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

Queral Salazar, José Cesar and Gonzalez Cadelo, Juan and Jiménez Varas, Gonzalo and 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.

Description

Title: Accident Management Actions In An Upper-Head Small-Break Loss-Of-Coolant Accident With High-Pressure Safety Injection Failed
Author/s:
  • Queral Salazar, José Cesar
  • Gonzalez Cadelo, Juan
  • Jiménez Varas, Gonzalo
  • Villaba, Ernesto
Item Type: Article
Título de Revista/Publicación: Nuclear Technology
Date: September 2011
ISSN: 0029-5450
Volume: 175
Subjects:
Faculty: E.T.S.I. Minas (UPM)
Department: Sistemas Energéticos [hasta 2014]
Creative Commons Licenses: Recognition - No derivative works - Non commercial

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Abstract

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.

More information

Item ID: 11398
DC Identifier: http://oa.upm.es/11398/
OAI Identifier: oai:oa.upm.es:11398
Official URL: http://www.new.ans.org/store/j_12507
Deposited by: Memoria Investigacion
Deposited on: 25 Jul 2012 12:31
Last Modified: 01 Mar 2017 16:37
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