Reactor Physics &
Computational Analysis Laboratory
KEPCO International Graduate Nuclear Graduate School
Advancing the frontiers of computational reactor physics, multi-physics and SMR/PWR and GEN-VI Reactor core design at KEPCO International Nuclear Graduate School, Ulsan, Korea.
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About the Lab
Reactor Physics & Computational Analysis Lab. (RPL)
The Reactor Physics Laboratory & Computational Analysis (RPL) at KINGS (KEPCO International Nuclear Graduate School) is dedicated to cutting-edge research in nuclear reactor physics, computational methods, and advanced core design. Located in Ulsan, Korea — the heart of Korea's nuclear energy industry — RPL bridges rigorous academic research with real-world nuclear engineering applications.
Our work spans the full spectrum of reactor physics computation: from first-principles Monte Carlo neutron transport to industry-grade whole-core diffusion solvers, and from traditional PWR core design to innovative Small Modular Reactor (SMR) concepts. We develop, validate, and deploy in-house simulation tools used to support next-generation nuclear energy systems.
With a strong international team of graduate researchers and deep ties to KEPCO Nuclear Fuel, RPL is uniquely positioned at the intersection of academia and industry.
Quick Facts
  • Institution: KINGS, Ulsan, Korea
  • Department: Nuclear Power Plant Engineering
  • PI: Prof. Joo-il Yoon
  • Focus: Computational Reactor Physics
  • Industry Partner: KEPCO Nuclear Fuel
Principal Investigator
Prof. Jooil Yoon (윤주일)
Associate Professor & Head, Department of Nuclear Power Plant Engineering, KINGS
Prof. Joo-il Yoon is an Associate Professor and the Head of the Department of Nuclear Power Plant Engineering at KINGS. He earned his B.S., M.S., and Ph.D. degrees in Nuclear Engineering from Seoul National University — one of Asia's premier research institutions — completing his doctoral degree in 2021. His academic training is complemented by over a decade of applied industry experience at KEPCO Nuclear Fuel (KNF) from 2008 to 2023, where he contributed to the development of next-generation nuclear design codes and reactor core analysis tools.
Prof. Yoon's research program is defined by a commitment to high-fidelity computational methods for reactor physics, with particular emphasis on Monte Carlo neutron transport, deterministic transport solvers, and AI-enhanced simulation techniques. He leads the development of in-house codes including MONTEX (MONTEcarlo code for neutroniX), SPHINCS(SPH-based Pin-Homogenized Innovative Neutronics Core Simulator) and CROMA (Characteristics-based Reactor Operation and Multigroup Analysis).
Education
01
Ph.D. — Nuclear Engineering
Seoul National University, 2021
02
M.S. — Nuclear Engineering
Seoul National University, 2009
03
B.S. — Nuclear Engineering
Seoul National University, 2007
Professional Experience
Mar. 2026 ~ Present
Associate Professor & Head, Nuclear Power Plant Eng. Dept., KINGS
Jan. 2024 ~ Feb. 2026
Assistant Professor & Head, Nuclear Power Plant Eng. Dept., KINGS
Jan. 2019 ~ Dec. 2023
Manager, Nuclear Fuel Design Department, KEPCO NF
Jan. 2012 ~ Dec. 2018
Principal Nuclear Engineer, KEPCO NF
May 2008 ~ Dec. 2011
Nuclear Engineer, KEPCO NF
Our Team
Lab Members
RPL brings together a diverse, international cohort of graduate researchers united by a shared commitment to advancing computational reactor physics. Our team combines Korean and international scholars pursuing doctoral and master's-level research under the mentorship of Prof. Yoon.
