Lin-wen Hu

Profile Information

Name: Dr. Lin-wen Hu
Institution: Massachusetts Institute of Technology
Position: Director, Research and Services
h-Index
ORCID: 0000-0003-3126-2225
Biography: Dr. Lin-wen Hu has more than 25 years of experience in nuclear systems design, safety, operations, and nuclear technology applications. She is a Senior Research Scientist and Director for research and services at the MIT Nuclear Reactor Laboratory, which operates the 6 MW MIT Reactor. Dr. Hu served on numerous U.S. and international technical committees including National Academies of Sciences study committees, International Group of Research Reactors steering committee, High Performance Research Reactors working group, American Nuclear Society Isotope and Radiation division as chair and executive committee. Dr. Hu is an expert in a wide range of non-power reactor applications including advanced nuclear fuel and materials irradiation tests, radioisotope production and safety analysis, licensing, and operations of nuclear systems. She held a Senior Reactor Operator license for the MIT Research Reactor issued by the U.S. Nuclear Regulatory Commission, and is a licensed Professional Engineer in the Commonwealth of Massachusetts. In addition to research and program management, Dr. Hu also contributes in education as advisor ofmore than 30 MIT SB, SM, PhD student theses and mentored several postdocs and early-career research scientists. Dr. Hu is a Fellow and Life-Time Member of the American Nuclear Society and has authored or co-authored more than 230 technical reports and peer-reviewed conference or journal papers.
Expertise: Microreactors, Test Reactors

Publications:

		"Experimental Measurement and Multiphysics Simulation of Tritium Transport in Neutron Irradiated Flibe Salt" Kieran Dolan, Guiqiu Zheng, Michael Ames, David Carpenter, Lin-wen Hu, Nuclear Technology Vol. 209 2023 515-531 Link Predicting the distribution and release of tritium remains a technical challenge for advanced nuclear reactors with molten Flibe (2LiF-BeF2) salt coolants. Tritium transport models, which are currently used to forecast release behavior, are limited by uncertainty in Flibe-related tritium transport properties and by a lack of relevant benchmark experiments to test input parameters and solution methods. A new test facility has been developed at the Massachusetts Institute of Technology Research Reactor (MITR) to irradiate a molten Flibe target in an ex-core neutron beam port to further investigate tritium transport mechanisms at prototypical reactor conditions. The experiment monitored the time-dependent release of tritium from the salt-free surface and the permeation rate of tritium through the stainless steel Flibe-containing test stand. Measured results were benchmarked with a multiphysics tritium transport simulation to resolve complex effects in the test. Trends in tritium release rates from the irradiation were in agreement with the multiphysics simulation of the test, which combined computational fluid dynamics, radiative heat transfer in participating media, and tritium transport in STAR-CCM+.
		"Hydride fuel irradiation in MITR-II: Thermal design and validation results" Sung Joong Kim, David Carpenter, Gordon Kohse, Lin-wen Hu, Nuclear Engineering and Design Vol. 277 2014 1-14 Link
		"Investigation of Nuclear Graphite and Cf/C Composite in High-Temperature Molten FLiBe Salt in the MIT Reactor" David Carpenter, Guiqiu Zheng, Lin-wen Hu, MIT technical report MIT-NRL-17-03 Vol. 2018 75
		"Radiation resistant fiber Bragg grating in random air-line fibers for sensing applications in nuclear reactor cores" David Carpenter, Lin-wen Hu, Joshua Daw, Kevin Chen, OSA Publishing, Optics Express Vol. 26 2018 11775 Link
		"Tritium Content and Chemical Form in Nuclear Graphite from Molten Fluoride Salt Irradiations" Kieran Dolan, Guiqiu Zheng, David Carpenter, Steven Huang, Lin-wen Hu, Fusion Science and Technology Vol. 76 2020 398-403 Link Advanced reactor applications that use a molten fluoride salt coolant and graphite moderator are under consideration as next-generation energy technologies. For molten salts with lithium or beryllium, such as flibe (2LiF-BeF2), the production of tritium from neutron irradiation is a significant technical challenge. Understanding the expected quantities and mechanisms for tritium retention in graphite is important for designing tritium management strategies in these advanced reactors. In this work, the tritium content of IG-110U graphite from a 2013 in-core flibe irradiation experiment was measured by leaching in water and thermal desorption. Five total samples were tested, with an average measured tritium content per salt-contacting surface area of 3.83 ± 0.25 Ci/m2. The tritium measured from the thermal desorption experiments was primarily in a water-insoluble form. Compared to the overall tritium generation during the irradiation, the total amount of retention in graphite predicted by the desorption measurements is significant.

Presentations:

	"Investigation of Nuclear Graphite and C/C Composite in Molten Li2BeF4 (FLiBe) under Neutron Irradiation" Guiqiu Zheng, David Carpenter, Kieran Dolan, Lin-wen Hu, 19th International Nuclear Graphite Specialists' Meeting September 2-6, (2018) Link