Enhancement of COBRA-EN capability for VVER reactors calculations

Author:

Aghaie M.,Zolfaghari A.,Minuchehr M.,Norouzi A.

Publisher

Elsevier BV

Subject

Nuclear Energy and Engineering

Reference7 articles.

1. Basile, D. et al., 1999. COBRA-EN: An Upgraded Version of the COBRA-3C/MIT Code for Thermal–Hydraulic Transient Analysis of Light Water Reactor Fuel Assemblies and Cores. Report 1010/1, ENELCRTN, Milano.

2. Downar, T., Xu, Y., Kozlowski, T., 2006. PARCS v2.7 U.S. NRC Core Neutronics Simulator USER MANUAL August, 2006. School of Nuclear Engineering, Purdue University, W. Lafayette, Indiana.

3. A problem in the COBRA-EN code related to the void fraction calculation;Braz Filho;Annals of Nuclear Energy,2005

4. Rowe, D.S., 1973. COBRA-IIIC: A Digital Computer Program for Steady State and Transient Thermal–Hydraulic Analysis of rod Bundle Nuclear Fuel Elements. BNWL-1695, Battelle-Northwest, Richland.

5. Modified COBRA-EN code to investigate thermal–hydraulic analysis of the Iranian VVER-1000 core;Safaei Arshi;Progress in Nuclear Energy,2010

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