Prof. Joo-il Yoon
Principal Investigator
Associate Professor & Department Head, Nuclear Power Plant Engineering
Hwang Se-yeon
황세연
Ph.D. Student
Ph.D. Student
Wazif
Ph.D. Student
Caesar Katabarwa
Ph.D. Student
Kim Jin-sun
김진선
Ph.D. Student
Yang Kyung-min
양경민
M.S. Student

RPL welcomes applications from motivated graduate students and postdoctoral researchers with backgrounds in nuclear engineering, computational physics, or high-performance computing. Inquiries may be directed to jiyoon@kings.ac.kr.
ph.d. Student
Kim Jin-sun (김진선)
Name: Kim Jin-sun (김진선)
Position: Senior Researcher, KEPCO Nuclear Fuel / Ph.D. Student
Email:
Research Topic:
Advanced core design and reactivity control for commercial PWRs and i-SMRs under boron-free and LEU+ operating conditions
EDUCATION
  • M.S. in Nuclear Engineering, Hanyang University, Seoul, Korea, Feb. 2023
  • B.S. in Nuclear Engineering, Hanyang University, Seoul, Korea, Feb. 2009
PUBLICATIONS
PROFESSIONAL EXPERIENCE
  • KEPCO Nuclear Fuel Co., Ltd. (KNF), Daejeon, Korea
    Senior Researcher, Core Design Team
    Nov. 2011 – Present
  • Reactor core design and analysis for PWRs
  • Burnable absorber design (Gd2O3, Er2O3, HIGA)
  • Fuel management and loading pattern optimization
  • Core analysis using KARMA / ASTRA code system
  • Development of boron-free and flexible operation core design for i-SMR
  • Core design and analysis for commercial PWRs (e.g., APR1400)
RESEARCH INTERESTS
  • Reactor Core Design for PWR and SMR)
  • Burnable Absorber Design (Gd2O3, Er2O3, HIGA)
  • Soluble Boron-Free Operation
  • Neutron Transport and Core Analysis
  • Fuel Management and Loading Pattern Optimization
  • High Burnup and LEU+ Core Design

CV
Ph.D. Student
Hwang Se-yeon (황세연)
Name: Hwang Se-yeon (황세연)
Position: Ph.D. Student
Research Topic: [Research Topic to be added]
EDUCATION
  • [Degree], [University], [Year]
  • B.S[Degree], [University], [Year]
RESEARCH INTERESTS
  • [Interest 1]
  • [Interest 2]
PUBLICATIONS
  • [Publication 1]

CV to be updated
Ph.D. Student
Dawid
Name: Dawid Szymczyna Pawel
Position: Ph.D. Student
GitHub:
Research Topic: Development of DeCART–MANTIS Coupling Framework for High-Fidelity Cross-Section Reconstruction and Whole-Core Neutronic Analysis
EDUCATION
  • Ph.D. in Nuclear Engineering, KEPCO International Nuclear Graduate School (KINGS), Expected 2029
  • M.Sc. in Nuclear Engineering, KEPCO International Nuclear Graduate School (KINGS), 2025
  • M.Sc. in Nuclear Engineering, Wroclaw University of Science and Technology, 2024
  • B.Sc. in Automation and Robotics, Wroclaw University of Science and Technology, 2022
RESEARCH INTERESTS
  • Computational Reactor Physics
  • Cross-Section Functionalization and Multi-Parameter Interpolation for Core Analysis
PUBLICATIONS

CV
Ph.D. Student
Wazif
Name:
Muhammad Wazif Mohd Sallehhudin
Position: Ph.D. Student
Research Topic:
CYNUS code verification for MSR core.
EDUCATION
  • Ph.D in Nuclear Engineering, KEPCO International Nuclear Graduate School(KINGS), Expected 2028
  • M.En in Nuclear Power Plant Engineering, KEPCO International Nuclear Graduate School(KINGS), 2021
  • Bachelor of Engineering(Nuclear),Universiti Teknologi Malaysia(UTM), 2018
RESEARCH INTERESTS
  • Computational Reactor Physics.
  • Generation III+ & IV Reactors.
  • Monte-Carlo Modelling.
PUBLICATIONS
  • Using Machine Learning to Predict the Fuel Peak Cladding Temperature for a Large Break Loss of Coolant Accident

CV to be updated
Ph.D. Student
Caesar Katabarwa
Name: Caesar Katabarwa
Position: Ph.D. Student
Research Topic: [Research Topic to be added]
EDUCATION
  • [Degree], [University], [Year]
  • [Degree], [University], [Year]
RESEARCH INTERESTS
  • [Interest 1]
  • [Interest 2]
PUBLICATIONS
  • [Publication 1]

CV to be updated
M.S. Student
Yang Kyung-min (양경민)
Name: Yang Kyung-min (양경민)
Position: Ph.D. Student
Research Topic: [Research Topic to be added]
EDUCATION
  • [Degree], [University], [Year]
  • [Degree], [University], [Year]
RESEARCH INTERESTS
  • [Interest 1]
  • [Interest 2]
PUBLICATIONS
  • [Publication 1]

CV to be updated
Research
Research Areas
RPL pursues five interconnected research thrusts spanning fundamental neutron physics, advanced deterministic transport, innovative reactor design, computational mathematics, and high-performance computing. Together, these form a comprehensive ecosystem for next-generation reactor simulation.
