scholarly journals Gasification Kinetics in Continuous Supercritical Water Reactors

2020 ◽  
Author(s):  
Brian Pinkard ◽  
John Kramlich ◽  
Per Reinhall ◽  
Igor Novosselov
Keyword(s):  
Supercritical Water ◽  
Water Reactors

scholarly journals Investigation of Corrosion Resistance of Alloys with Potential Application in Supercritical Water-cooled Nuclear Reactors

2019 ◽  
Vol 63 (2) ◽  
pp. 328-332 ◽  
Author(s):  
Ákos Horváth ◽  
Attila R. Imre ◽  
György Jákli
Keyword(s):  
Corrosion Resistance ◽  
Supercritical Water ◽  
Power Plants ◽  
Fuel Cladding ◽  
Pressurized Water ◽  
Water Reactors ◽  
Reactor Types ◽  
Materials Used ◽  
Supercritical Water Cooled Reactor ◽  
Nuclear Applications

The Supercritical Water Cooled Reactor (SCWR) is one of the Generation IV reactor types, which has improved safety and economics, compared to the present fleet of pressurized water reactors. For nuclear applications, most of the traditional materials used for power plants are not applicable, therefore new types of materials have to be developed. For this purpose corrosion tests were designed and performed in a supercritical pressure autoclave in order to get data for the design of an in-pile high temperature and high-pressure corrosion loop. Here, we are presenting some results, related to corrosion resistance of some potential structural and fuel cladding materials.

Design and test of an update of TRAC-BF1 for the modelling of supercritical water reactors

2020 ◽  
Vol 14 (3) ◽  
pp. 191
Author(s):  
Emilio Martinez Camacho ◽  
Jaime Baltazar Morales Sandoval ◽  
J. Manuel Gallardo Villarreal ◽  
Raymundo A. Sánchez Salazar
Keyword(s):  
Supercritical Water ◽  
Water Reactors

scholarly journals Developing and Evaluating Candidate Materials for Generation IV Supercritical Water Reactors - Final Technical Report, INERI 2003-008-K

2006 ◽  
Author(s):  
J. I. Cole ◽  
J. L. Rempe ◽  
T. C. Totemeier ◽  
G. S. Was ◽  
K. Sridharin ◽  
...  
Keyword(s):  
Supercritical Water ◽  
Technical Report ◽  
Water Reactors

Numerical Analysis of Supercritical Water Heat Transfer in Vertical Tube With Different Obstacles

2009 ◽  
Author(s):  
Bo Zhang ◽  
Jianqiang Shan ◽  
Jing Jiang
Keyword(s):  
Heat Transfer ◽  
Critical Point ◽  
Supercritical Water ◽  
Transfer Characteristic ◽  
Vertical Tube ◽  
Bulk Temperature ◽  
Transfer Enhancement ◽  
Heat Transfer Characteristics ◽  
Water Reactors ◽  
The Heat Transfer Coefficient

Supercritical Water Reactors (SCWRs) are essentially water reactors operating at pressure and temperature above critical point. The heat transfer coefficient is relative low when the bulk temperature is above the pseudo-critical point due to the properties of vapor-like fluid. To obtain better heat transfer characteristics, increasing the fluctuation using obstacles is the conventional method. Heat transfer characteristic in vertical tube with different obstacles is numerically investigated under supercritical condition. Numerical simulation is carried out with commercial CFD code Fluent 6.1 and adaptive grid. The results show that The RNG k-ε model with enhanced wall treatment can obtain a reliable result; the blockage ratio and the local temperature have large influence on the heat transfer enhancement. The influence region and decay trend of obstacles are also studied and compared with existing correlations.

Accident analysis of supercritical water reactors during startup

2020 ◽  
Vol 121 ◽  
pp. 103227
Author(s):  
Yuan Yuan ◽  
Jianqiang Shan ◽  
Xiaoying Zhang ◽  
Li Wang
Keyword(s):  
Supercritical Water ◽  
Accident Analysis ◽  
Water Reactors

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

2015 ◽  
Vol 1 (2) ◽  
Author(s):  
W. Peiman ◽  
I. Pioro ◽  
K. Gabriel
Keyword(s):  
Supercritical Water ◽  
Thermal Efficiency ◽  
Nuclear Reactors ◽  
Outlet Temperature ◽  
Channel Design ◽  
Fuel Channel ◽  
Temperature And Pressure ◽  
Water Reactors ◽  
Pressure Channel ◽  
Centerline Temperature

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.

Review of Gasification of Organic Compounds in Continuous-Flow, Supercritical Water Reactors

2018 ◽  
Vol 57 (10) ◽  
pp. 3471-3481 ◽  
Author(s):  
Brian R. Pinkard ◽  
David J. Gorman ◽  
Kartik Tiwari ◽  
John C. Kramlich ◽  
Per G. Reinhall ◽  
...  
Keyword(s):  
Organic Compounds ◽  
Supercritical Water ◽  
Continuous Flow ◽  
Water Reactors

SCWR Rod Bundle Thermal Analysis by a CFD Code

2008 ◽  
Author(s):  
M. Sharabi ◽  
W. Ambrosini ◽  
N. Forgione ◽  
S. He
Keyword(s):  
Heat Transfer ◽  
Supercritical Water ◽  
Power Distribution ◽  
Transport Model ◽  
Rod Bundles ◽  
Fuel Rod ◽  
Reynolds Stress Transport Model ◽  
Rod Bundle ◽  
Transfer Behavior ◽  
Water Reactors

The present paper describes the results of the application of the FLUENT code in the analysis of rod bundle configurations proposed for high pressure supercritical water reactors. The model considers a 1/8 slice of a rod bundle. The details from CFD calculations offer predictions of the circumferential clad surface temperature and of the effect of axial power distribution on the mass exchange between subchannels and on the maximum surface rod temperature. Geometry and boundary conditions are adopted from a previous work that made use of subchannel programs, allowing for a direct comparison between the two techniques. Both the standard k-ε model and the Reynolds stress transport model are used. Conclusions are drawn about the present capabilities in predicting heat transfer behavior in fuel rod bundles proposed for supercritical water reactors.

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