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Research Papers

Axial Power and Coolant-Temperature Profiles for a Non-Re-Entrant Pressure-Tube Supercritical Water-Cooled Reactor Fuel Channel

[+] Author and Article Information
Vitali Kovaltchouk

Faculty of Energy Systems and Nuclear Science,
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,
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,
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 2(1), 011009 (Dec 09, 2015) (4 pages) Paper No: NERS-15-1058; doi: 10.1115/1.4031200 History: Received April 14, 2015; Accepted August 07, 2015

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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References

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Figures

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Fig. 1

Block diagram of the coupled calculations

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Fig. 2

Preliminary concept of the PT-SCWR [7]

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Fig. 3

Cross-sectional view of 78-element PT-SCWR fuel assembly with non-re-entrant fuel channel [7]

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Fig. 4

Axial power production

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Fig. 5

Axial temperature distributions for fuel and coolant

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Fig. 6

Axial distribution of coolant density

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