Monte Carlo Neutron Transport
Development of in-house Monte Carlo code (MONTEX), GPU-accelerated CUDA-based neutron transport, and multigroup cross-section generation using Monte Carlo methods.
Whole-Core Transport & Diffusion Codes
Montecarlo and Pin-by-pin diffusion two-step method development (MONTEX/SPHINCS), MOC-based 2D/1D whole-core transport (CROMA).
SMR Core Design
Innovative Small Modular Reactor core design concepts, boron-free reactivity control strategies, and APR1400 nuclear core design for next-generation power systems.
Computational Methods
Method of Characteristics (MOC), Coarse Mesh Finite Difference (CMFD) acceleration techniques, and depletion methods including predictor-corrector schemes and CRAM.
AI & HPC for Nuclear Applications
GPU/CUDA acceleration for Monte Carlo codes, MPI/OpenMP parallel computing frameworks, and machine learning applications in reactor physics simulation and optimization.
Research Projects
Current Research Projects
RPL is actively engaged in cutting-edge research across several key areas of nuclear engineering. Our projects are supported by various national and international grants, contributing to the advancement of SMR technology, multiphysics simulations, and next-generation nuclear design codes.
SMR Projects
  • Development of a Digital Twin Platform-based Rapid Behavior Prediction System for SMR and Optimal Operator Action Recommendation Technology (Apr. 2025 ~ Dec. 2029, MSIT, Korea)
  • 혁신형 SMR 노심운전지원시스템 계산 엔진 핵심 기술 개발 (May 2024 ~ Dec. 2028, iSMR개발사업단(FNC), Korea)
  • 확산노달방법 기초 3차원 전노심해석 전산코드용 핵심모듈 개발 (Aug. 2024 ~ Dec. 2026, KOFONS(HYU), Korea)
Multiphysics
  • 중성자확산 모델 기반 3차원 중성자속 계산 엔진 개발 (CUPID/SPHINCS Code System) (Dec. 2024 ~ May. 2026, KAERI(ANTS), Korea)
  • Development of Multiphysics Code System with SPHINCS/GIFT (July. 2025 ~ Dec. 2026, SNU)
  • 빅데이터 기반 하이브리드 노심/열수력 코드 개발 (May. 2025 ~ Dec. 2029, MSIT(FNC), Korea)
  • 대형원전 탄력운전 모사 과도계산 노심해석코드 개발 (Oct. 2025 ~ Dec. 2028)
Next-Generation Nuclear Design Code System
  • 군정수 생산을 위한 몬테칼로 코드 개발 (Apr. 2025 ~ Nov. 2027, KEPCO NF, Korea)
  • 초격차 가압경수형 핵연로봉 단위 노심설계 기술 개발 2단계 (Jan. 2026 ~ Dec. 2028, KHNP, Korea)

GEN-IV Reactors (MSR, SFR 등) 과제 기획 중
Methodology
Conventional Two-step Neutronics Calculation Procedure
The conventional two-step neutronics calculation procedure consists of cross-section generation followed by whole-core diffusion calculation and pin power reconstruction.
01
Generation of XS
Lattice calculation with homogenization and group condensing to produce two-group constants from fuel assembly models.
02
Neutron Diffusion Calc.
Whole-core nodal diffusion calculation using homogenized two-group cross-sections on assembly-wise mesh.
03
Pin Power Recons.
Reconstruction of pin-by-pin power distribution from assembly-averaged flux solutions.
