Research

Ongoing Research

Development and Verification of Control Rod Depletion Module for Innovative SMR Design. Korea Nuclear Fuel Co., Ltd. (2023.11.06~2025.12.31)
Development of multi-physics integrated platform for 3D reactor core validation and safety/performance analysis of i-SMR. Innovative Small Modular Reactor (2024.01.01~2028.12.31)
Development and verification of the group constants generation method for prompt responding in-core SPND. Korea Nuclear Fuel Co., Ltd. (2024.02.23~2027.12.31)
Development of Core Design with Flexible Operation and Core Protection and Monitoring System for i-SMR. Innovative Small Modular Reactor (2024.04.01~2028.12.31)
Development of Base-Technology for LEU+/ATF Fuel and Core Application. National Research Foundation of Korea (2024.04.01~2026.12.31)
Development and Coupling of Reactor Core Analysis Code, RAST-K, to NPP Simulator. Korea Hydro & Nuclear Power Co., Ltd. (2024.06.13~2026.09.30)
Development of a core-integrated code (FAMILY) for safety regulation. Korea Institute of Nuclear Safety (2025.01.01~2025.12.31)
Development of Optimal Flexible Operation Strategy of i-SMR smart Net-zero City. Pohang University of Science and Technology Industry-Academic Cooperation Group (2025.02.25~2028.02.24)
Next-Generation Sodium-Cooled Fast Reactor Specialist Training Center. National Research Foundation of Korea (2025.05.01~2029.12.31)
Development of Monte Carlo High-Reliability Core Analysis and Multiphysics Coupling Technology to Reduce Design Uncertainty for Small Modular Reactors. National Research Foundation of Korea (2025.04.01~2029.12.31)

Finished

Development of World-Leading Full-Core Transport Numerical Reactor for Nuclear Design of Commercial Nuclear Power Plant (2019.04 – 2024.12)
Localization of In-core Nuclear Instrument Signal Measurement System with High Performance and Prompt Response Instrumentation based Technology Development (2020.05 – 2023.04)
Development of Key Technologies for Conceptual Design of Non-refueling Full-life Micro Reactor for Marine Applications (2019.04 – 2022.12)
IAEA CRP on Neutronics Benchmark of CEFR Start-Up Tests (2018.04 – 2022.12)
Development of operation supporting technology based on artificial intelligence for nuclear power plant start up and shutdown operation (2017.12 – 2022.11)
Development of high-performance 3D multiphysics computer code for reactor transient analysis (2019.03 – 2022.02)
Development of accident tolerance fuel utilization technology of integrated analysis platform (2020.01 – 2021.12)
Development of Lead-cooled Fast Reactor Technology (2017.04 – 2021.12)
Development of multi-physical safety analysis data compression code using PCA (2019.09 – 2021.08)
Development of AI Based Reactor Core Diagnostics System (2019.06 – 2021.05)
Evaluation of the rod-base core fuel characteristics for the high precision FFRD model (2020.08 – 2021.04)
Development of Multicycle Optimal Core Design Technology based on Simulation Annealing Method (2019.03 – 2021.02)
Development of High-Fidelity Multi-physics Computer Code Using Monte Carlo Method for PWR Analysis (2017.06 – 2020.05)
Development of Advanced Super-Critical Water-cooled Reactor and the Dedicated High Performance Computer System (2015.01 – 2020.02)
Establishment of a virtual-universial simulation platform for NPP systems (2018.02 – 2020.01)
Optimization Evaluation System Development for High Burnup Spent Nuclear Fuel Transportation/Storage Casks (2018.06 – 2019.12)
Safety evaluation system development for high burnup spent nuclear fuel Transportation/Storage casks (2018.07 – 2019.12)
크러드 침적 저감 피복관 및 미소셀 소결체 핵연료의 노심 핵적 특성 분석 (2018.09 – 2019.12)
Development of High Performance Three-Dimensional Neutron Transport Analysis Code (2016.11 – 2019.10)
Development and Coupling of Dynamic Reactor Core Nodal Computational Code and Nuclear Fuel Performance Analysis Code (2017.04 – 2019.09)
Assesment of Neutron Transport Methodology for Fast Reactor Application (2018.05 – 2018.12)
Development of Monitoring Equipment for Boron Concentration in Reactor Coolant (2015.09 – 2018.08)
Development of Neutron Transport and Kinetics Codes for Boron Free SMR (2014.03 – 2018.04)
Generation and Verification of Multi-group Cross Section Library of Lattice Physics Code for High Temperature (2017.02 – 2018.01)
Development and Analysis of Experimental Model for the Verification of Prototype Sodium-cooled Fast Reactor (2017.05 – 2017.12)
Depletion Benchmark Development for High-Fidelity Reactor Analysis (2014.12 – 2017.11)
Development of High Performance Monte Carlo Computer System for Light Water Reactor Whole Core Analysis (2014.06 – 2017.05)
Development of Advanced Long Life Small Modular Fast Reactor (2013.12 – 2016.11)
Development of Big Data Based Demonstration Platform to Improve the Safety of NPP (2014.12 – 2016.11)
Development of Resonance Treatment Method for VHTR Fuel (2015.03 – 2016.02)
Accuracy Enhancement of Neutron Transport Analysis by Improving Energy Resonance Self-Shielding Methods (2012.06 – 2015.05)
IAEA ADSR benchmark : kyoto university critical assembly – type A (2014.06 – 2015.05)
Accuracy Enhancement of Neutron Transport Alaysis by Improving Energy Resonance Self-Shielding Methods (2013.06 – 2015.05)
Establishment of Technology Supporting Foundations for Global Capability Enhancement of BOP in Nuclear Power Plants (2012.11 – 2015.04)
Future Nuclear Human Resource Development by Attending Advanced Nuclear Summer Schools (2013.06 – 2014.05)
Smart-grid based nuclear load-following operation technology (2011.06 – 2014.05)
Development of Advanced Reactor Core Analysis Code Employing Fusion Technique of Probabilistic Method and Deterministic Method (2011.05 – 2014.04)
Development of Ultra-long-life Core Fast Reactor (UCFR) Concept and Key Technology (2012.04 – 2014.03)
Basic research for Th cycle MSR development (2013.03 – 2014.02)
Summer School MeV, MIT NPS (2012.07 – 2013.05)
Premium Power Plant Design Requirements and Safety Research (2012.04 – 2013.05)
Feasibility Study of Daily Load Follow Operation for OPR1000 (2012.06 – 2013.02)
Summer School MeV, FJOH (2011.07 – 2012.07)

