주한규 | Joo, Han-Gyu

학력 (Education)

• 1996 : Ph.D., Department of Nuclear Engineering, Purdue University, U.S.A.
• 1986 : M.S., Department of Nuclear Engineering, Seoul National University, Korea
• 1984 : B.S., Department of Nuclear Engineering, Seoul National University, Korea

경력 (Career)

• 2009.10 - Present Professor, Dept. of Nuclear Eng., Seoul National University
• 2004.08 - 2009.09 Associate Professor, Dept. of Nuclear Eng., Seoul National University
• 2002.09 - 2004.08 Principal Researcher, Reactor Engineering Department, Korea Atomic Energy Research Institute(KAERI)
• 1998.06 - 2002.08 Senior Researcher, Reactor Engineering Department, KAERI
• 1997.08 - 1998.06 Post Doctoral Researcher, Dept. of Reactor Engineering, KAERI
• 1996.05 - 1997.07 Post Doctoral Research Associate, Dept. of Nuclear Engineering, Purdue University
• 1993.01 - 1996.05 Research Assistant, Dept. of Nuclear Engineering, Purdue University
• 1990.09 - 1992.12 Senior Researcher, Initial Core Design Department, KAERI
• 1986.02 - 1990.08 Researcher, Initial Core Design Department, KAERI

수상 및 영예 (Awards & Honors)

• National Overseas Study Scholarship: Korean Government, 1993-1996
• ANS Best Paper Award: ANS Reactor Physics Division, Winter Meeting, 1995
• KNS Best Paper Award: KNS Fall Meeting, 1999
• KNS Best Paper Award: KNS Spring Meeting, 2000

연구분야 (Research Interest)

• Nuclear Reactor Neutron Transport: authored the DeCART high-fidelity direct whole core transport calculation code which is capable of generating subpin level power distributions for the whole core in one-step with the thermal-hydraulic feedback fully incorporated.
• Reactor Kinetics: authored the initial version of the PARCS spatial kinetics code which has been a member of the NRC’s safety analysis code system and is being used worldwide. (http://www.nrc.gov/about-nrc/regulatory/research/comp-codes.html)
• Neutronics-Thermal/Hydraulics Coupled Reactor Simulation: developed the RELAP/ PARCS and MARS/MASTER coupled code systems for integrated simulation of reactor kinetics and system thermal hydraulics; developed the Numerical Nuclear Reactor which performs high-fidelity, high-resolution, integrated simulation of nuclear reactors by the coupling the whole core neutron transport solutions and the ultra fine computational fluid dynamics (CFD) solutions.
• Numerical Methods for Reactor Simulation: Nodal methods, Krylov subspace linear system solution methods, fuel loading optimization methods
• Fusion Blanket Neutronics: Tritium breeding optimization, shielding analysis

최근 논문 (Paper)

• Vasiliev, A., Canepa, S., Ferroukhi, H., Boyarinov, V. F., Fomichenko, P. A., Joo, H. G., & Ryu, M. (2019). Cross-verification of SUHAM-TD and nTracer reactivity insertion transient solutions without materials homogenization approximation using OECD/NEA C5G7-TD benchmark. Annals of Nuclear Energy, 134, 235-243.

• Lee, J., Facchini, A., & Joo, H. G. (2019). Development of a drift-flux model based core thermal-hydraulics code for efficient high-fidelity multiphysics calculation. Nuclear Engineering and Technology, 51(6), 1487-1503.

• Park, H., & Joo, H. G. (2019). Effective subgroup method employing macro level grid optimization for LWR applications. Annals of Nuclear Energy, 129, 461-471.

• Cherezov, A., Sanchez, R., & Joo, H. G. (2018). A reduced-basis element method for pin-by-pin reactor core calculations in diffusion and SP3 approximations. Annals of Nuclear Energy, 116, 195-209.

• Ban, Y. S., Masiello, E., Lenain, R., Joo, H. G., & Sanchez, R. (2018). Code-to-code comparisons on spatial solution capabilities and performances between nTRACER and the standalone IDT solver of APOLLO3®. Annals of Nuclear Energy, 115, 573-594.

• Lim, C., Joo, H. G., & Yang, W. S. (2018). Development of a fast reactor multigroup cross section generation code EXUS-F capable of direct processing of evaluated nuclear data files. Nuclear Engineering and Technology, 50(3), 340-355.

• Choi, N., Kang, J., & Joo, H. G. (2018, January). Massively parallel method of characteristics neutron transport calculation with anisotropic scattering treatment on GPUs. In Proceedings of the International Conference on High Performance Computing in Asia-Pacific Region (pp. 148-158).

• Park, H., & Joo, H. G. (2018). The Impact of Hydrogen Asymptotic Scattering Approximations on Thermal Reactor Analysis in Deterministic Multi-Group Approach. Transactions, 119(1), 1120-1123.

• Park, H., & Joo, H. G. (2017). Practical resolution of angle dependency of multigroup resonance cross sections using parametrized spectral superhomogenization factors. Nuclear Engineering and Technology, 49(6), 1287-1300.

• Sanchez, R., Tomatis, D., Zmijarevic, I., & Joo, H. G. (2017). Analysis of alpha modes in multigroup diffusion. Nuclear Engineering and Technology, 49(6), 1259-1268.

• Lim, C., Joo, H. G., & Yang, W. S. (2017). Application of Probability Table Method for Unresolved Resonance Self-Shielding in Deterministic Codes. Transactions, 117(1), 1278-1281.