Proprietary Reactor Core Analysis Code System in Korea
Multigroup Source Expansion Nodal Method: Better accuracy compared to typical Nodal Expansion Method; Numerical Stability compared to Analytic Nodal Method
Two-Level CMFD Acceleration Method: Accelerating Multigroup nodal calculation with CMFD formulation
Multigroup Pin Power Reconstruction Method: Based on 2D Source Expansion Nodal Method; Better pin power accuracy as incorporating the corner discontinuity factor (CDF)
SMR Design
iSMR Design Status
All-in-one reactor features:
  • Removal of large piping (No LB-LOCA)
  • Canned motor RCP
  • Spiral steam generator
  • Top-mounted ICI
  • In-Vessel CEDM (No REA Acc.)
Core Design specifications:
  • 520 MWth (17x17, 2.4m, 69 FAs)
  • Soluble boron-free operation design
  • Daily Flexible operation
  • 24-month long cycle operation
Key requirements highlighted:
  1. Strong Burnable Absorbers
  1. High Fidelity Nuclear Design Code System
Burnable Absorbers
Highly Intensive Gadolinia/Alumina BA (HIGA)
HIGA (Highly Intensive Gadolinia/Alumina Burnable Absorber)
  • Gd₂O₃-Al₂O₃ composition (Gd₂O₃ 10~20 mol.%)
  • HANA-6 Cladding
  • Active height: 2.4m
  • Al₂O₃ sections (10cm~50cm)
  • Pellet structure with CAP, SPRING, SPACER components
  • Diameter: ∅9.5 / ∅8.35 / ∅8.18
Pin-by-Pin Code
High Performance Pin-by-Pin Nuclear Design Code for iSMR
Purpose of the Research
  • Developing high-fidelity neutronics code for reducing 1% of box and peaking factor uncertainties in commercial nuclear design
  • Box Calculation Uncertainty: Fq,box = 4.78%, Fr,box = 3.66%, Fxy,box = 4.33% → target: 3~4%
For Soluble Boron-Free SMR Neutronics Calculations
  • Control rod inserted operations, flexible operations without soluble boron → Highly and locally heterogeneous flux distribution
  • Too much poison burnable absorbers for controlling excess reactivity → Highly and locally heterogeneous depletion effect
Performance Targets
  • ~20,000 neutronics calculations for nuclear design of one cycle
  • Less than a minute per calculation to complete in a week (2400 minutes)
  • Conventional CPU multi-core cluster systems as target hardware platform
Research Achievements
2D/1D Decoupling Method
Implementing diffusion-based 2D/1D decoupling method with parallelization for pin-by-pin neutronics calculation
CMFD Acceleration Scheme
Developing CMFD acceleration scheme for spatial homogenization and energy group condensation
Equivalence Factors
Determining pin-by-pin equivalence factor and appropriate coarse energy group (GET vs SPH Method)
Two-step Calculation Procedure
Establishing two-step calculation procedure for pin-by-pin nuclear design
Pin-by-Pin Transient Calculation
Developing pin-by-pin transient calculation by modifying node-wise transient calculation scheme
Benchmark Evaluation
Evaluating pin-by-pin code system with benchmark problems
Benchmark
C5G7 MOX Benchmark Problem (Ext. Cases)
Problem Configuration:
  • Assembly Width: 21.42 cm (17x17)
  • Number of Axial Planes: 4 (14.18cm Fuel, 21.42cm Reflector)
  • 7-Group Macroscopic cross-sections: pellet, clad, fission chamber, moderator and control rod
  • Three cases: Unrodded, Rodded A, Rodded B
Results table for Unrodded Case:
Results table for Rodded A Case:
Results table for Rodded B Case:
Key finding: Pin (7G) shows significantly better accuracy than RENUS nodal codes, with RMS errors below 1% for all cases.
Benchmark
VERA Benchmark Problem
Specification:
  • CASL (Consortium for Advanced Simulation of Light Water Reactors) Project published 10 benchmark problem sets based on the Watts Bar NPP 1
  • Westinghouse four-loop type, 193 fuel assemblies (V5H)
Modelling:
  • 4G and 8G XS Generation: nTRACER with ENDF/B-VII.0
  • Reflector XS Generation
  • Pin-wise Calc. with SPH Factor: Pin code
  • Nodal Calculation: RENUS
Results table for CASE 32:
Results table for CASE 27:
Key finding: Pin code shows significantly lower errors than RENUS nodal codes, especially where control rods are inserted. Pin-wise solutions follow severe flux gradients well.
