The development of Module In-vessel degraded severe accident Analysis Code MIDAC and the relevant research for CPR1000 during the station blackout scenario-Reference-Cited by-同舟云学术

The development of Module In-vessel degraded severe accident Analysis Code MIDAC and the relevant research for CPR1000 during the station blackout scenario

Published:2014-09 Issue: Volume:76 Page:44-54
ISSN:0149-1970
Container-title:Progress in Nuclear Energy
language:en
Short-container-title:Progress in Nuclear Energy

Author:

Wang Jun^ORCID,Tian Wenxi,Zhang Yapei,Chen Lie,Li Longze,Zhang Luteng,Zhou Yukun,Su Guanghui,Qiu Suizheng

Publisher

Elsevier BV

Subject

Waste Management and Disposal,Energy Engineering and Power Technology,Safety, Risk, Reliability and Quality,Nuclear Energy and Engineering

Reference44 articles.

1. An experimental study of natural convection in a volumetrically heated spherical pool bounded on top with a rigid wall;Asfia;Nucl. Eng. Des.,1996

2. Breakaway phenomenon of Zr-based alloys during a high-temperature oxidation;Baek;J. Nucl. Mater.,2008

3. Validation of CATHARE V2.5 thermal-hydraulic code against full-scale PERSEO tests for decay heat removal in LWRs;Bandini;Nucl. Eng. Des.,2011

4. Severe accident codes status and future development;Bolshov;Nucl. Eng. Des.,1997

5. Integrated analysis of core debris interactions and their effects on containment integrity using the contain computer code;Carroll;Nucl. Eng. Des.,1987

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