Thermo-hydraulic analysis of the supercritical water-cooled reactor core by porous media approach

2016 ◽  
Vol 110 ◽  
pp. 275-282 ◽  
Author(s):  
M.H. Rahimi ◽  
G. Jahanfarnia
Keyword(s):  
Porous Media ◽  
Supercritical Water ◽  
Reactor Core ◽  
Hydraulic Analysis ◽  
Supercritical Water Cooled Reactor

Pressure Drop Analysis of a Pressure-Channel Type SuperCritical Water-Cooled Reactor

2013 ◽  
Author(s):  
A. Dragunov ◽  
W. Peiman
Keyword(s):  
Pressure Drop ◽  
Supercritical Water ◽  
Nuclear Reactor ◽  
Hydraulic Analysis ◽  
Pressure Drops ◽  
One Dimensional ◽  
Supercritical Water Cooled Reactor ◽  
Thermal Hydraulic Analysis ◽  
Pressure Channel ◽  
Channel Type

Pressure drop calculation and temperature profiles associated with fuel and sheath are important aspects of a nuclear reactor design. The main objective of this paper is to determine the pressure drop in a fuel channel of a SuperCritical Water-cooled Reactor (SCWR) and to calculate the temperature profile of the sheath and the fuel bundles. One-dimensional steady-state thermal-hydraulic analysis was conducted. In this study, the pressure drops due to friction, acceleration, local losses, and gravity were calculated at supercritical conditions.

Various Design Aspects of the Canadian Supercritical Water-Cooled Reactor Core

2015 ◽  
Vol 2 (1) ◽  
Author(s):  
Metin Yetisir ◽  
Rui Xu ◽  
Michel Gaudet ◽  
Mohammad Movassat ◽  
Holly Hamilton ◽  
...  
Keyword(s):  
Supercritical Water ◽  
Thermal Efficiency ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Operating Conditions ◽  
Stress Distributions ◽  
Light Water Reactors ◽  
Core Components ◽  
Water Reactors ◽  
Supercritical Water Cooled Reactor

The Canadian Supercritical Water-Cooled Reactor (SCWR) is a 1200 MW(e) channel-type nuclear reactor. The reactor core includes 336 vertical pressurized fuel channels immersed in a low-pressure heavy water moderator and calandria vessel containment. The supercritical water (SCW) coolant flows into the fuel channels through a common inlet plenum and exits through a common outlet header. One of the main features of the Canadian SCWR concept is the high-pressure (25 MPa) and high-temperature (350°C at the inlet, 625°C at the outlet) operating conditions that result in an estimated thermal efficiency of 48%. This is significantly higher than the thermal efficiency of the present light water reactors, which is about 33%. This paper presents a description of the Canadian SCWR core design concept; various numerical analyses performed to understand the temperature, flow, and stress distributions of various core components; and how the analyses results provided input for improved concept development.

A Supercritical Water-Cooled Small Modular Reactor

2011 ◽  
Author(s):  
R. Duffey ◽  
L. K. H. Leung ◽  
D. Martin ◽  
B. Sur ◽  
M. Yetisir
Keyword(s):  
Supercritical Water ◽  
Reactor Core ◽  
Design Feature ◽  
Normal Operation ◽  
Fluid Temperature ◽  
Balance Of Plant ◽  
Small Modular Reactor ◽  
Accident Scenarios ◽  
Steel Sleeve ◽  
Supercritical Water Cooled Reactor

A new small modular reactor (SMR) is proposed for a 300 MW(e) nuclear generating station. It is referred as the SuperSafe© Reactor (SSR) and is a scaled down version of the Canadian supercritical water-cooled reactor (SCWR), which is designed to operate at supercritical conditions (pressure of 25 MPa and fluid temperature of 625°C) at the turbine inlet with high cycle efficiencies (∼45%). The supercritical turbine technology and associated components used in the balance of plant (BOP) are similar to those in existing supercritical fossil-fired plants. The reactor core consists of fuel channels, which are submerged in a subcooled heavy-water moderator pool inside a low-pressure calandria vessel. Each fuel channel consists of a pressure tube and a ceramic insulator enclosed inside a porous stainless steel sleeve. The moderator provides cooling to fuel channels under normal operation and postulated accident scenarios. This design feature enables the use of a flash-driven passive moderator cooling — an inherent safety feature of the proposed design. A major safety goal is to achieve a passive “no core melt” configuration for the channels and fuel.