Code System
MONTEX/SPHINCS Two-step Nuclear Design Code System
MONTEX (or PRAGMA):
  • GPU-Based High Performance Monte Carlo Code
  • Utilized for pin-wise homogeneous cross-sections generation
  • Multigroup neutron energy spectrum calculation
SPHINCS:
  • High Fidelity Pin-wise 3D Nuclear Design Code (Fully written in C++)
  • 4 seconds for single iSMR core calculation
  • 1 minute for one cycle depletion calculation
  • Diffusion-based 2D/1D pin-by-pin method
Workflow: Pin-wise XS Generation (MONTEX or PRAGMA) → Pin-wise XS Generation (Hybrid) → Whole Core Pin-wise 3D Calculation (SPHINCS)
Benchmark
APR1400 Benchmark Problem with PRAGMA/SPHINCS
Validation of the PRAGMA/SPHINCS code system against the APR1400 benchmark using MCCARD as reference.
Table 1 - Comparisons of the multiplication factors (K-EFF errors in pcm):
Table 2 - RMS errors of assembly power distributions (%):
Future Research
Future (Ongoing) Researches
SFR Core Analysis Code System Development:
  • Development of high-fidelity neutronics code for Sodium-cooled Fast Reactor (SFR) core analysis
  • Hexagonal fuel assembly geometry with pin-by-pin power distribution
  • Core pin power distribution visualization and error analysis
High Fidelity Multiphysics Code System:
  • Coupling of CFD solver (CUPID) with neutronics codes via preCICE framework
  • FEM-based fuel performance analysis
  • SPHINCS and Monte Carlo code integration
  • Radiation heat transfer and thermal-hydraulics coupling
High-fidelity Numerical Methods for MSR (Molten Salt Reactor):
  • Numerical methods for Molten Salt Fast Reactor (MSFR) core analysis
  • Griffin/Pronghorn code system application
  • Fuel salt, reflector, shield, pump, and heat exchanger modeling
Publications
Publications
319
Total Citations
9
h-index
7
i10-index
2025
  • Full-core fuel analysis of a soluble boron-free SMR: Pellet-cladding interaction issue and enhancing fuel safety through loading pattern design C. Lee, K. Shim, H. Rho, J. Yoon, H. Jeong, Y. Lee Nuclear Engineering and Technology, 57(10), 103709, 2025
  • Performance analysis of a two-step calculation procedure based on Monte Carlo and pin-wise diffusion methods for PWR core design C. Lim, S.J. Kwon, J. Yoon Nuclear Engineering and Technology, 57(8), 103596, 2025
  • Transient analysis of the SPERT III E-core experiment with nTRACER/SPHINCS code system S.D. Pawel, J. Yoon Nuclear Engineering and Technology, 104011, 2025
  • Development of cross-section model of two-step code system considering control rod depletion for soluble boron free operation of SMR J. Park, W. Kim, Y. Kang, W. Jeong, C. Lim, J. Yoon, D. Lee Progress in Nuclear Energy, 189, 105951, 2025
2024
  • Reactor core design with practical gadolinia burnable absorbers for soluble boron-free operation in the innovative SMR J.S. Kim, T.S. Jung, J. Yoon Nuclear Engineering and Technology, 56(8), 3144–3154, 2024
  • Reactor core design with enriched gadolinia burnable absorbers for soluble Boron-Free operation in the innovative SMR J.S. Kim, G. Bae, J. Yoon Nuclear Engineering and Design, 428, 113557, 2024
  • Solution of OECD/NEA PWR MOX/UO2 benchmark with a high-performance pin-by-pin core calculation code H. Hong, J. Yoon Nuclear Engineering and Technology, 56(9), 3654–3667, 2024
2023
  • GPU-accelerated Method of Characteristics with Discontinuous Galerkin Method in Unstructured Mesh Geometry K.M. Kim, H.G. Lee, J. Yoon, H.G. Joo M&C 2023
2021
  • High performance 3D pin-by-pin neutron diffusion calculation based on 2D/1D decoupling method for accurate pin power estimation J. Yoon, H.C. Lee, H.G. Joo, H.S. Kim Nuclear Engineering and Technology, 53(11), 3543–3562, 2021
  • Analysis of C5G7-TD benchmark with a multi-group pin homogenized SP3 code SPHINCS H.H. Cho, J. Kang, J.I. Yoon, H.G. Joo Nuclear Engineering and Technology, 53(5), 1403–1415, 2021
2016
