This is a laboratory working for the research and development of nuclear fuels and materials.
Research area includes Pressurized Water Reactor (PWR), Research Reactor (RR), and GEN-Ⅳ Reactors such as Fast and Molten Salt Reactors.

핵연료 및 재료의 연구 개발을 위해 노력하는 연구실입니다. 연구 분야에는 가압 경수로 (PWR), 연구로 (RR) 및 고속 및 용융 염로와 같은 GEN-Ⅳ 원자로가 포함됩니다.

Dependence of thickness, morphology, and
crystallographic properties of Mo and ZrN coatings on U-Mo substrate size

Effect of coating thickness and annealing
temperature on ZrN coating failure of U-Mo particles under heat treatment

Development of PRIME for irradiation
performance analysis of U-Mo/Al dispersion fuel

Thermophysical properties of sintered aluminum-silicon

Thermophysical properties of heattreated U-7Mo/Al

Thermal properties of U-7Mo/Al dispersion fuel

Effect of stress evolution on microstructural behavior in U-Mo/Al dispersion fuel

Pore growth in U-Mo/Al dispersion fuel

Thermal Conductivity Modeling of U-Mo/Al Dispersion Fuel

Mechanical analysis of UMo/Al dispersion fuel

Thermal conductivity of U–Mo/Al dispersion fuel: effects of particle shape and size, stereography, and heat generation

Short communication on “self-crack-healing behavior of oxide formed on a zirconium alloy cladding tube

Study of mechanism of oxidation resistance enhancement induced by preformed oxide on zirconium alloys

Performance analysis of UO2 fuel rod for low-power research reactor application

Effect of pre-oxidation temperature on the ductility of a zirconium alloy cladding under simulated accident conditions

Microstructural analysis of preformed oxides on a zirconium alloy before and after subsequent oxidation at 1000–1200 °C

Microstructural characteristics and effects of 800–1200°C preformed oxides on high-temperature steam oxidation of a zirconium alloy cladding

Enhanced high-temperature oxidation resistance of a zirconium alloy cladding by high-temperature preformed oxide on the cladding

High-temperature steam oxidation and oxide crack effects of Zr-1Nb-1Sn-0.1Fe fuel cladding

Gas releases and rod internal pressure of BN-added UO2 fuel

Fuel performance code COSMO for analysis of LWR UO2 and MOX fuel

Characteristics of GdxMyOz (M=Ti, Zr, Al) as a burnable absorber

An extension of the two-zone method for evaluating a fission gas release under an irradiation-induced resolution flux

Zircaloy-4 cladding corrosion model covering a wide range of PWR experiences

Molecular Dynamics Simulation of the Pressure-Volume-Temperature Data of Xenon for a Nuclear Fuel

Fabrication method for UO2 pellets with large grains or a single grain by sintering in air

Sintering behavior of U-80at％Zr powder compacts in a vacuum environment

Thermal conductivities of actinides (U, Pu, Np, Cm, Am) and uranium-alloys (U-Zr, U-Pu-Zr and U-Pu-TRU-Zr)

Development of integral type spent fuel pool storage rack with gadolinium and europium-containing structure materials

Understanding of electrochemical behaviors of niobium in
molten LiCl–KCl eutectic for pyrochemical decontamination process

Spatial distribution of
CeBi2 in Bi-Ce alloy facilitating density-based separation between actinides and lanthanides

Spatial distribution of HfBi in metallurgically and electrochemically
formed Bi-Hf alloy

Application of a Dynamic-nanoindentation Method to Analyze the Local Structure of an Fe-18 at.% Gd Cast Alloy

Corrosion and Wear Properties of Cold Rolled 0.087% Gd Lean Duplex Stainless Steels for Neutron Absorbing Material

Effect of gadolinium addition on the corrosion, wear, and neutron absorbing behaviors of duplex stainless steel sheet

Fabrication of Gd containing duplex stainless steel sheet for neutron absorbing structural materials

Thermal Conductivity Modeling of U-Mo/Al Dispersion Fuel 

Short communication on “self-crack-healing behavior of oxide formed on a zirconium alloy cladding tube
“

Spent Fuel Gripper using Two oval Shape Cams

Method for manufacturing nuclear fuel rod and billetof used thereof

Method for manufacturing UO2 pellet by adding Gd2O3

Method and Furnace for Sintering of U-Zr PowderCompacts in a Low-vacuum Environment

Innovative Nuclear FUEL (INFUEL)

교수님 소개

손동성

연구실 소개

연구실 키워드

알럼나이가 간 회사

연구실 지원 방법

대우 조건

연도별 논문 인용수

저널 & 학회 논문 출판 수

연구실 논문 목록

Short communication on “self-crack-healing behavior of oxide formed on a zirconium alloy cladding tube

Study of mechanism of oxidation resistance enhancement induced by preformed oxide on zirconium alloys

Performance analysis of UO2 fuel rod for low-power research reactor application

연구실 보유 특허

대학별 입시끈

우리 연구실 모집

교수님 소개

손동성

연구실 키워드

알럼나이가 간 회사

연구실 지원 방법

대우 조건

연도별 논문 인용수

저널 & 학회 논문 출판 수

연구실 논문 목록

Short communication on “self-crack-healing behavior of oxide formed on a zirconium alloy cladding tube

Study of mechanism of oxidation resistance enhancement induced by preformed oxide on zirconium alloys

Performance analysis of UO2 fuel rod for low-power research reactor application

연구실 보유 특허

최근 3년간 연구실적 높은 연구실

대학별 입시끈

우리 연구실 모집