Browse "Dept. of Nuclear and Quantum Engineering(원자력및양자공학과)" by Title

Showing results 2321 to 2340 of 9051

Crack-tip plasticity and intrinsic toughening in nano-sized brittle amorphous carbon

Shin, Dahye; Jang, Dongchanresearcher, INTERNATIONAL JOURNAL OF PLASTICITY, v.127, 2020-04

Cracking effect on gamma-ray shielding performance in concrete structure

Lee, Chang-Min; Lee, Yoon Hee; Lee, Kun Jairesearcher, PROGRESS IN NUCLEAR ENERGY, v.49, no.4, pp.303 - 312, 2007

CREAM-based communication error analysis method (CEAM) for nuclear power plant operators communication

Lee, Seung Min; Ha, Jun Su; Seong, Poong-Hyunresearcher, JOURNAL OF LOSS PREVENTION IN THE PROCESS INDUSTRIES, v.24, no.1, pp.90 - 97, 2011-01

Creating a Synthetic Mammogram From a Digital Breast Tomosynthesis Image

Hong, Joon Pyo; Cho, Seung Ryongresearcher; Lee, Tae Won; Kim, Hyung Seok; Choi, Young Wook, AAPM : 57th Annual Meeting & Exhibition, The American Association of Physicists in Medicine, 2015-07-14

Creep and Oxidation Behavior of Nickel-base Superalloys in High Temperature Helium Environments with Various Oxygen Concentrations

구자현; 장창희researcher; 김대종, JOURNAL OF KOREA PRESSURE VESSEL AND PIPING, 2011-02

Creep behavior of Alloy 617 in high temperature air and helium environments-effect of oxidation damage

Jang, Changheuiresearcher; Kim, Sung Hwan; Sah, Injin; Kim, Daejong, JOURNAL OF MECHANICAL SCIENCE AND TECHNOLOGY, v.30, no.10, pp.4433 - 4438, 2016-10

Creep behaviors of alloy 617 at temperatures between 800 and 1000WOC

Kim, D.; Jang, Changheuiresearcher; Ryu, W.S., International Conference on Advances in Nuclear Power Plants, ICAPP 2008, pp.2158 - 2162, 2008-06-08

Creep Behaviors of Alloy 617 in High Temperature Steam Environments

Jang, Changheuiresearcher; Donghoon Kim; Injin Sah; Jahyun Koo, NFSM 2012, American Nuclear Society, 2012-06-26

Creep Behaviors of Nickel-Base Superalloys Alloy 617 and Haynes 230 at High Temperature

Jang, Changheuiresearcher, 2009 International Congress on Advances in Nuclear Power Plants, 2009-05-12

Creep behaviour of advanced nuclear materials measured by the SPS system

Faris, Sweidan; Ryu, Ho Jinresearcher, The Nuclear Materials Conference, NUMAT 2018, 2018-10-15

Creep resistance of Alloy 617 in high temperature helium environments

Jang, Changheuiresearcher; Ho Jung Lee; Injin Sah; Jahyun Koo; Dae Jong Kim, Asia Pacific Conference on Fracture Strength-Machanical and materials 2012, Korean Society of Mechanical Engineering(KSME), 2012-05-14

Creep-environment interactions of Alloy 617 at elevated temperature

Kim, D.; Jang, Changheuiresearcher; Ryu, W.S., ASME 2008 Pressure Vessels and Piping Conference, PVP2008, pp.1215 - 1220, 2008-07-27

Criteria for near-surface disposal of wastes from peacer reactor

Sung I.K.; Lee, Kun Jairesearcher, American Nuclear Society - International Congress on Advances in Nuclear Power Plants 2005, ICAPP'05, v.4, pp.2309 - 2317, 2005-05-15

Critical Assessment of Transuranic Element Storage Facility with Wet Conditions: Monte Carlo Study

고길영; 김진환; 이민주; 김우섭; 안승규; 박세환; 조규성researcher, 한국원자력학회, 한국원자력학회, 2018-05-17

Critical conditions for pit initiation and growth of austenitic stainless steels

Al Ameri, M.; Yi, Y.; Cho, P.; Al Saadi, S.; Jang, Chang-Heuiresearcher; Beeley, P., CORROSION SCIENCE, v.92, pp.209 - 216, 2015-03

Critical Flow Rates of Subcooled Water Through Short Pipes with Small Diameters

Chun, Moon Hyun, Proc. of the 8th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-8), pp.759 - 766, 1997

Critical heat flux and flow pattern for water flow in annular geometry

Park, JW; Baek, WP; Chang, Soon-Heungresearcher, NUCLEAR ENGINEERING AND DESIGN, v.172, no.1-2, pp.137 - 155, 1997

Critical heat flux and onset of nucleate boiling tests in a finned rod bundle

Chae, HT; Park, JH; Kim, H; Chang, Soon-Heungresearcher, NUCLEAR TECHNOLOGY, v.148, no.3, pp.287 - 293, 2004-12

Critical heat flux characteristic of magnetite-water nanofluid in pool boiling

Lee, Jong Hyuk; Lee, Taeseung; Jeong, Yong Hoonresearcher, The 8th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS8), Thermal Hydraulics Divisions of AESJ, KNS, 2012-12

Critical Heat Flux during Flow Boiling Experiment with Surfactant Solutions

Sarwar, Mohammad Sohail; Jeong, Yong Hoonresearcher; Chang, Soon-Heungresearcher, 한국원자력학회 추계학술발표대회, 한국원자력학회, 2007-05-11



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