摘要:
The apparatus includes a flowmeter coupled a surface exposed to a flow channel. The flowmeter monitors a flow of coolant. The flowmeter includes a first temperature sensor that generates first temperature data based on measuring a first temperature of a first flowstream, a heating element coupled to the first temperature sensor where the heating element applies heat to the first temperature sensor through an interface, a second temperature sensor generates second temperature data based on measuring a second temperature of a second flowstream, the second temperature sensor being spaced apart from the heating element, and the second temperature sensor being at least partially insulated from the heating element so the second temperature data generated by the second temperature sensor is independent of heat generated by the heating element. A processor calculates a flowrate of the coolant based on the second temperature data and a temperature of the coolant fluid.
摘要:
Provided is a reactor secondary side passive residual heat removal system, comprising: a containment vessel; a steam generator provided with a steam outlet and a water supply inlet; a water tank, the water tank being internally provided with a heat exchanger, the heat exchanger having a heat exchanger inlet and a heat exchanger outlet; and a steam driven pump provided with a steam port, a water inlet and a water outlet, wherein the steam generator, the water tank and the steam driven pump are arranged in the containment vessel, the heat exchanger inlet is in communication with the steam outlet of the steam generator by means of a first pipeline, the heat exchanger outlet is in communication with the water inlet of the steam driven pump by means of a second pipeline, the water outlet of the steam driven pump is in communication with the water supply inlet of the steam generator by means of a third pipeline, and the steam port of the steam driven pump is in communication with the first pipeline by means of a fourth pipeline. The present invention does not rely on an external driving force, thereby greatly reducing the failure probability of the system and improving the safety of the system.
摘要:
A non-invasive eddy current flow meter embedded into a coolant channel for measuring the coolant flow velocity of liquid metal coolant in a nuclear reactor. The eddy current flow meter measures the coolant flow velocity in pool-type nuclear reactors and narrow coolant channels without creating bottlenecks that impede the coolant flow within the nuclear reactors.
摘要:
An electromagnetic flow meter comprises: an excitation member (5) in which a plurality of magnets (5a, 5b, 5c) are arranged spaced apart from one another along an outer circumferential surface of a flow channel (1) through which liquid metal flows, and in which a magnetic field is formed in a direction perpendicular to the outer circumferential surface of the flow channel (1); and electrodes (6a, 6b) provided between the magnets (5a, 5b, 5c) of the excitation member (5), for measuring the voltage generated when the liquid metal crosses the magnetic field. A pulse excitation electric-current supply device (7a) for supplying a pulsed excitation current to the excitation member (5) is provided, whereby a flow speed distribution can be suppressed from being generated in the circumferential direction of the channel (1) even when the flow speed of the liquid metal is low, and the flow rate can be accurately measured.
摘要:
This invention relates to the field of coolant flow in a fluoride salt cooled high temperature reactor. In particular, the invention relates to the discovery of use of nitrogen-16 and/or fluoride-20 decay signature to measure coolant flow. The method of the invention comprises detecting gamma radiation emanating from nitrogen-16 activity and/or fluorine-20 activity within the reactor coolant at a first position along the reactor coolant loop; detecting the gamma radiation emanating from the nitrogen-16 activity and/or fluorine-20 activity within the reactor coolant at a second position along the reactor coolant loop downstream of said first position. The gamma radiation detected from the first position and second position are then cross-correlated, thereby determining the transit time of corresponding gamma activity perturbations viewed at the two detector locations.
摘要:
Disclosed embodiments include electromagnetic flow regulators for regulating flow of an electrically conductive fluid, systems for regulating flow of an electrically conductive fluid, methods of regulating flow of an electrically conductive fluid, nuclear fission reactors, systems for regulating flow of an electrically conductive reactor coolant, and methods of regulating flow of an electrically conductive reactor coolant in a nuclear fission reactor.
摘要:
A system and method of detecting and monitoring flow conditions in the coolant of a nuclear reactor that relies upon acoustic or optical differences in the various flow conditions. The system uses a database of acoustic or optical characteristics of the various known flow conditions being monitored, and a processor that compares the detected acoustic signals with the known acoustic characteristics. The processor uses various methods of discrimination, such as altering or decaying the transmitted signal, to aid in the interpretation and comparison of the signals. The acoustic detection is provided by a pair of sensor assemblies positioned a distance from each other to detect variations in the acoustic patterns associated with the coolant flow. A transmitter and receiver of each sensor assembly can be positioned on opposite sides of the pipe in which the coolant is flowing, or on the same side of the pipe, depending upon which configuration provides the best discrimination between the flow conditions being monitored. The monitoring system is effective for determining, among other things, the existence of bubbles entrained in the coolant, the existence and level of a free surface, the existence of vortex or whirlpool formations, and the existence of entrained solid particulates.
摘要:
A flow measurement probe measures reactor coolant water flow in a cold leg pipe of a nuclear reactor system wherein a lack of elbows precludes the use of elbow tap flow measurements. An elongated probe body disposed in the pipe includes a common dynamic pressure tap in the peripheral wall of the probe body facing about into the direction of coolant water flow. A plurality of static pressure taps are oriented about normal to the direction of coolant water flow. Each of the static pressure taps is operatively connected witch one input of a different differential pressure transmitter. The other input of each of the transmitters is operatively connected with the common dynamic pressure tap. The differential pressure measured by each transmitter is proportional to the square of the magnitude of reactor coolant water flow. The multiple taps in the probe allow multiple redundant flow measurements to be made with a single device, thus avoiding the need for multiple penetrations in the pipe.
摘要:
A flow meter and temperature measuring device comprising a tube with a body centered therein for restricting flow and a sleeve at the upper end of the tube to carry several channels formed longitudinally in the sleeve to the appropriate axial location where they penetrate the tube to allow pressure measurements and temperature measurements with thermocouples. The high pressure measurement is made using a channel penetrating the tube away from the body and the low pressure measurement is made at a location at the widest part of the body. An end plug seals the end of the device and holes at its upper end allow fluid to pass from the interior of the tube into a plenum. The channels are made by cutting grooves in the sleeve, the grooves widened at the surface of the sleeve and then a strip of sleeve material is welded to the grooves closing the channels. Preferably the sleeve is packed with powdered graphite before cutting the grooves and welding the strips.
摘要:
This invention relates to the monitoring and diagnosing of nuclear power plants for its thermal performance using the NCV Method. Its applicability comprises any nuclear reactor such as used for research producing a useful output. Its greatest applicability lies with conventional Pressurized Water Reactor and Boiling Water Reactor nuclear plants generating an electric power. Its teachings of treating fission as an inertial process, a phenomena which is self-contained following incident neutron capture, allows the determination of an absolute neutron flux. This process is best treated by Second Law principles producing a total fission exergy. This invention also applies to the design of fusion thermal systems regards the determination of its Second Law viability and absolute plasma flux.