Practical numerical reactor employing direct whole core neutron transport and subchannel thermal/hydraulic solvers-Reference-Cited by-同舟云学术

Practical numerical reactor employing direct whole core neutron transport and subchannel thermal/hydraulic solvers

Published:2013-12 Issue: Volume:62 Page:357-374
ISSN:0306-4549
Container-title:Annals of Nuclear Energy
language:en
Short-container-title:Annals of Nuclear Energy

Author:

Jung Yeon Sang,Shim Cheon Bo,Lim Chang Hyun,Joo Han Gyu

Funder

National Research Foundation of Korea (NRF)

Publisher

Elsevier BV

Subject

Nuclear Energy and Engineering

Reference25 articles.

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4. Hwang, D.H. et al., 2008. Validation of a Subchannel Analysis Code MATRA Version 1.0. Korea Atomic Energy Research Institute, Report, KAERI/TR-3639/2008.

5. Accuracy and uncertainty analysis of PSBT benchmark exercises using a subchannel code MATRA;Hwang;Sci. Technol. Nucl. Inst,2012

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