  • Assessment of assembly homogenized two-steps core dynamic calculations using direct whole core transport solutions M. Hursin, T.J. Downar, J.I. Yoon, H.G. Joo Annals of Nuclear Energy, 87, 356–365, 2016
2013
  • Core follow calculation with the nTRACER numerical reactor and verification using power reactor measurement data Y.S. Jung, H.G. Joo, J.I. Yoon American Nuclear Society, 2013
2012
  • Implementation of Advanced Multigroup Nodal and Pin Power Reconstruction Methods into PARCS 3.1 J. Yoon, H.G. Joo, S.H. Ahn US NRC, NUREG/IA-025
2011
  • Application of backward differentiation formula to spatial reactor kinetics calculation with adaptive time step control C.B. Shim, Y.S. Jung, J.I. Yoon, H.G. Joo Nuclear Engineering and Technology, 43(6), 531–546, 2011
  • Benchmark verification of the KARMA/ASTRA code with OECD/NEA and US NRC PWR MOX/UO2 transient problem T.Y. Han, J.I. Yoon, J.H. Kim, C.K. Lee, B.J. Cho 2011
  • Verification & validation of KARMA/ASTRA with benchmark and core-follow analyses J.I. Yoon, S.W. Park, H.S. Park Transactions of the American Nuclear Society, 105, 801–802, 2011
2010
  • Verification of ASTRA Code with PWR MOX/UO2 Transient Benchmark Problem T.Y. Han, J.I. Yoon, J.H. Kim, C.K. Lee, B.J. Cho 2010
2009
  • Multigroup pin power reconstruction with two-dimensional source expansion and corner flux discontinuity H.G. Joo, J.I. Yoon, S.G. Baek Annals of Nuclear Energy, 36(1), 85–97, 2009
2008
  • Two-level coarse mesh finite difference formulation with multigroup source expansion nodal kernels J.I. Yoon, H.G. Joo Journal of Nuclear Science and Technology, 45(7), 668–682, 2008
Teaching
Teaching Courses
Prof. Yoon's teaching portfolio spans both foundational and advanced topics in nuclear engineering, reflecting his dual expertise in theoretical reactor physics and applied computational methods. Courses are offered to graduate students in the Department of Nuclear Power Plant Engineering at KINGS.
Reactor Physics
Foundational graduate course covering neutron lifecycle, criticality, diffusion theory, and the physics of sustained fission chain reactions in thermal and fast reactor systems.
Nuclear Core Design
Practical APR1400 Nuclear Design (PAND) course offering hands-on training in the full-cycle design methodology of Korea's flagship advanced pressurized water reactor.
Computational Methods in Nuclear Engineering
Advanced survey of numerical methods applied to reactor physics problems, including finite difference, finite element, and spectral methods for neutron transport equations.
Neutron Transport Methods
Specialized course on stochastic and deterministic simulation for neutron transport, covering random number generation, variance reduction techniques, tallying, and parallel implementation strategies.
Contact
Get in Touch
We welcome inquiries from prospective graduate students, academic collaborators, and industry partners. Whether you are interested in joining the lab, exploring research partnerships, or discussing collaborative projects in reactor physics or computational nuclear engineering, please don't hesitate to reach out.
Mailing Address
Room 411, Main Building 658-91 Haemaji-ro, Seosaeng-myeon Ulju-gun, Ulsan 45014 Republic of Korea
Direct Contact
📧 jiyoon@kings.ac.kr 📞 +82-52-712-7367
Institution
Lab Website
Visit the official RPL faculty page for the latest updates on research activities, publications, and graduate admissions.
Graduate Admissions
KINGS offers internationally competitive graduate programs in nuclear power plant engineering. Prospective students are encouraged to review the KINGS admissions portal and contact Prof. Yoon directly regarding research fit and available positions.
Pioneering Nuclear Simulation for a Carbon-Free Future
The Reactor Physics Laboratory at KINGS is committed to developing the computational tools, trained researchers, and scientific insights that will define the next generation of safe, efficient, and economical nuclear power. Join us in shaping the future of nuclear engineering.