Research

The Versatile Test Reactor (VTR) program under the U.S. Department of Energy (DOE) aims to design an irradiation facility to provide a broad range of testing capabilities for nuclear fuels and materials. The VTR facility is preliminarily proposed as a sodium-cooled 300 MW fast spectrum test reactor to aid in the efforts of research and develop advanced fuels, components, and
instrumentation. The proposed reactor will have five potential locations for instrumented assemblies and/or cartridge loops (CL). These CL will have the ability to be cooled using Sodium, Lead, Molten Salt, or Helium

Texas A&M University is participating in the project to develop instrumentation and tools to use in a proposed high-pressure/high-temperature gas-cooled closed-loop test vehicle. The goal of the research project is to provide measurement techniques and numerical tools to quantify the transport and deposition of fission products in test vehicles and auxiliary components. In the first year of the multi-year project, proposed measurement techniques were performed in a proof-of-concept horizontal test facility. Numerical models are compared to experimental results as validation. Subsequent years of the project will include the implementation of measurement techniques in a scaled model of the proposed VTR high-pressure/high-temperature gas cooled closed-loop test vehicle.

The scopes of the proposed project have following tasks:

• Phenomena Identification and Ranking Table (PIRT): A PIRT will be developed for the VTR GFR experimental facility to identify and rank the important phenomena involved in the generation, transport, deposition, and resuspension of fission products in the VTR GFR Cartridge Loop
• Scaling: It is expected that the scaling approach developed in this study can be applied to investigate the FPVS of the VTR GFR Cartridge Loop.
• Development and Application of Measurement Techniques: The goal is to develop and apply techniques to measure solid fission product concentrations. This will be achieved by combining a source of illumination (laser) and high speed and resolution cameras

• R. Chavez, D. Orea, B. Choi, T. D. Nguyen, N. K. Anand, Y. Hassan & P. Sabharwall (2020): An experimental study of solid and liquid aerosol transport in a horizontal square channel, Aerosol Science and Technology, DOI: 10.1080/02786826.2020.1786002
• Sabharwall, Piyush, Orea, Daniel, Choi, Byung-Hee, Nguyen, Thien, Vaghetto, Rodolfo, Hassan, Yassin A., and Anand, N. K. Fri , Development of Innovative Measurement Techniques for Fission Product Transport Quantification, USDOE Office of Nuclear Energy (NE), DOI:10.2172/1631739

• B.H. Choi, D. Orea, D.T. Nguyen, N.K. Anand, Y. Hassan, P. Sabharwall, Numerical Study of Particle Transport and Deposition in a Horizontal Channel Using a Lagrangian-based Modelling Approach, International Mechanical Engineering Congress & Exposition (IMECE 2019), November 8th-14th, Salt Lake City, Utah, 2019.
• Orea, B.H. Choi, D.T. Nguyen, R. Vaghetto, N.K. Anand, Y.A. Hassan, P. Sabharwall, Experimental Study of Surrogate Particle Transport and Deposition In a Square Channel Using Particle Tracking Technique, International Mechanical Engineering Congress & Exposition (IMECE 2019), November 8th-14th, Salt Lake City, Utah, 2019.
• Orea, B.H. Choi, D.T. Nguyen, R. Vaghetto, N.K. Anand, Y.A. Hassan, P. Sabharwall, An Investigation to Develop Measurement Techniques for Quantifying Fission Product Transport in a Gas-cooled Fast Reactor-Versatile Test Reactor Program, Transactions of American Nuclear Society 120 (1), pp. 1015-1018.
• B.H. Choi, D. Orea, D.T. Nguyen, N.K. Anand, Y. Hassan, P. Sabharwall, Numerical Investigation of Fluid Flow in a Square Channel-Versatile Test Reactor Program, Transactions of American Nuclear Society 121 (1), pp. 1149-1152.
• Orea, R. Chavez, D.T. Nguyen, R. Vaghetto, N.K. Anand, Y.A. Hassan, P. Sabharwall, Experimental Investigation of Surrogate Particle Transport in a Turbulent Channel Flow: Versatile Test Reactor Program, Transactions of American Nuclear Society 121 (1), pp. 1153-1156
• An Investigation to Develop Measurement Techniques for Quantifying Fission Product in a GasCooled Fast Reactor – Versatile Reactor Program, D. Orea, B. H. Choi, D. T. Nguyen, R. Vaghetto, N. K. Anand, Y. A. Hassan, and Piyush Sabharwall, Transactions of American Nuclear Society, Vol. 120, pp. 1015-1018, Minneapolis, Minnesota, June 9-13, 2019.

