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Research Papers

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

[+] Author and Article Information
Marija Miletic

Faculty of Nuclear Sciences and
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, University of Ontario Institute of Technology (UOIT),
e-mail: wargha.peiman@gmail.com

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 1(1), 011006 (Feb 09, 2015) (10 pages) Paper No: NERS-14-1002; doi: 10.1115/1.4026387 History: Received April 26, 2014; Accepted November 14, 2014; Online February 09, 2015

## 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 $1200-MWel$ 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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## Figures

Fig. 1

Direct single-reheat regenerative 1200-MWel NPP with PCh SCWR [78910]: HP, high pressure; LP, low pressure; IP, intermediate pressure; and HTR, HeaTeR

Fig. 2

Cross-sectional view of PCh-SCWR core

Fig. 3

Three-dimensional view of high-efficiency fuel channel (HEC) [7,9,10]

Fig. 4

Variation of k-effective and burnup as a function of time

Fig. 5

Methodology used for thermalhydraulics/neutronics calculations

Fig. 6

Fuel-centerline-temperature calculations

Fig. 7

Pressure drop along fuel channel

Fig. 8

Two-dimensional view of unit cell based on HEC with 73-element fuel bundle

Fig. 9

Enthalpy, density, specific-heat, and coolant-temperature profiles along heated length of fuel channel

Fig. 10

Dynamic-viscosity, Prandtl-number, thermal-conductivity, and coolant-temperature profiles along heated length of fuel channel

Fig. 11

Actual AHFPs of Rings 1–4 and average AHFP of 73-element bundle string with maximum channel power of 9.3  MWth

Fig. 12

Coolant-, sheath-, and fuel-centerline temperatures and HTC profiles at maximum heat flux

Fig. 13

Temperature variations along radial direction of fuel pellet at maximum heat flux

Fig. 14

Coolant-, sheath-, and fuel-centerline temperatures and HTC profiles at average heat flux

Fig. 15

Temperature variations along radial direction of fuel pellet at average heat flux

Fig. 16

Graphical representation of two-dimensional mesh in three-dimensional space for 2-m tube with 20-cm flow development region

Fig. 17

Realizable k-ε (RKE) simulation for 2-m section of vertical bare tube cooled with SCW flow (experimental data by Kirillov et al. [6,13])

Fig. 18

Contours of radial variation of turbulent kinetic energy near the wall for various axial positions: P=24.1  MPa, G=1496  kg/m2 s, qave=1235  kW/m2, and Tin=320°C

Fig. 19

Cross-sectional and isometric view of subchannel in fuel bundle

Fig. 20

Axial cross-sectional velocity variation at distance (z) from subchannel inlet (z=0.1, 0.25, and 0.4 m) at flow parameters: P=25  MPa, m=4.37  kg/s, q=250  kW/m2, and Tin=357°C

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