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 (). 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 pressure-channel (PCh) SCWR. Standard neutronics codes DRAGON and DONJON have been coupled with a new thermalhydraulics code developed based on the latest empirical heat-transfer correlation, which allowed for more accurate estimation of basic characteristics of a PCh SCWR. In addition, the computational fluid dynamics (CFD) Fluent code has been used for better understanding of the specifics of heat transfer in supercritical water. Future studies will be dedicated to materials and fuels testing in an in-pile supercritical water loop and developing passive safety systems.
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January 2015
Research Papers
Study on Neutronics and Thermalhydraulics Characteristics of 1200-MWel Pressure-Channel Supercritical Water-Cooled Reactor
Marija Miletic,
Marija Miletic
1
Faculty of Nuclear Sciences and
Physical Engineering,
V Holešovičkách 2,
180 00 Praha 8, Czech Republic
e-mail: marija_miletic@live.com
Physical Engineering,
Czech Technical University
, V Holešovičkách 2,
180 00 Praha 8, Czech Republic
e-mail: marija_miletic@live.com
1Corresponding author.
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Wargha Peiman,
Wargha Peiman
Faculty of Energy Systems and Nuclear Science,
e-mail: wargha.peiman@gmail.com
University of Ontario Institute of Technology (UOIT)
, Oshawa, Ontario L1H 7K4
, Canada
e-mail: wargha.peiman@gmail.com
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Amjad Farah,
Amjad Farah
Faculty of Energy Systems and Nuclear Science,
University of Ontario Institute of Technology (UOIT)
, Oshawa, Ontario L1H 7K4
, Canada
e-mail: amjad.farah@yahoo.com
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Jeffrey Samuel,
Jeffrey Samuel
Faculty of Energy Systems and Nuclear Science,
University of Ontario Institute of Technology (UOIT)
, Oshawa, Ontario L1H 7K4
, Canada
e-mail: Jeffrey.Samuel@uoit.ca
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Alexey Dragunov,
Alexey Dragunov
Faculty of Energy Systems and Nuclear Science,
University of Ontario Institute of Technology (UOIT)
, Oshawa, Ontario L1H 7K4
, Canada
e-mail: Alexey.Dragunov@uoit.ca
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Igor Pioro
Igor Pioro
Faculty of Energy Systems and Nuclear Science,
University of Ontario Institute of Technology (UOIT)
, Oshawa, Ontario L1H 7K4
, Canada
e-mail: Igor.Pioro@uoit.ca
Search for other works by this author on:
Marija Miletic
Faculty of Nuclear Sciences and
Physical Engineering,
V Holešovičkách 2,
180 00 Praha 8, Czech Republic
e-mail: marija_miletic@live.com
Physical Engineering,
Czech Technical University
, V Holešovičkách 2,
180 00 Praha 8, Czech Republic
e-mail: marija_miletic@live.com
Wargha Peiman
Faculty of Energy Systems and Nuclear Science,
e-mail: wargha.peiman@gmail.com
University of Ontario Institute of Technology (UOIT)
, Oshawa, Ontario L1H 7K4
, Canada
e-mail: wargha.peiman@gmail.com
Amjad Farah
Faculty of Energy Systems and Nuclear Science,
University of Ontario Institute of Technology (UOIT)
, Oshawa, Ontario L1H 7K4
, Canada
e-mail: amjad.farah@yahoo.com
Jeffrey Samuel
Faculty of Energy Systems and Nuclear Science,
University of Ontario Institute of Technology (UOIT)
, Oshawa, Ontario L1H 7K4
, Canada
e-mail: Jeffrey.Samuel@uoit.ca
Alexey Dragunov
Faculty of Energy Systems and Nuclear Science,
University of Ontario Institute of Technology (UOIT)
, Oshawa, Ontario L1H 7K4
, Canada
e-mail: Alexey.Dragunov@uoit.ca
Igor Pioro
Faculty of Energy Systems and Nuclear Science,
University of Ontario Institute of Technology (UOIT)
, Oshawa, Ontario L1H 7K4
, Canada
e-mail: Igor.Pioro@uoit.ca
1Corresponding author.
Manuscript received April 26, 2014; final manuscript received September 11, 2014; published online February 9, 2015. Assoc. Editor: Leon Cizelj.
ASME J of Nuclear Rad Sci. Jan 2015, 1(1): 011006 (10 pages)
Published Online: February 9, 2015
Article history
Received:
April 26, 2014
Revision Received:
September 11, 2014
Accepted:
November 14, 2014
Online:
February 9, 2015
Citation
Miletic, M., Peiman, W., Farah, A., Samuel, J., Dragunov, A., and Pioro, I. (February 9, 2015). "Study on Neutronics and Thermalhydraulics Characteristics of 1200-MWel Pressure-Channel Supercritical Water-Cooled Reactor." ASME. ASME J of Nuclear Rad Sci. January 2015; 1(1): 011006. https://doi.org/10.1115/1.4026387
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