The axial power and coolant-temperature distributions in a fuel channel of the Generation IV pressure-tube super-critical water-cooled reactor (PT-SCWR) are found using coupled neutronics-thermal-hydraulics calculations. The simulations are performed for a channel loaded with a fresh, 78-element Th-Pu fuel assembly. Neutronics calculations are performed using the DONJON diffusion code using two-group homogenized cross sections produced using the lattice code DRAGON. The axial coolant temperature profile corresponding to a certain axial linear heat generation rate is found using a code developed in-house at University of Ontario Institute of Technology (UOIT). The effect of coolant density, coolant temperature, and fuel temperature variation along the channel is accounted for by generating macroscopic cross sections at several axial positions. Fixed-point iterations are performed between neutronics and thermal-hydraulics calculations. Neutronics calculations include the generation of two-group macroscopic cross sections at several axial positions, taking into account local parameters such as coolant temperature and density and average fuel temperature. The coolant flow rate is adjusted so that the outlet temperature of the coolant corresponds to the SCWR technical specifications. The converged axial power distribution is found to be asymmetric, resembling a cosine shape skewed toward the inlet (reactor top).
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January 2016
Research Papers
Axial Power and Coolant-Temperature Profiles for a Non-Re-Entrant Pressure-Tube Supercritical Water-Cooled Reactor Fuel Channel
Vitali Kovaltchouk,
Vitali Kovaltchouk
1
Faculty of Energy Systems and Nuclear Science,
e-mail: vitali.kovaltchouk@uoit.ca
University of Ontario Institute of Technology
, 2000 Simcoe Street North, Oshawa, ON L1H 7K4
, Canada
e-mail: vitali.kovaltchouk@uoit.ca
1Corresponding author.
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Eleodor Nichita,
Eleodor Nichita
Faculty of Energy Systems and Nuclear Science,
e-mail: eleodor.nichita@uoit.ca
University of Ontario Institute of Technology
, 2000 Simcoe Street North, Oshawa, ON L1H 7K4
, Canada
e-mail: eleodor.nichita@uoit.ca
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Eugene Saltanov
Eugene Saltanov
Faculty of Energy Systems and Nuclear Science,
e-mail: eugene.saltanov@uoit.ca
University of Ontario Institute of Technology
, 2000 Simcoe Street North, Oshawa, ON L1H 7K4
, Canada
e-mail: eugene.saltanov@uoit.ca
Search for other works by this author on:
Vitali Kovaltchouk
Faculty of Energy Systems and Nuclear Science,
e-mail: vitali.kovaltchouk@uoit.ca
University of Ontario Institute of Technology
, 2000 Simcoe Street North, Oshawa, ON L1H 7K4
, Canada
e-mail: vitali.kovaltchouk@uoit.ca
Eleodor Nichita
Faculty of Energy Systems and Nuclear Science,
e-mail: eleodor.nichita@uoit.ca
University of Ontario Institute of Technology
, 2000 Simcoe Street North, Oshawa, ON L1H 7K4
, Canada
e-mail: eleodor.nichita@uoit.ca
Eugene Saltanov
Faculty of Energy Systems and Nuclear Science,
e-mail: eugene.saltanov@uoit.ca
University of Ontario Institute of Technology
, 2000 Simcoe Street North, Oshawa, ON L1H 7K4
, Canada
e-mail: eugene.saltanov@uoit.ca
1Corresponding author.
Manuscript received April 14, 2015; final manuscript received July 19, 2015; published online December 9, 2015. Assoc. Editor: Thomas Schulenberg.
ASME J of Nuclear Rad Sci. Jan 2016, 2(1): 011009 (4 pages)
Published Online: December 9, 2015
Article history
Received:
April 14, 2015
Revision Received:
July 19, 2015
Accepted:
August 7, 2015
Citation
Kovaltchouk, V., Nichita, E., and Saltanov, E. (December 9, 2015). "Axial Power and Coolant-Temperature Profiles for a Non-Re-Entrant Pressure-Tube Supercritical Water-Cooled Reactor Fuel Channel." ASME. ASME J of Nuclear Rad Sci. January 2016; 2(1): 011009. https://doi.org/10.1115/1.4031200
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