The Advanced High-Temperature Reactor (AHTR) concept leverages a particle-based fuel format consisting of discrete spherical graphite pebbles arrayed in a packed bed architecture. Thermal regulation achieved via flow of gas (e.g., helium) or liquid (e.g., molten salt) coolants through the void spaces between pebbles in the bed (characteristic pebble diameters: ~ 6 cm, gas cooled; ~ 3 cm, liquid cooled).

Texas A&M is conducting pressure drop and velocity measurements in the versatile experimental facility of randomly packed spheres at various Reynolds numbers. High-fidelity velocity measurements using Time-resolved Stereoscopic Particle Image Velocimetry (TR-SPIV) at the pore scales and near the wall boundary are performed in the matching-refractive-index (MRI) facility.

• Thien Nguyen, Robert Muyshondt, Y. Hassan, and N. K. Anand “Experimental investigation of cross flow mixing in a randomly packed bed and streamwise vortex characteristics using particle image velocimetry and proper orthogonal decomposition analysis” Physics of Fluids, Vol. 31, 025101-025, 2019 https://doi.org/10.1063/1.5079303

• Muyshondt, D.T. Nguyen, Y. Hassan, N.K. Anand, Non-Intrusive Velocity and Temperature Measurements of Buoyant Flows from Inductively Heated Dual-Spheres, Transactions of American Nuclear Society 121 (1), pp. 1633-1636.
• Robert Muyshondt, Thien Nguyen, Yassin Hassan, and N. K. Anand, “High-fidelity Velocity Measurements in a Cross-Flow Plan of a Randomly Packed Bed using Time Resolved Particle Imaging Velocimetry,” ICONE27- 2389, International Conference on Nuclear Engineering -27, Tsukuba, Ibaraki, Japan, May 19-24, 2019.

One major accident of interest in the High Temperature Gas-Cooled Reactors is the Pressurized Conduction Cooldown (PCC) scenario. The PCC scenario results in a loss of forced convection to the core, while the loop stays pressurized since there is no breach in the boundary. The coolant flows through the core driven by buoyancy forces. Experimental data is needed to validate the performances of CFD codes.

An experimental facility representing the core and upper plenum of a HTGR has been designed and constructed to perform natural circulation experiments.  The test sections include multiple vertical heated channels representing the coolant flow paths within a typical prismatic core design. Channels are independently heated.

Particle Image Velocimetry (PIV) and Laser Doppler Velocimetry (LDV) techniques are being applied.

• Anas Alwafi, Thien Nguyen, Yassin Hassan, and N. K. Anand, “Time-resolved particle image velocimetry measurements of a single impinging jet in the upper plenum of a scaled facility of high-temperature gas-cooled reactors,” International Journal of Heat and Fluid Flow, Vol. 76, pp. 113 129, 2019 https://doi.org/10.1016/j.ijheatfluidflow.2019.02.003

