4713213 Nuclear reactor plant housed in a steel pressure vessel with a gas cooled small high temperature reactor

1988 ◽  
Vol 15 (5) ◽  
pp. VI
Keyword(s):  
High Temperature ◽  
Pressure Vessel ◽  
Nuclear Reactor ◽  
Reactor Plant ◽  
High Temperature Reactor

ASME Boiler and Pressure Vessel Code Roadmap for Compact Heat Exchangers in High Temperature Reactors

2020 ◽  
Vol 6 (4) ◽  
Author(s):  
Robert B. Keating ◽  
Suzanne P. McKillop ◽  
Todd Allen ◽  
Mark Anderson
Keyword(s):  
High Temperature ◽  
Pressure Vessel ◽  
Heat Exchangers ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Department Of Energy ◽  
Research Project ◽  
Compact Heat Exchangers ◽  
High Temperature Reactor ◽  
American Society

Abstract The mission of the U.S. Department of Energy (DOE), Office of Nuclear Energy is to advance nuclear power in order to meet the nation's energy, environmental, and energy security needs. Advanced high temperature reactor systems will require compact heat exchangers (CHXs) for the next generation of nuclear reactors. The DOE is sponsoring research to support the development and deployment of CHXs for use in high temperature advanced reactors. The project is being executed by an Integrated Research Project (IRP) that includes university research institutes, national laboratories, manufacturers, and industry experts. The objective is to enable the use of CHX designs in advanced reactor service. A necessary step for achieving this objective is to ensure that the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section III, Division 5 has rules for the construction of CHXs for nuclear service. However, construction rules alone are not sufficient to deploy a CHX in an advanced reactor. A strategy for ASME Boiler and Pressure Vessel Code, Section XI, Inservice Inspection (ISI) of a heat exchanger in an operating nuclear reactor will also be required. The purpose of this ASME Code Roadmap is to identify the research gaps impeding the development of suitable construction and ISI rules for CHXs for high temperature reactor service and to provide a framework to utilize the research project results consistent with the expectations and needs of the industry and future owners.

An Air-Brayton Nuclear-Hydrogen Combined-Cycle Peak- and Base-Load Electric Plant

2007 ◽  
Author(s):  
Charles Forsberg
Keyword(s):  
High Temperature ◽  
Power Output ◽  
Nuclear Reactor ◽  
Peak Load ◽  
Combined Cycle ◽  
Operating Conditions ◽  
Electricity Production ◽  
Spinning Reserve ◽  
High Temperature Reactor ◽  
Base Load

A combined-cycle power plant is proposed that uses heat from a high-temperature nuclear reactor and hydrogen produced by the high-temperature reactor to meet base-load and peak-load electrical demands. For base-load electricity production, air is compressed; flows through a heat exchanger, where it is heated to between 700 and 900°C; and exits through a high-temperature gas turbine to produce electricity. The heat, via an intermediate heat-transport loop, is provided by a high-temperature reactor. The hot exhaust from the Brayton-cycle turbine is then fed to a heat recovery steam generator that provides steam to a steam turbine for added electrical power production. To meet peak electricity demand, after nuclear heating of the compressed air, hydrogen is injected into the combustion chamber, combusts, and heats the air to 1300°C—the operating conditions for a standard natural-gas-fired combined-cycle plant. This process increases the plant efficiency and power output. Hydrogen is produced at night by electrolysis or other methods using energy from the nuclear reactor and is stored until needed. Therefore, the electricity output to the electric grid can vary from zero (i.e., when hydrogen is being produced) to the maximum peak power while the nuclear reactor operates at constant load. Because nuclear heat raises air temperatures above the auto-ignition temperatures of the hydrogen and powers the air compressor, the power output can be varied rapidly (compared with the capabilities of fossil-fired turbines) to meet spinning reserve requirements and stabilize the grid.