Nuclear Reactor Physics

UNIST CORE (COmputational Reactor physics and Experiment laboratory) is an academic laboratory dedicated to the research in nuclear reactor physics. The laboratory actively develops theoretical methods and computer codes to tackle neutron transport and diffusion theory in all its aspects: collision with medium, slowing-down, scattering, absorption, nuclear fission, chain reaction, secondary particle production and nuclide transmutation. The laboratory manages the knowledge required for reactor core design, analysis and operation and conducts studies in the fields of criticality, reactivity feedback and reactivity control, reactor kinetics, nuclear fuel depletion, perturbation theory, deep-penetration shielding, steady-state and transient simulations, reactor design and safety analysis, and many more.

유니스트 원자로물리연구실은 원자로물리에 대하여 연구하고 있습니다. 원자로물리란 중성자 수송이론 및 확산이론을 기초로 원자로 내에서 중성자의 거동 즉, 매질과 충돌, 산란, 흡수, 핵분열 및 연쇄 반응 등을 연구하고, 중성자의 시공간 및 에너지 분포를 예측 및 제어하는 방법을 연구하는 공학의 한 분야입니다. 임계, 반응도 제어, 중성자 감속, 중성자 확산이론, 중성자 수송이론, 동특성, 반응도 궤환 등 원자로 설계, 해석 및 운전에 필요한 지식을 다룹니다. [네이버 지식백과] 원자로물리학 [Nuclear reactor physics] (학문명백과 : 공학, 형설출판사)

Interests

Code Development
Methodology Development
Nuclear Reactor Design
Machine Learning and Artificial Intelligence
Multi-Physics with Accuracy and Uncertainty (BEPU)

Research Projects

Advanced Reactor Analysis Computer Codes Developments
Reactor analysis code

Monte Carlo Code Development – MCS Neutron Transport Code Development – STREAM MC-MoC Hybrid Methodology Development Resonance Treatment Method of Characteristics (MoC) Solver CMFD implementation, Linear source approximation, Memory optimization Future study: Depletion, Adjoint flux (perturbation), 2-D/1-D coupling, Parallelization of 2-D/1-D, MoC at transient state
BEAVRS by MCS
MCS - Monte Carlo Code

MCS: Large scale reactor analysis with accelerated Monte Carlo simulation BEAVRS benchmark analysis results
Methodology Development
Methodology development of reactor physics

Monte Carlo Methods Method of Characteristics Unified Nodal Method Acceleration Techniques
Nuclear Reactor Design
SM-SFR, PWR, and MSR

Design study of Ultra-long Cycle Fast Reactor UCFR 1000MWe and short-term deployable 100MWe design Strategies: Breed-and burn, power flattening, inherent safety Fuel study: LEU zoning, blanket breeding, PWR spent fuel loading, uranium-thorium mixed fuel performance, metallic fuel analysis Safety analysis: Inherent safety, temperature coefficient, sodium void worth, control rod worth, fast neutron fluence Design of SM-SFR Small modular fast reactor with breed-and burn strategy and liquid metal coolant Feasible design criteria study for nuclear power plant type and components Take advantage of the merit of SMR and SFR simultaneously Benchmark-Modeling for Calculating Neutronic Parameters To Solve Benchmark Problems with MC code / MoC and Nodal code KUCA Kyoto university critical assembly Molten Salt Reactor Design Molten salt reactor (MSR) design and code development GEN-III+ PWR Burnable absorper design for low-boron operation Reflector design
Multi-Physics Multi-Scale Simulations with Accuracy and Uncertainty (BEPU)
Multy-Physics Coupling, Uncertainty Quantification, Accuracy Improvements

Worldwide attention from nuclear researchers are poured into the calculation of more realistic results in terms of reactor core safety margins against critical core conditions. The analysis of non-quantified uncertainties on account of multi-physics phenomena involves the coupled modeling of neutron kinetics, coolant thermal-hydraulics and nuclear fuel performance using the numerical integration methods with built-in precision and accuracy control.
Machine Learning and Artificial Intelligence
Cross-section Generation, Core Diagnostics System, Surrogate Model for Core Parameters

Application of CNN, ANN in predicting the core parameters, cross-section generation for nodal code, calculation acceleration for transport code, data generation and management for training and testing of developed networks. AI-based diagnostic for reactor operation and Simulated Annealing based algorithm for loading pattern optimization are also being conducted.

Laboratory Resource

LINUX Cluster
PARAO, SPHINX, and PYRAMID

PARAO

SPHINX

PYRAMID
Retained Codes & Nuclear Data Libraries

Reactor Analysis Computer

Nuclear Data Library