The pursuit of advanced Light Water Reactor (LWR) fuel candidates, such as Accident Tolerant Fuel (ATF) and high burnup fuel with modern Zircaloy cladding represents a remarkable departure from the traditional UO2 pellet and Zircaloy cladding fuel composition. This shift necessitates the establishment of new safety limits, as well as the definition of design and operational envelopes.
To address this need, our research endeavors involve conducting various cutting-edge experiments to investigate the behavior of fuel materials in extreme environments during accidents. By doing so, we aim to develop new safety criteria and refine the fuel design and operational envelope for advanced LWR fuels and reactor designs including SMRs.
Research topics under investigation
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Accident Tolerant Fuel (ATF) safety and attainable performance
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Safety limits of high burnup fuel with Low Enriched Uranium Plus (LEU+) and its application for Small Modular Reactors (SMRs)
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Integral LOCA (i-LOCA) experiments for Fuel Fragmentation Relocation and Dispersal (FFRD)
< i -LOCA facility for integral fuel accident safety study >
< Mechanical Testing of fuel materials >
< Post-accident Cr-coated ATF cladding characterization >
Collaborators
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IAEA
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Karlsruhe Institute of Technology (Germany)
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Oak Ridge National Laboratory (USA)
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Massachusetts Institute of Technology (USA)
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Ceramic Tubular Product (USA)
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KEPCO Nuclear Fuel (KNF)
Many performance metrics of Light Water Reactors (LWRs), such as discharge burnup and cycle length, are constrained by the allowable uranium enrichment. The introduction of LEU+ fuel (enrichment 5-10%) in the next generation of LWRs poses several technical challenges to address. However, it also promises to unlock the potential of future LWRs.
An important aspect to note is that the benefit of increased uranium enrichment can be fully capitalized in Small Modular Reactors (SMRs), where the degradation of nuclear fuel materials is more manageable. Our research focuses on developing an innovative LEU+ fueled SMR core and fuel design using high-performance fuel materials. This effort is poised to shape the future of SMRs with critical technological breakthroughs.
Research topics under investigation
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SMR design with LEU+ fuels
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Safety assessment of ultra-high burnup LEU+ fuels in SMR
Collaborators
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Innovative Small Modular Reactor Development Agency
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Korea Atomic Energy Research Institute (KAERI)
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KAIST
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Hanyang University
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Keimyung University
Innovative Small Modular Reactor
Development Agency
Analyzing nuclear fuel behavior is a fundamental aspect of nuclear reactor design and analysis. We are developing a nuclear fuel simulation code that enables high-fidelity modeling of next-generation nuclear fuel designs and their behaviors, encompassing a wide range of applications including Small Modular Reactor (SMR) fuels, multi-layered Accident Tolerant Fuel (ATF) cladding, high burnup Light Water Reactor (LWR) fuels, spent fuels during dry storage, and TRISO fuels for advanced reactors.
This code represents the pinnacle of Korean nuclear fuel simulation technology, with broad implications for fuel design and analysis. It integrates separate effect models derived from experimental investigations of individual fuel behavior investigated in our group.
Research topics under investigation
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Development of advanced LWR fuel code GIFT
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Development of particle-based high temperature fuel (i.e., TRISO) code TRIPLE
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Multi-physics (Fuel - Reactor Physics - Thermal Hydraulics) code coupling
: SNU-developed LWR fuel analysis code
< Advanced fuel mechanical model and code implementation >
< Multi-physics: Fuel - Reactor Physics - Thermal Hydraulics coupled analysis >
< GIFT - COBRA SFS coupled analysis for dry storage of spent nuclear fuels >
Reactor Physics
(Vanguard)
Fuel
(GIFT)
Thermal Hydraulics
(Cupid)
: Advanced fuel cladding oxidation model
< Code development >
< Experimental validation >
: SNU-developed particle-based fuel (TRISO) fuel code
< Code development >
< TRIPLE Code >
: SNU-developed image analysis code
Collaborators
-
Korea Atomic Energy Research Institute (KAERI)
Interim storage of spent nuclear fuel is a pressing concern for the sustainability of nuclear energy in Korea. In alignment with the national interest in pursuing interim dry storage solutions, our group is investigating the safety of spent nuclear fuel during extended dry storage, with a particular focus on hydride-induced cladding embrittlement, which is considered a key safety-limiting factor.
We are employing various cutting-edge material characterization techniques, including Electron Backscatter Diffraction (EBSD), Scanning Electron Microscopy (SEM), Transmission Electron Microscopy (TEM), and in-house developed image-analysis code to advance the understanding of hydrogen-induced cladding embrittlement in Zircaloy. These efforts aim to enhance the field's knowledge and enable informed decisions regarding the safety of spent nuclear fuel during extended dry storage.
