1069results about "Cooling arrangement" patented technology

The invention relates to the technical field of nuclear reactor safety, and particularly discloses a reactor waste heat export, recovery and comprehensive utilization system and method based on a heat pipe, and the system comprises a nuclear energy generation system, a secondary loop energy conversion system, a primary loop waste heat discharge system, a secondary loop waste heat discharge system, and a waste heat recovery and comprehensive utilization system. According to the waste heat recycling system, the heat pipe can be used for heat transfer and storage under the peak regulation operation condition, the nuclear heat use efficiency is remarkably improved, the requirements of various types of energy sources such as electricity and heat are met, and passive export of reactor core waste heat can be achieved through multi-natural-circulation coupling cooperative operation under the condition of a whole-field outage accident. By means of the efficient heat transfer characteristic and the modularization characteristic of the heat pipes, independent operation of the heat pipes can be achieved, the problem of single-point failure of the system can be avoided, meanwhile, separation of a reactor coolant and a cooling medium of a waste heat removal system can be achieved, the inherent safety of a nuclear reactor is remarkably improved, and the requirements of multi-purpose scenes are met; and the purpose of fully utilizing resources is achieved.

The invention discloses a pressurized water reactor primary loop boron concentration prediction and control method and system, and the method comprises the steps: obtaining an original helium flow sequence of a pressurized water reactor primary loop, and obtaining M boron concentration IMF components IMF1,..., IMFM; the M boron concentration IMF components IMF1,..., IMFM are put into a trained LSTM model for prediction, predicted values Y1, Y2,..., YM of the IMF components are obtained, and error coefficients delta Y1, delta Y2,..., delta YM corresponding to the boron concentration IMF components are calculated according to the predicted values Y1, Y2,..., YM of the IMF components; and inputting error coefficients delta Y1, delta Y2,..., delta YM corresponding to each boron concentration IMF component into the trained DenseNet prediction model to obtain an optimal IMF component predicted value # imgabs0 # of the boron concentration signal sequence, based on a DEM-DAC algorithm, controlling the boron concentration of the pressurized water reactor primary loop according to the optimal IMF component predicted value # imgabs1 # of the boron concentration signal sequence, and calculating the boron concentration of the pressurized water reactor primary loop according to the optimal IMF component predicted value # imgabs0 # of the boron concentration signal sequence. The method and the system can solve the technical problems that the boron concentration fluctuation of the primary loop is large and the unit load regulation performance is poor during the reactor variable load operation period.

The present application relates to a method and system for filling water in a primary loop after overhaul charging of a nuclear power unit, comprising: performing preparation work before filling water; after the preparation work is completed, performing switch operation test of a to-be-tested actuator and water filling and exhaust work of a flowmeter; after the switch operation test of the to-be-tested actuator is completed, performing performance test of a high-pressure safety injection pump; after the performance test of the high-pressure safety injection pump is completed, performing performance test of a low-pressure safety injection pump; after the performance test of the low-pressure safety injection pump is completed, filling water in the primary loop according to a preset water filling mode; judging whether a real-time water level of a reactor pool reaches a full water level; if yes, performing alarm action test. The present application can solve the problem of long time consumption, reduce workload, significantly improve work efficiency, reduce work risk and effectively save production cost by completing the preparation work in advance and arranging the alarm action test at the end.

The invention provides a reactor core structure capable of controlling an assembly clearance and a manufacturing method thereof. The reactor core structure comprises a reactor core heat management element and a filling material; wherein the base body is provided with a plurality of assembly holes; the reactor core heat management element comprises a fuel rod or a heat pipe; the reactor core heat management element is assembled in the assembly hole of the base body; the filling material is filled in an assembly gap between the reactor core heat management element and the assembly hole; and the filling material is a shape memory alloy. The reactor core structure can improve the safety margin of the heat pipe reactor.

