scholarly journals Analysis for Performance of Passive Heat Removal by a Water Cooling Panel from a High-Temperature Gas-Cooled Reactor.

1999 ◽  
Vol 65 (635) ◽  
pp. 2489-2497 ◽  
Author(s):  
Shoji TAKADA ◽  
Kunihiko SUZUKI ◽  
Yoshiyuki INAGAKI ◽  
Yukio SUDO
Keyword(s):  
High Temperature ◽  
Heat Removal ◽  
Water Cooling ◽  
High Temperature Gas

On the Design of High-Temperature Gas Turbine Blade Water-Cooling Channels

1978 ◽  
Vol 100 (4) ◽  
pp. 586-591 ◽  
Author(s):  
M. A. El-Masri ◽  
J. F. Louis
Keyword(s):  
High Temperature ◽  
Gas Turbines ◽  
Critical Thickness ◽  
Ambient Pressure ◽  
Rectangular Channel ◽  
Heat Fluxes ◽  
Strong Impact ◽  
Water Cooling ◽  
Cooling Passage ◽  
High Temperature Gas

Centrifugal and Coriolis accelerations have a strong impact on the fluid dynamics and heat transfer in the various schemes of water cooling being actively considered for rotating blades in very high temperature gas turbines. Analytical studies of a thin water film in a rotating rectangular channel open to ambient pressure are presented. First, the dynamics of a thin rotating film indicate that for a certain flow rate it thins out into a stable film under the action of the Coriolis force only for a flow depth below a critical thickness. The value of the critical thickness is a function of the tilt angle between the axis of the cooling passage and the radial. Criteria for nucleation and burnout in high speed liquid films are proposed. These criteria are used to estimate the coolant requirements for representative heat fluxes at different ambient pressures. They suggest that coolant demand increases drastically with pressure. The maximum coolant demand at a fixed heat flux would occur in the neighborhood of five bars.

Goals, Requirements, and Design Implications for the Advanced High-Temperature Reactor

2006 ◽  
Author(s):  
Charles W. Forsberg
Keyword(s):  
High Temperature ◽  
Power Output ◽  
Nuclear Power ◽  
Heat Removal ◽  
Security Requirements ◽  
Electricity Production ◽  
Plant Design ◽  
High Temperature Reactor ◽  
High Temperature Gas ◽  
Electrical Output

The Advanced High-Temperature Reactor (AHTR), also called the liquid-salt-cooled Very High-Temperature Reactor (LS-VHTR), is a new reactor concept that has been under development for several years. The AHTR combines four existing technologies to create a new reactor option: graphite-matrix, coated-particle fuels (the same fuel as used in high-temperature gas-cooled reactors); a liquid-fluoride-salt coolant with a boiling point near 1400°C; plant designs and decay-heat-removal safety systems similar to those in sodium-cooled fast reactors; and a helium or nitrogen Brayton power cycle. This paper describes the basis for the selection of goals and requirements, the preliminary goals and requirements, and some of the design implications. For electricity production, the draft AHTR goals include peak coolant temperatures between 700 and 800°C and a maximum power output of about 4000 MW(t), for an electrical output of ∼2000 MW(e). The electrical output matches that expected for a large advanced light-water reactor (ALWR) built in 2025. Plant capital cost per kilowatt electric is to be at least one-third less than those for ALWRs with the long-term potential to significantly exceed this goal. For hydrogen production, the peak temperatures may be as high as 950°C, with a power output of 2400 MW(t). The safety goals are to equal or surpass those of the modular high-temperature gas-cooled reactor with a beyond-design-basis accident capability to withstand large system and structural failures (vessel failure, etc.) without significant fuel failure or off-site radionuclide releases. These safety goals may eliminate the technical need for evacuation zones and reduce security requirements and significantly exceed the safety goals of ALWRs. The plant design should enable economic dry cooling to make possible wider nuclear-power-plant siting options. Uranium consumption is to be less than that for a LWR, with major improvements in repository performance and nonproliferation characteristics.

scholarly journals Procedure of Active Residual Heat Removal after Emergency Shutdown of High-Temperature-Gas-Cooled Reactor

2014 ◽  
Vol 2014 ◽  
pp. 1-10 ◽  
Author(s):  
Xingtuan Yang ◽  
Yanfei Sun ◽  
Huaiming Ju ◽  
Shengyao Jiang
Keyword(s):  
High Temperature ◽  
Heat Shock ◽  
Steam Generator ◽  
Flow Rate ◽  
Heat Removal ◽  
Emergency Shutdown ◽  
Start Up ◽  
Sudden Load ◽  
High Temperature Gas ◽  
Residual Heat

After emergency shutdown of high-temperature-gas-cooled reactor, the residual heat of the reactor core should be removed. As the natural circulation process spends too long period of time to be utilized, an active residual heat removal procedure is needed, which makes use of steam generator and start-up loop. During this procedure, the structure of steam generator may suffer cold/heat shock because of the sudden load of coolant or hot helium at the first few minutes. Transient analysis was carried out based on a one-dimensional mathematical model for steam generator and steam pipe of start-up loop to achieve safety and reliability. The results show that steam generator should be discharged and precooled; otherwise, boiling will arise and introduce a cold shock to the boiling tubes and tube sheet when coolant began to circulate prior to the helium. Additionally, in avoiding heat shock caused by the sudden load of helium, the helium circulation should be restricted to start with an extreme low flow rate; meanwhile, the coolant of steam generator (water) should have flow rate as large as possible. Finally, a four-step procedure with precooling process of steam generator was recommended; sensitive study for the main parameters was conducted.

Simulation and Analysis of Helium Circulator Trip ATWS Test at Full Power on the HTR-10

2008 ◽  
Author(s):  
Yujie Dong ◽  
Fubing Chen ◽  
Zuoyi Zhang ◽  
Shouyin Hu ◽  
Lei Shi ◽  
...  
Keyword(s):  
High Temperature ◽  
Cooling System ◽  
Power Level ◽  
Reactor Core ◽  
Heat Removal ◽  
Reactor Power ◽  
Code System ◽  
Safety Features ◽  
Forced Cooling ◽  
High Temperature Gas

Safety demonstration tests on the 10 MW High Temperature Gas-cooled Reactor-Test Module (HTR-10) were conducted to verify the inherent safety characteristics of modular High Temperature Gas-cooled Reactors (HTGRs) as well as to obtain the reactor core and primary cooling system transient data for validation of HTGR safety analysis models and codes. As one of these safety demonstration tests, a simulated anticipated transient without scram (ATWS) test called loss of forced cooling by tripping the helium circulator without reactor scram was carried out at 100% rated power level in July, 2005. This paper simulates the reactor transient behaviour during the test by using the THERMIX code system. The reactor power transition and a comparison with the test result are presented. Owing to the negative temperature coefficient of reactivity, the reactor undergoes a self-shutdown after the stop of the helium circulator and keeps subcritical till the end of the test. Due to the loss of forced cooling, the residual heat is slowly transferred from the core to the Reactor Cavity Cooling System (RCCS) by conduction, radiation and natural convection. The thermal response of this heat removal process is investigated. The calculated and test temperature transients of the measuring points in the reactor internals are given and the differences are preliminarily discussed. With respect to the safety features of the HTR-10, it is of most importance that the maximum fuel center temperature is always lower than 1230 °C which is the limited value at the first phase of the HTR-10 project. The simulation and test results show that the HTR-10 has the built-in passive safety features, and the THERMIX code system is applicable and reasonable for simulating and analyzing the helium circulator trip ATWS test.

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