Modified COBRA-EN code to investigate thermal-hydraulic analysis of the Iranian VVER-1000 core-Reference-Cited by-同舟云学术

Modified COBRA-EN code to investigate thermal-hydraulic analysis of the Iranian VVER-1000 core

Published:2010-08 Issue:6 Volume:52 Page:589-595
ISSN:0149-1970
Container-title:Progress in Nuclear Energy
language:en
Short-container-title:Progress in Nuclear Energy

Author:

Safaei Arshi S.,Mirvakili S.M.,Faghihi F.

Publisher

Elsevier BV

Subject

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

Reference11 articles.

1. COBRA-EN: an Upgraded Version of the COBRA-3C/MIT Code for Thermal Hydraulic Transient Analysis of Light Water Reactor Fuel Assemblies and Cores;Basile,1999

2. Nuclear Heat Transport;El-Wakil,1993

3. Reactivity coefficients simulation of the Iranian VVER-1000 nuclear reactor using WIMS and CITATION codes;Faghihi;Prog. Nucl. Energ.,2007

4. Burn up calculations for the Iranian miniature reactor: a reliable and safe research reactor;Faghihi;Nucl. Eng. Des.,2009

5. Fowler, T.B., Vondy, D.R., Cunningham, G.W., 1971. Nuclear Reactor Core Analysis Code: CITATION.ORNL–TM–2496, Rev. 2.

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