The embodiment of the invention provides an integrated small reactor passive residual heat removal test heat exchange structure, system and method, and relates to the technical field of nuclear reactor safety system tests. The system aims to solve the problems that in an integrated small reactor passive quay crane test system, split type and pipeline connection are adopted for containment cooling, so that the system is tedious and large in occupied space. And the heat exchange structure is directly connected with the containment simulation device and the cooling water tank through the shared heat exchange part. The test system comprises the test heat exchange structure, and the test method is implemented by adopting the test system. And heat is transferred from the high-temperature containment simulation device to the low-temperature cooling water tank fluid through the heat exchange part. Redundant connecting parts of the common heat exchange part are omitted, and occupied space is reduced. Efficient and rapid heat transfer is achieved, waste heat is discharged in time, safety is guaranteed, and due to the compact structural design, the whole test device is more reasonable in layout, saves space and is convenient to install, debug and maintain.

An installation includes: a plurality of pilings securable to a bed under a surface of a body of water; a base structure disposed atop the plurality of pilings; and a module disposable on the base structure, wherein the module is positioned and securable on the base structure after being floated on the surface of the body of water over the base structure.

The invention provides a nuclear power plant primary side post-accident passive residual heat removal system and a control method, the system comprises a reactor coolant system, a containment inner water tank, a collecting mechanism and a passive water replenishing tank, and the reactor coolant system comprises a hot section and a cold section; the containment inner water tank is installed on the inner wall of the steel containment, the heat exchanger is arranged in the containment inner water tank, the hot section is connected with an inlet pipeline of the heat exchanger, the cold section is connected with an outlet pipeline of the heat exchanger, and the collecting mechanism is arranged on the inner wall of the steel containment and communicated with the containment inner water tank through a connecting pipe. The collecting device is used for collecting condensate on the inner wall of the steel containment in a grading manner along the axial direction of the containment and enabling the condensate to flow to a water tank in the containment through a connecting pipe; one end of the passive water replenishing tank is connected with the water tank in the containment through a runner pipe, and the passive water replenishing tank replenishes water to the water tank in the containment when the liquid level of the water tank in the containment drops to a target value.

The invention provides a long-term passive cooling system of a compact small reactor, which is characterized in that a reactor and steam generators are arranged in a containment, the steam generators are connected with the reactor, and each steam generator is provided with a secondary side residual heat removal system; the liquid storage chamber is mounted on the outer wall surface of the containment in a surrounding manner; the passive heat exchanger is arranged in the liquid storage chamber and is connected with a secondary side residual heat removal system of the corresponding steam generator, so that steam of the steam generator enters the passive heat exchanger for heat exchange; the plurality of passive heat pipes are mounted on the wall surface of the containment in a penetrating manner and are used for guiding heat in the containment to the liquid storage chamber. The secondary side waste heat removal system is arranged on the secondary side of the steam generator and used for relieving non-loss-of-water accidents, short-term and long-term reactor core cooling is achieved outside the containment through the passive heat exchanger combining water cooling and air cooling, the peak pressure of the containment after the accidents is effectively reduced, long-term passive heat carrying is achieved, and the unit operation safety is improved.

This invention discloses a firecracker-style controlled nuclear fusion and fission synergistic power generation device and method. The power generation device includes a fission-fusion reaction system, a circulating cooling system, and other accessories. The fission-fusion reaction system includes a neutron gun, fuel spheres, a fuel sphere storage and delivery device, and a vacuum reactor. The circulating cooling system includes a circulating fan, a voltage regulator, a steam turbine, a steam generator, delivery pipelines, and a water pump. Other accessories include an energy storage and control system and a vacuum pump. This invention innovatively integrates the start-up mechanism, drawing on the firecracker-style multi-reactor relay concept, combining nuclear fission and nuclear fusion. It utilizes the high temperature and pressure generated by nuclear fission to start nuclear fusion, reducing start-up difficulty and cost. Simultaneously, it efficiently converts and utilizes energy, employing helium for heat transfer. Helium's high thermal conductivity, low density, high specific heat capacity, and chemical inertness allow it to effectively carry away the heat stored in the phase change microspheres, comprehensively collecting and converting nuclear reaction heat energy, thus improving energy conversion efficiency.