European Research Program on Supercritical Water-Cooled Reactor

2020 ◽  
Vol 7 (2) ◽  
Author(s):  
M. Kryková ◽  
T. Schulenberg ◽  
M. Arnoult Růžičková ◽  
A. Sáez-Maderuelo ◽  
I. Otic ◽  
...  
Keyword(s):  
Supercritical Water ◽  
Research Program ◽  
Power Plants ◽  
Large Scale ◽  
Mechanical Stability ◽  
Reactor Core ◽  
Small Modular Reactor ◽  
The Future ◽  
Future Direction ◽  
Supercritical Water Cooled Reactor

Abstract The article summarizes the major achievements and scope of the projects supported by the European Commission with the main focus on the supercritical water-cooled reactor, one of the concepts of the Generation IV reactors. The presented projects are focused mainly on the design of the future reactor, the study of the very specific knowledge gaps related to the technology cooled by supercritical water in large-scale installation as well as the small modular reactor type. Major research topics cover the fields of the constructional materials, their corrosion and mechanical stability, validation of the thermal-hydraulic computing codes and design of the future reactor core including the neutronic calculations. Main goals and future direction of the supercritical water-cooled reactor research program are stated to gain as much as possible from the already performed research not only in the nuclear field—but also in the area of fossil-fueled supercritical-water cooled power plants.

COUPLING FINITE ELEMENTS APPLIED TO HYDRAULIC ANALYSIS IN SATURATED POROUS MEDIA

2019 ◽  
Author(s):  
Murilo Camargo ◽  
Pedro Cleto ◽  
Eduardo Alexandre Rodrigues ◽  
Heber Agnelo Antonel Fabbri ◽  
Osvaldo Luís Manzoli
Keyword(s):  
Porous Media ◽  
Finite Elements ◽  
Saturated Porous Media ◽  
Hydraulic Analysis

scholarly journals Research on Influence of Different Simulation Methods of Bypass Flow in Thermal Hydraulic Analysis on Temperature Distribution in HTR-10

2020 ◽  
Vol 2020 ◽  
pp. 1-8
Author(s):  
Shiyan Sun ◽  
Youjie Zhang ◽  
Yanhua Zheng
Keyword(s):  
Temperature Distribution ◽  
Reactor Core ◽  
Flow Distribution ◽  
Maximum Temperature ◽  
Flow Ratio ◽  
Hydraulic Analysis ◽  
Bypass Flow ◽  
Pebble Bed ◽  
The Core ◽  
Thermal Hydraulic Analysis

In pebble-bed high temperature gas-cooled reactor, gaps widely exist between graphite blocks and carbon bricks in the reactor core vessel. The bypass helium flowing through the gaps affects the flow distribution of the core and weakens the effective cooling of the core by helium, which in turn affects the temperature distribution and the safety features of the reactor. In this paper, the thermal hydraulic analysis models of HTR-10 with bypass flow channels simulated at different positions are designed based on the flow distribution scheme of the original core models and combined with the actual position of the core bypass flow. The results show that the bypass coolant flowing through the reflectors enhances the heat transfer of the nearby components efficiently. The temperature of the side reflectors and the carbon bricks is much lower with more side bypass coolant. The temperature distribution of the central region in the pebble bed is affected by the bypass flow positions slightly, while that of the peripheral area is affected significantly. The maximum temperature of the helium, the surface, and center of the fuel elements rises as the bypass flow ratio becomes larger, while the temperature difference between them almost keeps constant. When the flow ratio of each part keeps constant, the maximum temperature almost does not change with different bypass flow positions.

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