Simulation and transient analyses of a complete passive heat removal system in a downward cooling pool-type material testing reactor against a complete station blackout and long-term natural convection mode using the RELAP5/3.2 code-Reference-Cited by-同舟云学术

Simulation and transient analyses of a complete passive heat removal system in a downward cooling pool-type material testing reactor against a complete station blackout and long-term natural convection mode using the RELAP5/3.2 code

Published:2017-08 Issue:5 Volume:49 Page:953-967
ISSN:1738-5733
Container-title:Nuclear Engineering and Technology
language:en
Short-container-title:Nuclear Engineering and Technology

Author:

Hedayat Afshin

Publisher

Elsevier BV

Subject

Nuclear Energy and Engineering

Reference38 articles.

1. Safety of Research Reactors. IAEA Safety Standards, Safety Requirements, No. NS-R-4;IAEA,2005

2. Fundamental Safety Principles. Safety Standard No. SF-1;IAEA,2006

3. Safety Analysis for Research Reactor. Safety Reports Series No. 55;IAEA,2008

4. Operational Limits and Conditions and Operating Procedures for Research Reactors. Safety Standard No. NS-G-4.4;IAEA,2008

5. Workshop on Strengthening the Role of the Regulatory Authority in Light of the Fukushima Accident. 5–8 Sep. 2014;INRA and IAEA,2014

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