Quá trình công tác

Duy Tan University, Faculty of Electrical and Electronic Engineering: Lecturer

Truong Giang Electric Research and Development Engineer: Conducting research on medium voltage switchgear with focus on vacuum circuit breaker and vacuum interrupter

Advantech, JSC: Channel partner of ANSYS in Vietnam Simulation Engineer: Working with Computational Fluid Dynamics Simulation Software: ANSYS Fluent and ANSYS CFX

Quá trình đào tạo

Purdue University, West Lafayette, Indiana, USA M.S., Nuclear Engineering, August 2015 - July 2018

Pohang University of Science and Technology, Pohang, South of Korea. M.S., Nuclear Engineering, March 2013 - February 2015

Ho Chi Minh University of Science, Ho Chi Minh, Vietnam. B.S., Physics (Honors), September 2008 - October 2012

Công trình nghiên cứu

• Shripad T. Revankar, Hung Nguyen and Mark Brown (2017). Regulatory Assessment of Leakage Through Cracks in Piping Components (Report No. CNSC/PU-NE-17/12). Canadian Nuclear Safety Commission.

• Brown MA, Nguyen H, Revankar ST, Riznic J. (2017). Assessment of Choking Flow Models in RELAP5 for Subcooled Choking Flow Through a Small Axial Crack of a Steam Generator Tube. ASME. International Conference on Nuclear Engineering, Volume 9: Student Paper Competition ():V009T15A046. doi:10.1115/ICONE25-67371.

• Nguyen H, Brown M, Revankar ST, Riznic J. (2016). Experimental Investigation of Subcooled Choking Flow in a Steam Generator Tube Crack. ASME. International Conference on Nuclear Engineering, Volume 5: Student Paper Competition ():V005T15A017. doi:10.1115/ICONE24-60309.

• Song, K. W., Nguyen, T. H., Ha, K. S., Kim, H. Y., Song, J., Park, H. S., ... & Kim, M. H. (2017). Experimental study on two-phase flow natural circulation in a core catcher cooling channel for EU-APR1400 using air-water system. Nuclear Engineering and Design, 316, 75-88.

• Nguyen, Thanh Hung, Song, Ki Won, Revankar, Shripad T., & Park, Hyun Sun (2015). Experimental measurement of two-phase flow parameters in large channel. Proceedings of the 23th International Conference on Nuclear Engineering (ICONE-23), (p. 3737). Japan

Đề tài dự án

Purdue University, Multiphase and Fuel Cell Research Laboratory. Research assistant

Research project: Regulatory assessment of leakage through cracks in piping components, sponsored by Canadian Nuclear Safety Commission (CNSC). The main goal of this project was to develop a comprehensive database and models on the steam generator tube degradation process and resulting leak rates that permit the evaluation of steam generator tubes integrity. The scope of the present research is to develop database on critical flow through steam generator (SG) tube cracks with subcooled liquid flow at the entrance. The experimental data was developed for predictions of leak rates through a crack on a US pressurize water reactor or CANDU steam generator tube. The knowledge of this maximum flow rate through a crack in the steam generator tubes will allow designers to calculate leak rates and design inventory levels accordingly while limiting losses during loss of coolant accidents. The leak rates models were updated for the assessment of steam generator tube inspections. The Homogeneous Equilibrium Model (HEM) and Homogeneous Non-Equilibrium Model (HNEM) were developed for the leak rate of SG tube cracks, providing a physical basis for prediction. The research includes: • An experimental program including development of Facility for Leak Rate Testing (FLRT) at Multiphase Flows and Fuel Cell Research Labs (MFRL), Purdue University; test matrix; fabrication of SG tube crack test specimens, conduction of experiments for various operating conditions. An extensive database on choked flow for small length to diameter ratio cracks is obtained by running tests on the facility. • Development of the choking flow models: Homogeneous equilibrium model and Homogeneous thermal non-equilibrium model. The models predictions are validated against the experimental data. • Assessment of the RELAP5 choking flow models for predicting the leak rate. • Develop a 3D model on CFD analysis software (ANSYS CFX) for the choked flow induced by liquid flashing.

Pohang University of Science and Technology, Division of Advanced Nuclear Engineering. Research assistant

Research project: Two-phase flow natural circulation in a core catcher cooling channel for EU-APR1400, sponsored by National Research Foundation of Korea and BK21+ program Core catcher system is used in the European versions of pressurized water reactor EU-APR1400 to arrests and retains the corium discharged from the reactor vessel in the severe accident scenario. Water at the upside and downside of the core catcher steel plate absorbs decay heat from the molten core. Due to the heating up and boiling of the water in coolant channel, natural circulation flow is generated and the decay heat is removed without external power. In this research, two test facilities were built to simulate the core catcher system: Air-water Experiment Passive Ex-vessel corium retaining and Cooling System (AE-PECS) facility designed and installed at POSTECH and Heated steam-water system Cooling Experiment-PECS constructed in Korea Atomic Energy Research Institute. The main goal of my work in this research is to investigate the two-phase flow characteristics in a large rectangular channel. The objectives of my study are to: • Obtain the database on local two-phase flow parameters such as flow regime, local void fraction, gas velocity, bubble chord length distribution, turbulence intensity. • Study the dependence of local flow parameters on the gas injection flow rate as well as the natural circulation liquid flow rate. • Study the flow transition when it passes the bend connecting the inclined with vertical channel. In order to achieve the objectives, I conducted the following tasks: • Constructing measurement devices including double conductivity probe and data acquisition system. The probes are well verified and calibrated in order to acquire good signal in the experiment. • Building data acquiring and processing program. The data is acquired in high frequency and long time so there is a large number of data points. Therefore, the program is able to manage a large amount of data in an acceptable processing time. • Analyzing the local two-phase flow parameters to study the flow characteristics in the large channel.

Giải thưởng

• Best student paper competition in 25th International Conference on Nuclear Engineering

• Best student poster competition in 24th International Conference on Nuclear Engineering

• Best student paper competition in the 23rd International Conference on Nuclear Engineering