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Innovative Approach to Correlate Heat Transfer Data to SuperCritical CO2 Flowing Upward in a Bare Tube in Forced Convection Regime

23rd International Conference On Nuclear Engineering (ICONE 23)

Chair and Panelist, National Future Energy Strategy, Nuclear Energy Strategic Plan, and Nuclear Power Basis for Future Energy Production in the World as

22rd International Conference on Nuclear Engineering (ICONE 22)

Specifics of Thermophysical Properties and Heat Transfer at Supercritical Pressures

Workshop on Supercritical Fluids and Energy (SFE'13)

Chair, Plenary Session, Nuclear Power Reactors and Fuel Cycle/Uranium Supply

BIT's 3rd New Energy Forum 2013

Keynote Lecture: Nuclear Power as a Basis for Future Electricity Production in the World

22nd International Conference on Nuclear Energy for New Europe (NENE) September, 2013

Power Cycles of Generation III and III+ Nuclear Power Plants

Abstract: Modern advanced thermal power plants have reached very high thermal efficiencies (55–62%). In spite of that, they are still the largest emitters of carbon dioxide into the atmosphere. Therefore, reliable non–fossil fuel energy generation, such as nuclear power, is becoming more and more attractive. However, current nuclear power plants (NPPs) are way behind in thermal efficiency (30–42%) compared to the efficiency of advanced thermal power plants. Therefore, it is important to consider various ways to enhance the thermal efficiency of NPPs. This paper presents a comparison of thermodynamic cycles and layouts of modern NPPs and discusses ways to improve their thermal efficiencies.

Thermal-Hydraulic and Neutronic Analysis of a Reentrant Fuel-Channel Design for Pressure-Channel Supercritical Water-Cooled Reactors

Abstract: To address the need to develop new nuclear reactors with higher thermal efficiency, a group of countries, including Canada, have initiated an international collaboration to develop the next generation of nuclear reactors called Generation IV. The Generation IV International Forum (GIF) Program has narrowed design options of the nuclear reactors to six concepts one of which is the SuperCritical Water-cooled Reactor (SCWR).

Study on Neutronics and Thermalhydraulics Characteristics of 1200-MWel Pressure-Channel Supercritical Water-Cooled Reactor

Abstract: Nuclear power becomes more and more important in many countries worldwide as a basis for current and future electrical energy generation. The largest group of operating nuclear power plants (NPPs) equipped with water-cooled reactors (96% of all NPPs) has gross thermal efficiencies ranging from 30–36%. Such relatively low values of thermal efficiencies are due to lower pressures/temperatures at the inlet to a turbine (4.5–7.8  MPa/257–293°C ). However, modern combined-cycle power plants (Brayton gas-turbine cycle and subcritical-pressure steam Rankine cycle, fueled by natural gas) and supercritical-pressure coal-fired power plants have reached gross thermal efficiencies of 62% and 55%, respectively. Therefore, next generation or Generation IV NPPs with water-cooled reactors should have thermal efficiencies as close as possible to those of modern thermal power plants. A significant increase in thermal efficiencies of water-cooled NPPs can be possible only due to increasing turbine inlet parameters above the critical point of water, i.e., supercritical water-cooled reactors (SCWRs) have to be designed. This path of increasing thermal efficiency is considered as a conventional way that coal-fired power plants followed more than 50 years ago. Therefore, an objective of the current paper is a study on neutronics and thermalhydraulics characteristics of a generic 1200-MW el pressure-channel (PCh) SCWR.

Study on Specifics of Forced-Convective Heat Transfer in Supercritical Carbon Dioxide

Abstract: The appropriate description of heat transfer to coolants at the supercritical state is limited by the current understanding. Thus, this poses one of the main challenges in the development of supercritical-fluids applications for Generation-IV reactors. Since the thermodynamic critical point of water is much higher than that of carbon dioxide (CO 2 ), it is more affordable to run heat-transfer experiments in supercritical CO 2 . The data for supercritical CO 2 can be later scaled and used for supercritical water-based reactor designs. The objective of this paper is, therefore, to discuss the basis for comparison of relatively recent experimental data on supercritical CO 2 obtained at the facilities of the Korea Atomic Energy Research Institute (KAERI) and Chalk River Laboratories (CRL) of the Atomic Energy of Canada Limited (AECL). Based on the available instrumental error, a thorough analysis of experimental errors in wall- and bulk-fluid temperatures and heat transfer coefficient was conducted. A revised heat-transfer correlation for the CRL data is presented. A dimensional criterion for the onset of the deteriorated heat transfer in the form of a linear relation between heat flux and mass flux is proposed. A preliminary heat-transfer correlation for the joint CRL and KAERI datasets is presented.