• Alwafi, D.T. Nguyen, Y. Hassan, N.K. Anand, Investigation of the flow near the wall of a single impinging jet at the scaled upper plenum of HTGR using TR-PIV, The Proceedings of the International Conference on Nuclear Engineering (ICONE), Tsukuba, Japan, May 2019.
• Investigation of Flow near the Wall of a Single Impinging Jet at the Scaled Upper Plenum of HTGR using TR-PIV, Anas Alwafi, Thien Nguyen, Yassin Hassan, and N. K. Anand, International Conference on Nuclear Engineering -27, Tsukuba, Ibaraki, Japan, May 19-24, 2019
• Anas Alwafi, Thien Nguyen, N. K. Anand, Yassin Hassan, “Time-Resolved Particle Image Velocimetry Measurements and Proper Orthogonal Decomposition Analysis of Jet Impingement in a HTGR Upper Plenum,” Transactions of the American Nuclear Society, Vol. 118, pp. 1120-1122, Philadelphia, Pennsylvania, June 17–21, 2018
• Anas Alwafi, Jae Hyung Park, Saya Lee, Carlos Estrada-Perez, N. K. Anand, Yassin A. Hassan, “Study of the Flow at the Upper Plenum of a Scaled VHTR using PTV,” 17 th International Conference on Emerging Nuclear Energy Systems (ICENES) 2015, Istanbul, Turkey, October 4-8, 2015.
• Jae Hyung Park, Anas Alwafi, Saya Lee, Yassin A. Hassan, N. K. Anand. “Particle Image Velocimetry on a Single Buoyant Plume of the Very High Temperature Gas- Cooled Reactor,” 2015 the American Nuclear Society (ANS) Winter Meeting, Washington, DC, USA, November 8-12, 2015

The helical coil steam generator is a specific type of Tube and Shell Heat Exchanger known for having a higher heat transfer coefficient than many other designs.

We have designed, constructed, and operated three test facilities with main focus on the shell side flow for a specific design of the helical coil steam generator. The three facilities include:

• one-to-one curved single interface between bundles,
• one-to-one curved five bundle test section and
• one-to-one straight five bundle test section.

Studies focus on the effect the complex geometry plays on flow properties. The state of the art facilities use refractive index matching Particle Image Velocimetry (PIV), Laser Doppler Anenometry (LDA), and Hot Film Anenometry, to study the cross-flow behavior of the fluid between bundles.

Pressure measurements across the bundles are also conducted to compare the effect that the different geometry has on current tube and shell pressure drop correlations. Dynamic Pressure Transducers and Pressure Sensitive Paint (PSP) are also used to study the change in pressure around the surface of the individual rods within a bundle.

Experimental Studies

Flow Visualization Studies

Flow Induced Vibration Studies

Computational Fluid Dynamics (CFD) Simulations 

The laboratory has conducted simulations of helical coil steam generator (HCSG) to further investigate the flow in these components using with large eddy simulation (LES) methods. Open-source, high-order spectral element CFD code Nek5000 is largely utilized in the laboratory. Special analysis techniques (Spectral analysis and wavelet analysis) are conducted to extract important information of the flow behavior. Validation of Nek5000 model was performed using the experimental results obtained from the test sections operated in the laboratory.

References

1) Mustafa Alper Yildiz, Haomin Yuan, et al. (2020). “Numerical Simulation of Isothermal Flow Across Slant Five-Tube Bundle with Spectral Element Method Code Nek5000”. Nuclear Technology 206 ,pp. 296–306.

2) Mustafa Alper Yildiz, Elia Merzari, and Yassin Hassan (2019). “Spectral and Modal Analysis of the Flow in a Helical Coil Steam Generator Experiment with Large Eddy Simulation”. International Journal of Heat and Fluid Flow, 2019, 80.108486,pp. 1–16.

• M. Delgado, Y. A. Hassan, and N. K. Anand, “Experimental flow visualization study using particle image velocimetry in a helical coil steam generator with changing lateral pitch geometry,” International Journal of Heat and Mass Transfer, Vol. 133, pp. 756-768, 2019, https://doi.org/10.1016/j.ijheatmasstransfer.2018.12.152
• M. Delgado, S. Lee, Y. A. Hassan, and N. K. Anand, “Flow Visualization Study at the Interface of Alternating Pitch Tube Bundles in a Model Helical Coil Steam Generator using Particle Image Velocimetry “, International Journal of Heat and Mass Transfer, Vol. 122, pp. 614-628, 2018, https://doi.org/10.1016/j.ijheatmasstransfer.2018.02.014