Thermo-Economic Performance of a Cogeneration Medium–Small Modular Nuclear Reactor Plant in Canada

2017 ◽  
Vol 3 (3) ◽  
Author(s):  
Gaoming Ge ◽  
Tomohiko Ikegawa ◽  
Koji Nishida ◽  
Carey J. Simonson
Keyword(s):  
High Temperature ◽  
Natural Gas ◽  
Economic Performance ◽  
Nuclear Reactor ◽  
Carbon Tax ◽  
Point Of View ◽  
Integrated Systems ◽  
Cogeneration Plant ◽  
Reactor Plant ◽  
High Co2

Hitachi-GE developed a 300 MWel class modular simplified and medium small reactor (DMS) concept, and the DMS was originally designed for generating electricity only. In this study, the feasibility of a cogeneration DMS plant which supplies both electricity and heat is under investigation. The thermal performance of the DMS plant without or with low-, medium-, or high-temperature thermal utilization (TU) applications is evaluated by numerical simulations. The results show that the electricity generated reduces as the heating requirement of TU application becomes higher. Furthermore, the economic performance of the cogeneration DMS plant is compared with another two integrated systems: (i) DMS plus electric boilers and (ii) DMS plus natural gas boilers, for those three TU applications in Canada. The results illustrate that the DMS plus natural gas boilers system are most economic if there is no carbon tax, but with high-CO2 emissions (up to 180 kton per year). The cogeneration plant performs best as the carbon tax increases up to $40/ton. The cogeneration DMS plant is a promising scheme to supply both electricity and heat simultaneously in the economic-environmental point of view.

Heat Transfer and Fluid Flow Characteristic of U-Shaped Flow Channel for Applications of VHTR

2021 ◽  
Author(s):  
Yasuaki Takayama ◽  
Tetsuaki Takeda
Keyword(s):  
Heat Transfer ◽  
Fluid Flow ◽  
High Temperature ◽  
Natural Convection ◽  
Pressure Vessel ◽  
Cooling System ◽  
Flow Channel ◽  
Working Fluid ◽  
Experimental Apparatus ◽  
High Temperature Reactor

Abstract A very high-temperature reactor (VHTR) is a next-generation nuclear reactor systems. A gas cooling system with natural circulation is considered as a candidate for the pressure vessel cooling system (VCS) of the VHTR. The Japan Atomic Energy Agency is pursuing the design and development of commercial systems such as the 300 MWe gas turbine high-temperature reactor 300 for cogeneration (GTHTR300C). In the VCS of the GTHTR300C, many rectangular flow channels are formed around the reactor pressure vessel (RPV). A cooling panel utilizing the natural convection of air has been proposed. However, the amount of removed heat is inferior to that of cooling by forced convection. In this study, we use an experimental apparatus to simulate the cooling panel of a VCS. The experimental apparatus is a U-shaped flow channel, and the heating surface side is a vertical rectangular flow channel. Air is used as the working fluid. A fine copper wire is used as the porous material. The porosity is varied from 0.996 to 0.999. We perform an experiment to investigate the heat transfer and fluid flow characteristics by natural convection in a vertical rectangular channel heated on one side. Additionally, we perform an experiment on a smooth surface for comparison.

Activation Analysis of the Reactor Pressure Vessel of HTR-PM Reactor Using FLUKA Code

2016 ◽  
Author(s):  
Sheng Fang ◽  
Wenqian Li ◽  
Sida Sun ◽  
Hong Li
Keyword(s):  
High Temperature ◽  
Activation Analysis ◽  
Pressure Vessel ◽  
Technical Specification ◽  
Reactor Pressure Vessel ◽  
Impurity Control ◽  
Activation Products ◽  
High Temperature Reactor ◽  
Specific Activities ◽  
Reactor Pressure

Activation products are the primary radiation source in the maintenance of the reactor pressure vessel (RPV) of High temperature reactor pebble-bed module (HTR-PM). In order to properly plan the maintenance and reduce the related occupational exposure, it is important to correctly evaluate the activation of the impurities in the metal material of HTR-PM’s RPV. In this study, activation of the material of HTR-PM’s RPV is performed using both FLUKA program and experimental formula. Based on the impurity control limit in the technical specification, various impurity nuclides are taken into account. The specific activities of activated radionuclides calculated by the two methods after 40 years irradiation and at the shutdown time are compared. It is demonstrated that FLUKA and experimental formula are in agreement. Also primary contributory activation products at 30 days, 1 year, 5 years and 10 years after the shutdown time are listed.

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