Reshaping key conservatism in nuclear fuel design and safety 

Nuclear Fuel Material Behavior in Extreme Condition


Nuclear fuel engineering is essential for the fundamental advancement of nuclear reactor technology. At the same time, it is a factor that remarkably slows down the innovation of nuclear power. The complexities of fuel, together with exceedingly high demands placed on performance predictability, calls for a series of exhaustive tests that can easily take a decade to achieve any improvements. In addition, a remarkable departure from the traditional UO2 pellet Zircaloy cladding fuel composition, sought by advanced reactors and Light Water Reactor (LWR) accident tolerant fuel candidates, further challenges timely performance evaluations and design optimization. Our multi-disciplinary expertise systematically integrates material research and nuclear system studies. This is what I think the field needs today in order to expedite development of unexplored nuclear fuel and reactor designs and safety regulations to unleash them.

Research subjects: High temperature fuel safety experiments, 
Accident Tolerant Fuel, High burnup fuel safety, Nuclear fuel safety criteria

Developing a Korean PWR Fuel Analysis Code

Nuclear Fuel Code Development


We are developing a nuclear fuel simulation code. The code aims at simulating 2D axial symmetric, real scale PWR nuclear fuel rod (4m) at appropriate computational costs. The developed model can handle the plastic deformation of cladding including creep and multilayered cladding, which is the structure of several Accident Tolerant Fuel candidates. We are expecting to propel research in nuclear fuel code development to advance the understanding of integral fuel behavior thereby improving fuel design and safety.

Research subjects: Fuel cladding and pellet model implementation, 
Accident Tolerant Fuel simulation, Accident fuel behavior simulation, Spent fuel simulation

Advancing the understanding of hydrogen-induced embrittlement to solve the near-term spent fuel issue

Spent Fuel Behavior and Safety


Managing pre-disposal spent nuclear fuel is a major technical issue currently faced by the nuclear industry. Pre-disposal spent fuel management is divided into wet and dry storage stages. We are currently investigating structural safety of spent fuels in the pre-disposal stages and support advanced safety spent fuel management regulations. Our research team is conducting both experimental and computational studies to understand key spent fuel behavior during dry-storage. By doing so, we are expecting to contribute to the dry-storage design, management protocols, and regulatory guidelines.


Research subjects: Hydrogen-induced cladding embrittlement of spent nuclear fuel, Hydride reoreintation in Zirconium-based alloys

A new insight into material behavior in extreme environments by removing barriers between material mechanics and thermal hydraulics

Material-Fluid Interface Engineering


During an accident in a nuclear reactor, a strong energetic interaction between reactor coolant and structural materials occurs. Such an aggressive energetic interaction is the root of most transient structural problems by which reactor design and safety systems are largely influenced. By limiting our perspective to individual disciplines, we will never be able to comprehensively understand these safety issues, unnecessarily relying on empiricism that slows down the introduction of new fuel designs with years of exhaustive experiments. Our research team's  multidisciplinary expertise in both material mechanics and thermal hydraulics has afforded a new insight that has led to scientific discoveries on thermal shock of materials.

In the ceramic science community, improving thermal shock tolerance has long been the goal of much research. We produced nanostructure coating to enhance the hydrophobicity of material surface and reduce stress from thermal shock caused by liquid droplet impingements, and as a result, produced material that is resistant to thermal shock. This work therefore opens up new avenues for resistance to thermal shocks in a wide range of applications.

Research subjects: Superior thermal shock tolerant material, 
Critical Heat Flux of advanced cladding surfaces

Rethinking the nuclear power density 

Innovative Next-generation LWR 


The high power operation has long been an unquestioned norm of nuclear reactor design for the economic reason. However, excess power is a key factor that negatively affects the safety of the power plant. The current reactor core power density necessitates complex Emergency Core Cooling System (ECCS).  The heavy reliance on ECCS for accident coping strategy made the reactor inherently vulnerable to accidents occurred by a coincidental alignment of failures in designed ECCS. We are currently re-examining the LWR power density and developing a groundbreaking next-generation water reactor with low power density

Research subjects: Reactor design of SNUSCALE

Solving the corrosion issue of LBE to revisit the potential of LFR

Material Corrosion of Lead-cooled Fast Reactor (LFR)


Our laboratory holds the world’s largest natural convection lead-bismuth experimental loop (PILLAR) thanks to the legacy of Prof. IS. Hwang.  We are conducting the lead-bismuth nuclear reactor flow accelerated corrosion experiment and collecting experimental data of LBE natural circulation for code validation. At the same time, we are conducting a small scale pool-type lead experiment equipment and simultaneously test material corrosion in a well-controlled environment.

Despite the breakthrough in structural material designs achieved in recent years and merits of Lead-cooled Fast Reactor (LFR), the potential of LFR has been inappropriately evaluated due to concerns surrounding the corrosion issue. By testing advanced alloys and advancing an understanding of flow accelerated corrosion phenomenon, we look forward to contributing to revisiting the potential of LFRs. In doing so, the LFR research of Seoul National University will continuously be bequeathed to future generations and most importantly, solving the material-coolant compatibility issue of a next-generation nuclear reactor.

Research subjects: Natural convection study of LBE using demo-scale facility PILLAR, flow-accelerated corrosion of structural materials in LBE and Pb.