The Spanish nuclear power generation industry proposed the development of an external Emergency Support Center as one of the measures to strengthen the nuclear safety and Emergency Preparedness and Response, as a consequence of the stress test developed after the Fukushima accident. The CAE project was carried out to define and establish a centralized service composed of intervention equipment and specialized personnel in the framework of an Emergency Support Center shared among all the Spanish nuclear power plants (NPPs). The emergency support service aims to strengthen the NPP emergency capabilities, by integrating with the Emergency Response Organization (ERO) of plants. This service successfully developed preoperational tests at each one of the Spanish NPPs in 2014. With these tests, the development of the different aspects that make up the Emergency Support Center service, at every Spanish NPP, was validated in four stages: (1) CAE mobilization in less than 24 hrs, (2) equipment deployment to its final location in the plant, (3) checking the connections with NPP’s interfaces, and (4) functionality of the equipment. The CAE proved in this way its capability to provide support to the Spanish NPPs, to strengthen their already strong characteristics in EP&R, to face these new extended damage/beyond design basis scenarios.
Skip Nav Destination
Article navigation
October 2016
Technical Briefs
CAE—The Spanish Emergency Support Center: A Centralized and Shared Emergency Support Service for Beyond Design Basis Events
Rafael J. Caro
Rafael J. Caro
Search for other works by this author on:
Rafael J. Caro
Manuscript received June 26, 2015; final manuscript received January 8, 2016; published online October 12, 2016. Assoc. Editor: Leon Cizelj.
ASME J of Nuclear Rad Sci. Oct 2016, 2(4): 044504 (5 pages)
Published Online: October 12, 2016
Article history
Received:
June 26, 2015
Revision Received:
January 8, 2016
Accepted:
January 21, 2016
Citation
Caro, R. J. (October 12, 2016). "CAE—The Spanish Emergency Support Center: A Centralized and Shared Emergency Support Service for Beyond Design Basis Events." ASME. ASME J of Nuclear Rad Sci. October 2016; 2(4): 044504. https://doi.org/10.1115/1.4032641
Download citation file:
Get Email Alerts
Cited By
Operation Optimization Framework for Advanced Reactors Using a Data Driven Digital Twin
ASME J of Nuclear Rad Sci
Investigation of Metal–H2O Systems at Elevated Temperatures: Part III: Solubility Data and New Zr Pourbaix Diagrams at 298.15 K and 373.15 K
ASME J of Nuclear Rad Sci (April 2025)
Investigation of Metal-H2O Systems at Elevated Temperatures: Part II. SnO2(s) Solubility Data and New Sn Pourbaix Diagrams at 298.15 K and 358.15 K
ASME J of Nuclear Rad Sci (April 2025)
Investigation of Metal–H2O Systems at Elevated Temperatures: Part I. Development of a Solubility Apparatus Specialized for Super-Ambient Conditions
ASME J of Nuclear Rad Sci (April 2025)
Related Articles
A Coarea Formulation for Grid-Based Evaluation of Volume Integrals
J. Comput. Inf. Sci. Eng (December,2020)
Special Issue: Highlights From ASME Computers and Information in Engineering (CIE) 2017
J. Comput. Inf. Sci. Eng (September,2018)
The Fabulous Nuclear Odyssey of Belgium
J. Pressure Vessel Technol (June,2009)
Field Testing to Validate Models Used in Explaining a Piston Problem in a Large Diesel Engine
J. Eng. Gas Turbines Power (October,1993)
Related Proceedings Papers
Related Chapters
A PSA Update to Reflect Procedural Changes (PSAM-0217)
Proceedings of the Eighth International Conference on Probabilistic Safety Assessment & Management (PSAM)
Lessons Learned: NRC Experience
Continuing and Changing Priorities of the ASME Boiler & Pressure Vessel Codes and Standards
The Fidelity of Operators' Performance Data Observed under Simulated Emergencies of Nuclear Power Plants (PSAM-0293)
Proceedings of the Eighth International Conference on Probabilistic Safety Assessment & Management (PSAM)