The ASME V&V 30 Subcommittee on Verification and Validation in Computational Nuclear System Thermal Fluids Behavior is supporting a series of verification and validation (V&V) benchmark problems designed to study the scope and key ingredients of the V&V 30 Subcommittee’s charter. This will be achieved by means of the following:

• New, high-quality, state-of-the-art validation data sets obtained using a scaled down experimental facility and instruments with measurement uncertainties estimated using accepted ASME practices for experimental uncertainty (ASME PTC 19.1 Test Uncertainty)
• A stepwise, progressive approach characterized by focusing on each key ingredient individually

The second V&V benchmark problem, Single-Jet Computational Fluid Dynamics (CFD) Numeric Model Validation, in the series has been initiated for the 2019 V&V Symposium and involves simulating a single-jet and plume for single jet experiments at different Reynolds numbers.

The first V&V benchmark problem, Twin-Jet Computational Fluid Dynamics (CFD) Numerical Model, which ran from the 2016 V&V Symposium to the 2018 V&V Symposium, is now closed and a technical report is being written on the results.

Objective of the Second Problem:

Using a select set of data from the single-jet experiments (provided by the ASME and organizers), apply the V&V practices necessary to ensure an appropriately validated computational solution is obtained. The V&V 30 Subcommittee encourages participants from the nuclear community to use whatever V&V practices they would normally use in the context of preparing a document which they might submit to the U.S. Nuclear Regulatory Commission for review. A summary description of the second benchmark problem, including test geometry, boundary conditions for the isothermal single-jet experiments, and experimental data are accessible at the end of the page.

Protocol for Participating in the Second Problem:

The participating organizations or individuals who would like to take part in the benchmark exercise are requested to perform their V&V assessment using the standard protocol and procedures accepted by their engineering communities and sponsoring organizations.  It is noted that this benchmark effort is not intended as a competition among companies or individuals, but rather is intended as a demonstration of the state of the practice in using and applying computational tools to support U.S. Nuclear Regulatory Commission or other regulatory reviews.
The outcomes of this second benchmark effort include lessons learned, review of V&V methods, and understanding of the effectiveness of V&V methods to support modeling and simulation reviews. The results of the various participants will be summarized and compared in a subsequent report.

The experimental data set produced by the Thermal-Hydraulic Research Laboratory, and information on experimental facility, are available for download from the link below.

V&V Benchmark Problem 2

• Anas Alwafi, Thien Nguyen, Yassin Hassan, and N. K. Anand, “Time-resolved particle image velocimetry measurements of a single impinging jet in the upper plenum of a scaled facility of high-temperature gas-cooled reactors,” International Journal of Heat and Fluid Flow, Vol. 76, pp. 113 129, 2019 https://doi.org/10.1016/j.ijheatfluidflow.2019.02.003

• Alwafi, D.T. Nguyen, Y. Hassan, N.K. Anand, Investigation of the flow near the wall of a single impinging jet at the scaled upper plenum of HTGR using TR-PIV, The Proceedings of the International Conference on Nuclear Engineering (ICONE), Tsukuba, Japan, May 2019.
• Investigation of Flow near the Wall of a Single Impinging Jet at the Scaled Upper Plenum of HTGR using TR-PIV, Anas Alwafi, Thien Nguyen, Yassin Hassan, and N. K. Anand, International Conference on Nuclear Engineering -27, Tsukuba, Ibaraki, Japan, May 19-24, 2019
• Anas Alwafi, Thien Nguyen, N. K. Anand, Yassin Hassan, “Time-Resolved Particle Image Velocimetry Measurements and Proper Orthogonal Decomposition Analysis of Jet Impingement in a HTGR Upper Plenum,” Transactions of the American Nuclear Society, Vol. 118, pp. 1120-1122, Philadelphia, Pennsylvania, June 17–21, 2018.