rafael; Carlos Alberola, Sofía; Castro Gonzalez, Emilio; Cuervo Gomez, Diana; Durán Vinuesa, Luis Felipe; Escrivá Castells, Facundo Alberto; Feria Márquez, Francisco; Fernández Cosials, Kevin; Martínez Quiroga, Víctor Manuel; Pérez-Ferragut, Marina; Pericas Casals, Raimon; Reventós Puigjaner, Francesc Josep (Elsevier, 2024-02-01)
      Article
      Accés restringit per política de l'editorial
      Best-Estimate Plus Uncertainty (BEPU) nuclear safety analysis is an approach widely used in the field of nuclear engineering and reactor safety assessment. It aims to provide a more realistic assessment of reactor safety ...
    • Uncertainty and sensitivity analysis of the LOFT L2-5 test: Results of the BEMUSE programme 

      de Crécy, Agnès; Bazin, P.; Glaeser, H.; Joucla, J.; Probst, P.; Chung, B.; Oh, D.Y.; Kyncl, M.; Pernica, R.; Batet Miracle, Lluís; Pérez-Ferragut, Marina; Reventós Puigjaner, Francesc Josep (2008-12)
      Article
      Accés restringit per política de l'editorial
      This paper presents the results and the main lessons learnt from the phase 3 of BEMUSE, an international benchmark activity sponsored by the Committee on the Safety of Nuclear Installations [CSNI: Committee on the Safety ...