Code assessment and modelling for Design Basis Accident Analysis of the European sodium fast reactor design. Part I: System description, modelling and benchmarking

Lázaro Núñez, Álvaro; Ammirabile, Luca; Bandini, G.; Darmet, G.; Massara, S.; Dufour, Jean Francois; Tosello, A.; Gallego Díaz, Eduardo F.; Jiménez Varas, Gonzalo; Mikityuk, K.; Schikorr, M.; Bubelis, E.; Ponomarev, A.; Kruessmann, R. y Stempniewicz, M. (2014). Code assessment and modelling for Design Basis Accident Analysis of the European sodium fast reactor design. Part I: System description, modelling and benchmarking. "Nuclear Engineering and Design", v. 266 (n. null); pp. 1-16. ISSN 0029-5493. https://doi.org/10.1016/j.nucengdes.2013.10.019.

Descripción

Título: Code assessment and modelling for Design Basis Accident Analysis of the European sodium fast reactor design. Part I: System description, modelling and benchmarking
Autor/es:
  • Lázaro Núñez, Álvaro
  • Ammirabile, Luca
  • Bandini, G.
  • Darmet, G.
  • Massara, S.
  • Dufour, Jean Francois
  • Tosello, A.
  • Gallego Díaz, Eduardo F.
  • Jiménez Varas, Gonzalo
  • Mikityuk, K.
  • Schikorr, M.
  • Bubelis, E.
  • Ponomarev, A.
  • Kruessmann, R.
  • Stempniewicz, M.
Tipo de Documento: Artículo
Título de Revista/Publicación: Nuclear Engineering and Design
Fecha: Enero 2014
Volumen: 266
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 (GIF) are conceived to improve the use of natural resources, reduce the amount of high-level radioactive waste and excel in their reliability and safe operation. Among these novel designs sodium fast reactors (SFRs) stand out due to their technological feasibility as demonstrated in several countries during the last decades. As part of the contribution of EURATOM to GIF the CP-ESFR is a collaborative project with the objective, among others, to perform extensive analysis on safety issues involving renewed SFR demonstrator designs. The verification of computational tools able to simulate the plant behaviour under postulated accidental conditions by code-to-code comparison was identified as a key point to ensure reactor safety. In this line, several organizations employed coupled neutronic and thermal-hydraulic system codes able to simulate complex and specific phenomena involving multi-physics studies adapted to this particular fast reactor technology. In the “Introduction” of this paper the framework of this study is discussed, the second section describes the envisaged plant design and the commonly agreed upon modelling guidelines. The third section presents a comparative analysis of the calculations performed by each organisation applying their models and codes to a common agreed transient with the objective to harmonize the models as well as validating the implementation of all relevant physical phenomena in the different system codes.

Proyectos asociados

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

Más información

ID de Registro: 35676
Identificador DC: http://oa.upm.es/35676/
Identificador OAI: oai:oa.upm.es:35676
Identificador DOI: 10.1016/j.nucengdes.2013.10.019
URL Oficial: http://www.sciencedirect.com/science/article/pii/S0029549313005736
Depositado por: Memoria Investigacion
Depositado el: 28 Sep 2015 16:53
Ultima Modificación: 01 Mar 2017 16:30
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