Research Papers

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

[+] Author and Article Information
W. Peiman

Faculty of Energy Systems and Nuclear Science, University of Ontario Institute of Technology, Oshawa, ON, Canada e-mail: wargha.peiman@uoit.ca

I. Pioro

Faculty of Energy Systems and Nuclear Science, University of Ontario Institute of Technology, Oshawa, ON, Canada e-mail: Igor.Pioro@uoit.ca

K. Gabriel

Faculty of Engineering and Applied Science, University of Ontario Institute of Technology, Oshawa, ON, Canada e-mail: kamiel.gabriel@uoit.ca

1Corresponding author.

Manuscript received August 1, 2014; final manuscript received October 5, 2014; published online March 24, 2015. Assoc. Editor: Asif Arastu.

ASME J of Nuclear Rad Sci 1(2), 021008 (Mar 24, 2015) (10 pages) Paper No: ; doi: 10.1115/1.4026393 History: Received August 01, 2014; Accepted December 09, 2014; Online March 24, 2015

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 supercritical water-cooled reactor (SCWR). Among the Generation IV nuclear-reactor concepts, only SCWRs use water as a coolant. The SCWR concept is considered to be an evolution of water-cooled reactors (pressurized water reactors (PWRs), boiling water reactors (BWRs), pressurized heavy water reactors (PHWRs), and light-water, graphite-moderated reactors (LGRs)), which comprise 96% of the current fleet of operating nuclear power reactors and are categorized under Generation II, III, and III+ nuclear reactors. The latter water-cooled reactors have thermal efficiencies of 30–36%, whereas the evolutionary SCWR will have a thermal efficiency of approximately 45–50%. In terms of a pressure boundary, SCWRs are classified into two categories, namely, pressure-vessel (PV) SCWRs and pressure-channel (PCh) SCWRs. A generic pressure-channel SCWR, which is the focus of this paper, operates at a pressure of 25 MPa with inlet and outlet coolant temperatures of 350°C and 625°C, respectively. The high outlet temperature and pressure of the coolant make it possible to improve thermal efficiency. On the other hand, high operating temperature and pressure of the coolant introduce a challenge for material selection and core design. In this view, there are two major issues that need to be addressed for further development of SCWR. First, the reactor core should be designed, which depends on a fuel-channel design. Second, a nuclear fuel and fuel cycle should be selected. Several fuel-channel designs have been proposed for SCWRs. These fuel-channel designs can be classified into two categories: direct-flow and reentrant channel concepts. The objective of this paper is to study thermal-hydraulic and neutronic aspects of a reentrant fuel-channel design. With this objective, a thermal-hydraulic code has been developed in MATLAB, which calculates fuel-centerline-temperature, sheath-temperature, coolant-temperature, and heat-transfer-coefficient profiles. A lattice code and diffusion code were used to determine a power distribution inside the core. Then, heat flux in a channel with the maximum thermal power was used as an input into the thermal-hydraulic code. This paper presents a fuel centerline temperature of a newly designed fuel bundle with UO2 as a reference fuel. The results show that the maximum fuel centerline temperature exceeds the design temperature limits of 1850°C for fuel.

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Fig. 6

Reentrant channel with annulus gas as thermal insulation for SCWR with liquid moderator

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Fig. 5

High-efficiency channel with ceramic insert (based on Chow and Khartabil [27])

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Fig. 4

Thermal conductivity of selected fuels [17-25]

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Fig. 3

Density profiles of water at 22.064 MPa and 25 MPa

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Fig. 2

Specific-heat profiles of water at 22.064 MPa and 25 MPa

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Fig. 1

Single-reheat cycle for SCW NPP [13]

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Fig. 7

Three-dimensional view of reentrant fuel channel with ceramic insulator

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Fig. 8

Methodology used for thermal-hydraulic/neutronic calculation

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Fig. 9

Heat flux profiles associated with UO2 fuel elements on Rings 1–4 of the fuel bundle as well as the average heat flux for a fuel channel with a maximum thermal power of 9.5 MW

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Fig. 10

Coolant, sheath, and fuel centerline temperature profiles for UO2 fuel

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Fig. 11

Temperature variation across a UO2 fuel element at a location with maximum temperature

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Fig. 12

Temperature variation across a UC fuel element at a location with maximum temperature




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