Code assessment and modelling for Design Basis Accident analysis of the European Sodium Fast Reactor design. Part II: Optimised core and representative transients analysis

Lázaro Núñez, Álvaro; Schikorr, M; Mikityuk, K.; Ammirabile, Luca; Bandini, G.; Darmet, G.; Schmitt, D.; Dufour, Jean Francois; Tosello, A.; Gallego Díaz, Eduardo F.; Jiménez Varas, Gonzalo; Bubelis, E.; Ponomarev, A.; Kruessmann, R.; Struwe, D. y Stempniewicz, M. (2014). Code assessment and modelling for Design Basis Accident analysis of the European Sodium Fast Reactor design. Part II: Optimised core and representative transients analysis. "Nuclear Engineering and Design", v. 277 (n. null); pp. 265-276. ISSN 0029-5493. https://doi.org/10.1016/j.nucengdes.2014.02.029.

Descripción

Título: Code assessment and modelling for Design Basis Accident analysis of the European Sodium Fast Reactor design. Part II: Optimised core and representative transients analysis
Autor/es:
  • Lázaro Núñez, Álvaro
  • Schikorr, M
  • Mikityuk, K.
  • Ammirabile, Luca
  • Bandini, G.
  • Darmet, G.
  • Schmitt, D.
  • Dufour, Jean Francois
  • Tosello, A.
  • Gallego Díaz, Eduardo F.
  • Jiménez Varas, Gonzalo
  • Bubelis, E.
  • Ponomarev, A.
  • Kruessmann, R.
  • Struwe, D.
  • Stempniewicz, M.
Tipo de Documento: Artículo
Título de Revista/Publicación: Nuclear Engineering and Design
Fecha: Octubre 2014
Volumen: 277
Materias:
Escuela: E.T.S.I. Industriales (UPM)
Departamento: Ingeniería Energética
Licencias Creative Commons: Reconocimiento - Sin obra derivada - No comercial

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Resumen

The new reactor concepts proposed in the Generation IV International Forum require the development and validation of computational tools able to assess their safety performance. In the first part of this paper the models of the ESFR design developed by several organisations in the framework of the CP-ESFR project were presented and their reliability validated via a benchmarking exercise. This second part of the paper includes the application of those tools for the analysis of design basis accident (DBC) scenarios of the reference design. Further, this paper also introduces the main features of the core optimisation process carried out within the project with the objective to enhance the core safety performance through the reduction of the positive coolant density reactivity effect. The influence of this optimised core design on the reactor safety performance during the previously analysed transients is also discussed. The conclusion provides an overview of the work performed by the partners involved in the project towards the development and enhancement of computational tools specifically tailored to the evaluation of the safety performance of the Generation IV innovative nuclear reactor designs.

Proyectos asociados

TipoCódigoAcrónimoResponsableTítulo
FP7EC/FP7/232658/EUCP-ESFRSin especificarCollaborative Project on European Sodium Fast Reactor

Más información

ID de Registro: 35677
Identificador DC: http://oa.upm.es/35677/
Identificador OAI: oai:oa.upm.es:35677
Identificador DOI: 10.1016/j.nucengdes.2014.02.029
URL Oficial: http://www.sciencedirect.com/science/journal/00295493/277
Depositado por: Memoria Investigacion
Depositado el: 29 Sep 2015 16:35
Ultima Modificación: 01 Mar 2017 16:28
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