Accident Management Actions In An Upper-Head Small-Break Loss-Of-Coolant Accident With High-Pressure Safety Injection Failed

Queral Salazar, José Cesar; Gonzalez Cadelo, Juan; Jiménez Varas, Gonzalo y Villaba, Ernesto (2011). Accident Management Actions In An Upper-Head Small-Break Loss-Of-Coolant Accident With High-Pressure Safety Injection Failed. "Nuclear Technology", v. 175 (n. 3); pp. 572-593. ISSN 0029-5450.

Descripción

Título: Accident Management Actions In An Upper-Head Small-Break Loss-Of-Coolant Accident With High-Pressure Safety Injection Failed
Autor/es:
  • Queral Salazar, José Cesar
  • Gonzalez Cadelo, Juan
  • Jiménez Varas, Gonzalo
  • Villaba, Ernesto
Tipo de Documento: Artículo
Título de Revista/Publicación: Nuclear Technology
Fecha: Septiembre 2011
Volumen: 175
Materias:
Escuela: E.T.S.I. Minas (UPM) [antigua denominación]
Departamento: Sistemas Energéticos [hasta 2014]
Licencias Creative Commons: Reconocimiento - Sin obra derivada - No comercial

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Resumen

Since the Three Mile Island accident, an important focus of pressurized water reactor (PWR) transient analyses has been a small-break loss-of-coolant accident (SBLOCA). In 2002, the discovery of thinning of the vessel head wall at the Davis Besse nuclear power plant reactor indicated the possibility of an SBLOCA in the upper head of the reactor vessel as a result of circumferential cracking of a control rod drive mechanism penetration nozzle - which has cast even greater importance on the study of SBLOCAs. Several experimental tests have been performed at the Large Scale Test Facility to simulate the behavior of a PWR during an upper-head SBLOCA. The last of these tests, Organisation for Economic Co-operation and Development Nuclear Energy Agency Rig of Safety Assessment (OECD/NEA ROSA) Test 6.1, was performed in 2005. This test was simulated with the TRACE 5.0 code, and good agreement with the experimental results was obtained. Additionally, a broad analysis of an upper-head SBLOCA with high-pressure safety injection failed in a Westinghouse PWR was performed taking into account different accident management actions and conditions in order to check their suitability. This issue has been analyzed also in the framework of the OECD/NEA ROSA project and the Code Applications and Maintenance Program (CAMP). The main conclusion is that the current emergency operating procedures for Westinghouse reactor design are adequate for these kinds of sequences, and they do not need to be modified.

Más información

ID de Registro: 11398
Identificador DC: http://oa.upm.es/11398/
Identificador OAI: oai:oa.upm.es:11398
URL Oficial: http://www.new.ans.org/store/j_12507
Depositado por: Memoria Investigacion
Depositado el: 25 Jul 2012 12:31
Ultima Modificación: 01 Mar 2017 16:37
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