Development of Gas Cooled Reactors and Experimental Setup of High Temperature Helium Loop for In-Pile Operation

Abstract: Rapidly increasing energy and electricity demands, global concerns over the climate changes and strong dependence on foreign fossil fuel supplies are powerfully influencing greater use of nuclear power. In order to establish the viability of next-generation reactor concepts to meet tomorrow's needs for clean and reliable energy production the fundamental research and development issues need to be addressed for the Generation-IV nuclear-energy systems. Generation-IV reactor concepts are being developed to use more advanced materials, coolants and higher burn-ups fuels, while keeping a nuclear reactor safe and reliable. One of the six Generation-IV concepts is a very high temperature reactor (VHTR).

Thermal Design Aspects of High-Efficiency Channel for SuperCritical Water-Cooled Reactors (SCWRs)

Abstract: This paper focuses on thermal-design aspects of a new pressure channel (i.e., High Efficiency Channel) for SuperCritical Water-cooled Reactors. Objectives of this paper are to estimate heat loss from the coolant to the moderator and to investigate the effects of the insulator thickness and moderator pressure on the overall heat loss. In order to meet the objectives of this study, a steady-state one-dimensional heat-transfer analysis was conducted.

Heat Transfer Profiles of a Vertical, Bare, 7-Element Bundle Cooled with SuperCritical Freon R-12

Abstract: Experimental datasets on simulated fuel bundles are very limited in availability. Supercritical water-cooled nuclear reactors (SCWRs), as one of the six concepts of Generation IV reactors, cannot be designed without such data. Therefore, a preliminary approach using modeling fluids such as carbon dioxide or refrigerants instead of water is practical. One of the supercritical modeling fluids typically used is Freon (R-12) with the critical pressure of 4.136 MPa and the critical temperature of 111.97 °C.

Supercritical-Water Experimental Setup for In-Pile Operation

Abstract: The main goal of the Generation-IV nuclear-energy systems is to address the fundamental research and development issues necessary for establishing the viability of next-generation reactor concepts to meet future needs for clean and reliable energy production. One of the six Generation-IV concepts is a supercritical water-cooled reactor (SCWR), which continues the utilization of well-known light-water-reactor technologies. Research Centre Rez Ltd. has taken part in a large European joint-research project dedicated to Generation-IV light-water reactors with objectives to contribute to the fundamental research and development of the SCWRs by designing and building a test facility called “supercritical water loop (SCWL)”. The main objective of this loop is to serve as an experimental facility for in-core and out-of-core corrosion studies of structural materials, testing and optimization of suitable water chemistry for future SCWRs, studies of water radiolysis at supercritical conditions and nuclear fuels. This paper summarizes the concept of the SCWL, its design, utilization and first results obtained from non-active tests already performed within the supercritical-water conditions.

Developing Empirical Heat-Transfer Correlations for Supercritical CO2 Flowing in Vertical Bare Tubes

Abstract: This paper presents an analysis of three new heat-transfer correlations developed for SuperCritical (SC) carbon dioxide (CO2) flowing in vertical bare tubes. These correlations were developed from the large set of experimental data obtained at Chalk River Laboratories (CRL), AECL (2003). The dataset consists of tests performed in upward flow of CO2 inside 8-mm ID vertical Inconel-600 tube with a 2.208-m heated length.

Progress of International Hydrogen Production Network for the Thermochemical Cu–Cl Cycle

Abstract: This paper presents recent advances by an international team which is developing the thermochemical copper–chlorine (Cu–Cl) cycle for hydrogen production. Development of the Cu–Cl cycle has been pursued by several countries within the framework of the Generation IV International Forum (GIF) for hydrogen production with the next generation of nuclear reactors.

2014 Service Recognition Award

Dr. Pioro received the 2014 Service Recognition Award from the ASME Nuclear Engineering Division for his contributions to nuclear engineering and his role in launching the ASME Journal of Nuclear Engineering and Radiation Science.

Honorary Degree

Received a 2013 Honorary Doctor of the National Technical University of Ukraine, Kiev Polytechnic Institute for his commitment to nuclear engineering education.

������ Research Excellence Award

Recipient of the Senior Research Award for his contributions to education and research.

Education and Communication Award

Recipient of the CNS Education and Communication Award for his significant efforts in improving the understanding of nuclear science and technology among educators, students, and the public.

Canadian Society of Mechanical Engineers

Engineering Institute of Canada

American Society of Mechanical Engineers

Canadian Nuclear Society

Professional Engineers ������

American Nuclear Society