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Lee, Deokjung (이덕중)

Department
Department of Nuclear Engineering(원자력공학과)
Website
https://sites.google.com/a/unistcore.com/main/
Lab
Computational Reactor physics & Experiment Lab. (원자로물리 연구실)
Research Keywords
중성자 수송이론, 중성자 확산이론, 원자로 물리학, 코드 개발, 고성능 컴퓨팅, 원자로 노심 시뮬레이터, 첨단 원자로 설계, 방법 개발, reactor physics, nuclear reactor physics, neutron transport theory, neutron diffusion theory, code development, method development, advanced reactor design, high-performance computing, reactor core simulator
Research Interests
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.
This table browses all dspace content
Issue DateTitleAuthor(s)TypeViewAltmetrics
2021-08Uncertainties of PWR spent nuclear fuel isotope inventory for back-end cycle analysis with STREAM/RAST-KJang, Jaerim; Kong, Chidong; Ebiwonjumi, Bamidele, et alARTICLE24 Uncertainties of PWR spent nuclear fuel isotope inventory for back-end cycle analysis with STREAM/RAST-K
2021-07Bayesian method and polynomial chaos expansion based inverse uncertainty quantification of spent fuel using decay heat measurementsEbiwonjumi, Bamidele; Lee, DeokjungARTICLE10 Bayesian method and polynomial chaos expansion based inverse uncertainty quantification of spent fuel using decay heat measurements
2021-07Verification and validation of isotope inventory prediction for back-end cycle management using two-step methodJang, Jaerim; Ebiwonjumi, Bamidele; Kim, Wonkyeong, et alARTICLE28 Verification and validation of isotope inventory prediction for back-end cycle management using two-step method
2021-04시뮬레이션 데이터 생성을 통한 기계학습 기반 원자로 노심 이상 탐지 방법론오용경; 김한주; 이덕중, et alARTICLE34 시뮬레이션 데이터 생성을 통한 기계학습 기반 원자로 노심 이상 탐지 방법론
2021-03MicroURANUS: Core design for long-cycle lead-bismuth-cooled fast reactor for marine applicationsNguyen, Tung Dong Cao; Khandaq, Muhammad Farid; Jung, Eun, et alARTICLE55 MicroURANUS: Core design for long-cycle lead-bismuth-cooled fast reactor for marine applications
2021-03Uncertainty quanti fi cation of PWR spent fuel due to nuclear data and modeling parametersEbiwonjumi, Bamidele; Kong, Chidong; Zhang, Peng, et alARTICLE51 Uncertainty quanti fi cation of PWR spent fuel due to nuclear data and modeling parameters
2021-02Criticality benchmark of Monte Carlo code MCS for light water reactor fuel in transportation and storage packagesLee, Hochul; Jang, Jaerim; Jang, Junkyung, et alARTICLE97 Criticality benchmark of Monte Carlo code MCS for light water reactor fuel in transportation and storage packages
2021-01Analysis of VVER-1000 mock-up criticality experiments with nuclear data library ENDF/B-VIII.0 and Monte Carlo code MCSSetiawan, Fathurrahman; Lemaire, Matthieu; Lee, DeokjungARTICLE106 Analysis of VVER-1000 mock-up criticality experiments with nuclear data library ENDF/B-VIII.0 and Monte Carlo code MCS
2021-01Validation of spent nuclear fuel decay heat calculation by a two-step methodJang, Jaerim; Ebiwonjumi, Bamidele; Kim, Wonkyeong, et alARTICLE165 Validation of spent nuclear fuel decay heat calculation by a two-step method
2020-12Neutronic simulation of China experimental fast reactor start-up tests- part II: MCS code Monte Carlo calculationTuan Quoc Tran; Choe, Jiwon; Du, Xianan, et alARTICLE130 Neutronic simulation of China experimental fast reactor start-up tests- part II: MCS code Monte Carlo calculation
2020-12A Multi-Physics Adaptive Time Step Coupling Algorithm for Light-Water Reactor Core Transient and Accident SimulationCherezov, Alexey; Park, Jinsu; Kim, Hanjoo, et alARTICLE133 A Multi-Physics Adaptive Time Step Coupling Algorithm for Light-Water Reactor Core Transient and Accident Simulation
2020-12RAST-K v2-Three-Dimensional Nodal Diffusion Code for Pressurized Water Reactor Core AnalysisPark, Jinsu; Jang, Jaerim; Kim, Hanjoo, et alARTICLE152 RAST-K v2-Three-Dimensional Nodal Diffusion Code for Pressurized Water Reactor Core Analysis
2020-10A PCA compression method for reactor core transient multiphysics simulationCherezov, Alexey; Jang, Jaerim; Lee, DeokjungARTICLE65 A PCA compression method for reactor core transient multiphysics simulation
2020-09Overcoming the challenges of Monte Carlo depletion: Application to a material-testing reactor with the MCS codeDos, Vutheam; Lee, Hyunsuk; Jo, Yunki, et alARTICLE297 Overcoming the challenges of Monte Carlo depletion: Application to a material-testing reactor with the MCS code
2020-09Validation of nuclide depletion capabilities in Monte Carlo code MCSEbiwonjumi, Bamidele; Lee, Hyunsuk; Kim, Wonkyeong, et alARTICLE277 Validation of nuclide depletion capabilities in Monte Carlo code MCS
2020-08Extension of Monte Carlo code MCS to spent fuel cask shielding analysisMai, Nhan Nguyen Trong; Zhang, Peng; Lemaire, Matthieu, et alARTICLE105 Extension of Monte Carlo code MCS to spent fuel cask shielding analysis
2020-07Analysis and comparison of direct inversion and Kalman filter methods for self -powered neutron detector compensationKhoshahval, Farrokh; Zhang, Peng; Lee, DeokjungARTICLE113 Analysis and comparison of direct inversion and Kalman filter methods for self -powered neutron detector compensation
2020-07Interpretation of two SINBAD photon-leakage benchmarks with nuclear library ENDF/B-VIII.0 and Monte Carlo code MCSLemaire, Matthieu; Lee, Hyunsuk; Zhang, Peng, et alARTICLE131 Interpretation of two SINBAD photon-leakage benchmarks with nuclear library ENDF/B-VIII.0 and Monte Carlo code MCS
2020-06Coupling of FRAPCON for fuel performance analysis in the Monte Carlo code MCSYu, Jiankai; Lee, Hyunsuk; Kim, Hanjoo, et alARTICLE316 Coupling of FRAPCON for fuel performance analysis in the Monte Carlo code MCS
2020-06Conceptual design of long-cycle boron-free small modular pressurized water reactor with control rod operationJang, Jaerim; Choe, Jiwon; Choi, Sooyoung, et alARTICLE130 Conceptual design of long-cycle boron-free small modular pressurized water reactor with control rod operation

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