Lin-wen Hu

Lin-wen Hu
Lin-wen
Hu
Director, Research and Services
Senior Research Scientist
617-258-5860
NW12-220

Dr. Hu has over 20 years of experience in nuclear reactor design, safety analysis and applications. She currently directs NRL’s research and irradiation services division, which consists of reactor experiments, neutron activation and elemental analysis, neutron beam applications, and reactor physics groups. She is the Technical Lead at MITR as a partner facility of Idaho National Laboratory’s Nuclear Science User Facilities (NSUF) and engaged in the development of a wide range of nuclear applications including advanced nuclear fuel and materials irradiation tests, radioisotope production, licensing, and safety analysis of nuclear reactors. She is currently the Principal Investigator or co-Principal Investigator on several research projects including MITR LEU fuel conversion feasibility study; Fluoride salt-cooled High-temperature Reactor (FHR) design, modeling and safety analysis, fluoride salt and materials irradiation testing; and Transient Reactor Test Facility (TREAT) modeling and instrumentation design. Dr. Hu has authored or co-authored over 190 technical publications.

Education

PhD Nuclear Engineering, MIT, 1996
MS Nuclear Engineering, MIT, 1993.
MS Nuclear Engineering, National Tsing-Hua University, Taiwan, 1991.
BS Nuclear Engineering, National Tsing-Hua University, Taiwan, 1989.

Research Interests

  • Molten salt reactor design and safety analysis.
  • Research reactor design, licensing, operation and applications including fuel, materials, instrumentation, in-pile tests, instrumental neutron activation analysis (INAA), and isotope production.
  • Nanofluids and nanostructured materials for nuclear and other heat transfer applications.
  • Multi-phase flow and heat transfer.
  • Enhanced heat transfer.

Selected Professional Activites

  • Member, National Academies of Sciences Study Committee “State of Molybdenum-99 Production and Utilization and Progress toward Eliminating Use of Highly Enriched Uranium”, 2015–2016.
  • Member, International Expert Diagnosis Assessment Panel, Shanghai Institute of Applied Physics, Chinese Academy of Sciences, 2014–
  • Convenor, International Organization for Standardization (ISO), Research and Test Reactors Working Group under TC085/SC6, 2013–
  • Associate Editor, Journal of Nuclear Engineering and Radiation Science, ASME, Jan 2014–
  • Steering Committee Member, International Group of Research Reactors

Selected Recent Publications

Neutronic design features of a transportable Fluoride-salt-cooled High-temperature Reactor, K. Sun, L. Hu, and C. Forsberg, Journal of Nuclear Engineering and Radiation Science (submitted 2015).

Thermal-hydraulic analyses of transportable Fluoride-salt-cooled High-temperature Reactor with CFD modeling, C. Wang, K. Sun, L. Hu, S. Qiu, and G. Su, Nuclear Technology (in press, 2016).

Development of a Thermal-Hydraulic Analysis Code and Transient Analysis for a Fluoride-salt-cooled High-Temperature Test Reactor, Y. Xiao, L. W. Hu, S. Qiu, D. Zhang, G. Su, and W. Tian, Nuclear Engineering and Radiation Science 1, 011007-1 (2015).

Investigation on mitigating neutron streaming effect for the water-loop facility used in the MIT reactor, K. Sun, D. Carpenter, G. Kohse, and L. Hu, Progress in Nuclear Energy 80, 37-44 (2015).

Computational fluid dynamics analysis for Asymmetric Power Generation in a Prismatic Fuel Block of Fluoride Salt-cooled high-temperature test Reactor, W.-C. Cheng, K. Sun, L.-W. Hu, and C.-C. Chieng, J. of Nuclear Engineering and Radiation Science 1, 011003-1 (2015).

Phenomenology, methods and experimental program for fluoride-salt cooled, high-temperature reactors (FHRs), N. Zweibaum, G. Cao, A.T. Cisneros, B. Kelleher, M.R. Laufer, R.O. Scarlat, J.E. Seifried, M.H. Anderson, C.W. Forsberg, E. Greenspan, L.-W. Hu, P.F. Peterson, and K. Sridharan, Progress in Nuclear Energy 77, 390–405, (2014).

Design and licensing strategies for the fluoride-salt-cooled, high-temperature reactor (FHR) technology, R.O. Scarlat, M.R. Laufer, E.D. Blandford, N. Zweibaum, D.L. Krumwiede, A.T. Cisneros, C. Andreades, C.W. Forsberg, E. Greenspan, L.-W. Hu, and P.F. Peterson, Progress in Nuclear Energy 77, 406–420 (2014).

Validation of a Fuel Management Code MCODE-FM Against Fission Product Poisoning and Flux Wire Measurements of the MIT Reactor, K. Sun, M. Ames, T.H. Newton, and L.W. Hu, Progress in Nuclear Energy 75, 42-48 (2014).

Hydride Fuel Irradiation in MITR-II: Thermal Design and Validation Results, S.J. Kim, D. Carpenter, G. Kohse, and L.W. Hu, Nuclear Engineering and Design 277, 1-14 (2014).

Analysis of the Limiting Safety Systems Settings of a Fluoride Salt Cooled High Temperature Test Reactor, Y. Xiao, L. W. Hu, C. Forsberg, S. Qiu, G. Su, K. Chen, and N. Wang, Nuclear Technology 187, 221-234 (2014).

Effect of Magnetic Field on Laminar Convective Heat Transfer of Magnetite Nanofluids, R. Azizian, E. Doroodchi, T. McKrell, J. Buongiorno, L.W. Hu, and B. Moghtadri, International Journal of Heat and Mass Transfer 68, 94-109 (2014).

Experimental Study of Critical Heat Flux with Alumina-water Nanofluids in Downward-Facing Channels for In-Vessel Retention Applications, G. Dewitt, T. McKrell, J. Buongiorno, and L. W. Hu, Nuclear Engineering and Technology 45, no. 3 (2013).

Experimental Investigation of Transient Critical Heat Flux of Water-Based Zinc-Oxide Nanofluids, V. I. Sharma, J. Buongiorno, T. J. McKrell, and L.W. Hu, International Journal of Heat and Mass Transfer 61, 425-431 (2013).

Convective Heat Transfer Enhancement in Nanofluids: Real Anomaly or Analysis Artifact, N. Prabhat, J. Buongiorno, and L.-W. Hu, Journal of Nanofluids 1, 55-62 (2012).

Pool Boiling Heat Transfer Performance of a Dielectric Fluid with Low Global Warming Potential, E. Forrest, L.-W. Hu, J. Buongiorno, and T. McKrell, Heat Transfer Engineering 34, no. 15 (2013).

Thermal-Hydraulic Analysis for High Enrichment Uranium (HEU) and Low Enrichment Uranium (LEU) Transitional Core Conversion of the MIT Research Reactor, S. J. Kim, L.-W. Hu, and F. Dunn, Nuclear Technology 182, no. 3, 315-332 (2013).