Research topics under investigation
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Advanced microstructural characterization of hydride in Zircaloy with EBSD and TEM
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Hydride embrittlement of high burnup Zircaloy and advanced Zircaloy such as HANA-6
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Multi-physics modeling of spent fuel behavior during dry storage
< Macroscopic Characterization >
Characterization of hydride morphologies using PROPHET and mechanical strength
< Microstructural Characterization >
EBSD characterization of hydrides
in CWSR Zircaloy cladding
Schematic diagrams of circumferential and radial hydrides in tubular gemetry
TEM characterization of hydrides
in CWSR Zircaloy cladding
< Modeling and Simulation >
SNU-developed hydride reorientation modeling code
SNU-developed
hydrogen diffusion modeling code
Collaborators
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Korean Institute of Materials Science (KIMS)
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Korea Atomic Energy Research Institute (KAERI)
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KEPCO Nuclear Fuel (KNF)
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Centrum výzkumu Řež (Chezch Republic)
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Karlsruhe Institute of Technology (Germany)
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Gesellschaft für Anlagen- und Reaktorsicherheit (Germany)
The qualification of nuclear materials against irradiation damage represents a critical factor that limits the realization of most advanced reactors. Simultaneously, it significantly impedes the innovation of nuclear reactor technology. To address this seminal challenge facing global advanced reactor development, we have recently initiated a research program aimed at mimicking neutron irradiation damage using ion beams.
In collaboration with MIT, we are developing an advanced ion-beam radiation damage quantification method using Transient Grating Spectroscopy (TGS). This innovative approach will enable precise and efficient assessment of material response to irradiation, facilitating the development of next-generation nuclear materials with enhanced tolerance against radiation damage.
Research topics under investigation
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Ion beam irradiation damage on nuclear materials
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Mimicking neutron irradiation damage with ion beam
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Utilizing Transient Grating Spectroscopy (TGS) for measuring swelling and property changes induced by ion beam radiation
*In the courtesy of Prof. Short group's (MIT) TGS research
Collaborators
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Massachusetts Institute of Technology (USA)
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Korea Atomic Energy Research Institute (KAERI)
Improving the construction process to enhance cost competitiveness is a global challenge faced by large-scale nuclear power plants. To gain deep insights into the underlying issues, we are conducting a comprehensive analysis of the construction processes employed in today's large nuclear power plants, such as APR 1400 and AP 1000. By comparing these processes of different power plants, we aim to identify essential factors for enhancing construction efficiency, thereby revitalizing the global nuclear industry.
Additionally, our research extends to the construction processes of Small Modular Reactors (SMRs), focusing on expected construction durations and the requirements for construction-friendly SMR designs. This investigation directly addresses the key uncertainties surrounding SMR deployment and proposes new insights and directions for the future development of nuclear power plants.
Research topics under investigation
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APR1400 and AP1000 construction process comparison
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SMR construction process analysis and investigation of construction-friendly SMR design
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Implementation of Building Information Management (BIM) in Nuclear Industry
APR1400 and AP1000
construction process comparison
in terms of....
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Construction schedule
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Concrete installation productivity (rate)
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Modular construction
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Construction method (e.g., Open Top construction)
<Big-6 Modules of AP1000 reactor>
< BIM Model >
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Construction schedule estimation
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Cash flow estimation
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Cost / Schedule optimization
SMR construction process analysis and
investigation of construction-friendly SMR design
On-going analysis
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SMART-100 (ROK)
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NuScale
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BWRX-300 (GE)
*Simulation video created by Prof. Youho LEE's group
Collaborators
-
Korea Hydro & Nuclear Power (KHNP)
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Korea faces a unique situation where the supply of enriched nuclear fuel is entirely dependent on overseas sources. This dependence naturally escalates to a national security concern. It is imperative for the front-end nuclear fuel cycle policy to consider a comprehensive perspective on the national fuel and reactor development outlook and prospects.
We are actively engaged in reevaluating the fuel supply landscape of Korea with the aim of revitalizing national security associated with the use of nuclear power. By addressing this critical issue, we seek to ensure a sustainable and secure energy future for Korea.
Research topics under investigation
-
Analyzing the impact of the global uranium supply market on the national fuel supply plan
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Korea-U.S. Cooperation in the front-end nuclear fuel cycle
농축 우라늄 수급
불안에 따른 가격 폭등
농축도 5% 초과
상업용 핵연료 수요
미국의 농축 우라늄
공급능력 부족,
동맹국 공조 필요성 대두
"우라늄 수급시장
지형도 변화"
< Global Uranium price trend>
Collaborators
-
Korea Hydro & Nuclear Power (KHNP)
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