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

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.

PRELIMINARY METHODOLOGY FOR THE ANALYSIS OF THE NATIONAL RESEARCH UNIVERSAL REACTOR USING INTEGRATED SEVERE ACCIDENT MODELLING CODES

2015 ◽  
Vol 4 (1) ◽  
pp. 53-65 ◽  
Author(s):  
A.C. Morreale ◽  
M.J. Brown ◽  
S.M. Petoukhov
Keyword(s):  
Large Scale ◽  
Aluminum Matrix ◽  
Heat Sinks ◽  
Severe Accident ◽  
Accident Analysis ◽  
Pressurized Water ◽  
Fuel Design ◽  
Water Reactors ◽  
Accident Case ◽  
Accident Modelling

The National Research Universal (NRU) Reactor is a multi-purpose research reactor located at Atomic Energy of Canada Limited (AECL) Chalk River Laboratories. The severe accident case for the NRU has been explored through deterministic and probabilistic safety analysis (PSA) including multi-level PSAs that detail the progression and consequences of a severe accident in the NRU. These previous calculations lack the interconnected and comprehensive features of a full severe accident modelling code that is now the standard for severe accident analysis of power reactors. It was of interest within AECL to evaluate modern severe accident modelling codes to the NRU reactor case to enhance the understanding of accident progression and predict the system damage and radiation release consequences of a severe accident, which is a very low probability event. The NRU is smaller and operates at a lower power than the large scale power reactors (e.g., pressurized heavy water reactors, pressurized water reactors, and boiling water reactors) that these codes were designed to analyze. Additionally, the NRU has a unique design different from the power reactors and several features relevant to severe accidents including filtered venting, large passive heat sinks, and a dispersion fuel design of uranium-silicide in an aluminum matrix. The major severe accident analysis codes available to AECL and their applicability to the NRU are explored in this paper. In addition, a preliminary strategy for employing the most applicable codes to the NRU for the purposes of severe accident modelling is proposed.

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

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.

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