N. Tak

发表

Thermo-fluid investigation on a double-side-cooled annular fuel for the prismatic very high temperature gas-cooled reactor

W. Lee, Yong-Hee Kim, N. Tak, 2008 .

Design of Scale-down Standard Fuel Block Test Section to Validate Core Thermo-fluid Analysis Code for Prismatic Gas-Cooled Reactor

C. S. Kim, Eungseon Kim, N. Tak, 2018 .

Study on Influence of Turbulence Model Selection on Prediction of Flow Distribution and Hot Spot Fuel Temperature in Prismatic HTGR Cores

J. Noh, Min-Hwan Kim, N. Tak, 2014 .

Thermal-fluid assessment of the design options for reactor vessel cooling in a prismatic core VHTR

Min-Hwan Kim, Hong-Sik Lim, N. Tak, 2010 .

Computational Assessment of the Reactor Vessel Cooling Options in a prismatic core VHTR

Min-Hwan Kim, W. Lee, N. Tak, 2009 .

CFD analysis of the HYPER spallation target

Yong-bum Lee, N. Tak, Jaehyuk Choi, 2008 .

Investigations of Hydraulic Behavior in Spallation Target Configurations

X. Cheng, A. Batta, N. Tak, 2006 .

Numerical and experimental investigation on labyrinth seal mechanism for bypass flow reduction in prismatic VHTR core

Kwang‐Yong Kim, Min-Hwan Kim, Sang-moon Lee, 2013 .

Numerical prediction of the fission product plate-out for a VHTR application

Hong-Sik Lim, J. Yoo, J. Chun, 2010 .

Thermo-mechanical analysis of the prismatic fuel assembly of VHTR in normal operational condition

Min-Hwan Kim, N. Tak, Ji-Ho Kang, 2012 .

Investigation on turbulent heat transfer to lead–bismuth eutectic flows in circular tubes for nuclear applications

Nam-il Tak, X. Cheng, X. Cheng, 2006 .

An uncertainty importance measure using a distance metric for the change in a cumulative distribution function

Moon-Hyun Chun, Seok-Jung Han, Nam-Il Tak, 2000, Reliab. Eng. Syst. Saf..

A combined procedure of RSM and LHS for uncertainty analyses in source term quantifications using MAAP3.0B code

Moon Hyun Chun, Seok-Jung Han, M. Chun, 1996 .

Time-dependent neutron diffusion analysis using finite element method for a block-type VHTR core design

Hong-Sik Lim, J. Cho, Seungsu Yuk, 2020 .

Water-jacket reactor cavity cooling system concept to mitigate severe accident consequence of high temperature gas-cooled reactor

Hong-Sik Lim, N. Tak, C. Jo, 2018, Nuclear Engineering and Design.

Numerical design of a 20 MW lead–bismuth spallation target for an accelerator-driven system

C. H Cho, T. Y Song, N. I Tak, 2004 .

Thermal hydraulic test with 6 MPa nitrogen gas loop for developing the Korean He cooled test blanket

Dong Won Lee, I. Bang, Sung-Deok Hong, 2010 .

Development of a Computer Code to Estimate the Fuel Rod Failure Using Primary Coolant Activities of Operating PWRs

M. Chun, Sang Kyu Lee, N. Tak, 1998 .

Numerical investigation of a heat transfer within the prismatic fuel assembly of a very high temperature reactor

Min-Hwan Kim, Won Jae Lee, Nam-il Tak, 2008 .

Improvement of GAMMA+ code for system transient and thermo-fluid safety analysis of sodium-cooled fast reactors

J. Eoh, N. Tak, Jung Yoon, 2022, Nuclear Engineering and Design.

Multi-physics steady-state analysis of OECD/NEA modular high temperature gas-cooled reactor MHTGR-350

Deokjung Lee, Matthieu Lemaire, Nam-il Tak, 2017 .

CAPP/GAMMA+ code system for coupled neutronics and thermo-fluid simulation of a prismatic VHTR core

T. Y. Han, Hyun Chul Lee, Hong-Sik Lim, 2016 .