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Presented By: Nuclear Engineering & Radiological Sciences

NERS Colloquium: Piyush Sabharwall, PhD, Idaho National Laboratory

Thermal Hydraulic Experiments and Modeling to Support Design, Development, and Deployment of Advanced Nuclear Reactors

Thermal Hydraulic Experiments and Modeling to Support Design, Development, and Deployment  of Advanced Nuclear Reactors Thermal Hydraulic Experiments and Modeling to Support Design, Development, and Deployment  of Advanced Nuclear Reactors
Thermal Hydraulic Experiments and Modeling to Support Design, Development, and Deployment of Advanced Nuclear Reactors
ABSTRACT: The development of more-efficient, reliable, and cost-effective nuclear technologies has been accomplished by testing and evaluating the performance of fissile and non-fissile materials in neutron-rich environments, such as Advanced Test Reactor (ATR), High Flux Isotope Reactor (HFIR), etc. In addition, irradiation tests have been done to support the verification and validation of systems and components of nuclear reactors for licensing purposes. Currently, there are very few fast-neutron sources for civilian research. Recently, access to fast-neutron technologies has been fulfilled by using foreign nuclear research reactors, but many research institutions and industries do not have access to this technology and resource, which can limit development of advanced nuclear energy technologies. Furthermore, this limits the expansion of practical knowledge and feasibility in the area of nuclear physics, chemistry, material science, and instrumentation and measurement. Therefore, efforts have begun to develop the Versatile Test Reactor (VTR), a bridge to advance nuclear future. The objective of which is to perform irradiation tests on fuels, materials, and components to understand and evaluate their performance. The access to VTR will significantly increase the knowledge base in terms of irradiation of materials, reactor fuels and components. The inclusion of these experiment vehicles will enable the VTR to perform multiple tests that can support various mission areas while enhancing technical readiness levels for its anticipated life of 50 to 100 years.

BIO: Dr. Piyush Sabharwall is a staff research scientist working in Nuclear System Design and Analysis Division at Idaho National Laboratory (INL). He has expertise in heat transfer, fluid mechanics, thermal design, thermodynamics, and nuclear safety analyses. Over the last few years, he has been researching high temperature heat exchanger design and optimization, system integration and power conversion systems, energy storage, and safety and reliability for Advanced Reactor Concepts. He has exhibited leadership qualities by leading several external partnerships both at regional/international levels, engagements with industry, national laboratories and academia. He has co-authored two books, contributed chapters to technical books on advanced reactors and thermal systems and process heat transfer and published over 100 peer-reviewed publications. He holds an Adjunct Associate Professor appointment in the Department of Mechanical Engineering at Texas A&M University and serves on the ASME Heat Transfer Division's K-9 and K-13 committees. Dr. Sabharwall received the ASME New Faces of Engineering Award in 2011, the ANS Young Member Excellence National Award in 2013, and the ANS Landis Young Member Engineering Achievement Award in 2019.
Thermal Hydraulic Experiments and Modeling to Support Design, Development, and Deployment  of Advanced Nuclear Reactors Thermal Hydraulic Experiments and Modeling to Support Design, Development, and Deployment  of Advanced Nuclear Reactors
Thermal Hydraulic Experiments and Modeling to Support Design, Development, and Deployment of Advanced Nuclear Reactors

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