A small-scale nuclear reactor cooling device and a cooling method are provided. The small-scale nuclear reactor cooling device includes a reactor vessel in which cooling water circulates along a flow path structure, and a cooling water storage unit that is provided inside the reactor vessel and is filled with the cooling water based on the circulating flow, and when a molten core occurs in the reactor vessel due to the occurrence of a serious accident, the cooling water storage unit comes into contact with the molten core via the cooling water filled inside to cool the molten core and prevent the molten core from leaking out to the outside.

The invention discloses a dual-mode self-starting residual heat removal system for a heat pipe reactor, which belongs to the technical field of nuclear reactor safety, and comprises a heat pipe heat exchanger for dissipating heat of a plurality of heat pipes; the air cooling system comprises an air channel formed in the heat pipe heat exchanger, the heat pipe is located in the air channel, circulation openings are formed in the top and the bottom of the air channel respectively, and when the heat pipe heat exchanger fails, the circulation openings are opened to communicate with the external atmosphere. The waste heat removal heat exchanger is provided with an inner-layer cavity and an outer-layer cavity, the inner-layer cavity and the outer-layer cavity are both filled with molten metal used for heat conduction, the heat pipe penetrates through the inner-layer cavity, and a self-opening assembly is arranged between the inner-layer cavity and the outer-layer cavity. And the self-opening assembly is automatically opened. According to the invention, self-starting dual-mode residual heat removal can be realized under extreme conditions, and the residual heat removal capability and safety margin of the heat pipe reactor system under accident and shutdown conditions are improved.

The invention relates to the technical field of cooling of particle accelerators and nuclear power equipment, and discloses a cooling device suitable for high-power rays, which comprises an absorber assembly, a cooling water system assembly, a vacuum tube assembly, a shield assembly and a support assembly. The absorber assembly is formed by welding internal graphite and external copper through a brazing process, and air holes penetrating to a graphite layer are formed in the copper and communicated with an air cavity in the graphite to form a vacuum sealing cavity. The cooling water system assembly comprises a water segregator with two layers of water inlet ring cavities and two layers of water outlet ring cavities, and uniform flow distribution is ensured. And the absorber cooling water pipe is embedded into the copper and is connected with the U-shaped pipe to form a cooling loop. The electron beam bombards the graphite to generate heat, and the heat is transferred to the copper and taken away by cooling water in the cooling water pipe embedded in the copper. The shield assembly is arranged around the vacuum tube assembly to prevent radiation leakage. The cooling device realizes effective heat dissipation of 8GeV high-power rays, and has excellent radiation resistance and extremely long fatigue life.

The utility model discloses a hydrogen removal system which comprises a first hydrogen removal module, a second hydrogen removal module and a third hydrogen removal module, the reactor coolant system comprises a reactor coolant system main loop and a voltage stabilizer; the first hydrogen removal module is connected with a main loop of the reactor coolant system, and the first hydrogen removal module is used for carrying out circulating degassing on a coolant in the main loop of the reactor coolant system before shutdown, so that the hydrogen concentration of the coolant is within a preset range; the second hydrogen removal module is connected with the voltage stabilizer; the second hydrogen removal module is used for reducing the hydrogen concentration of a gaseous space in the voltage stabilizer after shutdown; and the third hydrogen removal module is connected with the reactor coolant system main loop and is used for carrying out chemical hydrogen removal on the reactor coolant system main loop after the reactor is shut down and the temperature of the coolant is lower than the preset temperature until the hydrogen concentration is reduced to be lower than the preset concentration. Therefore, long hydrogen removal time of the reactor coolant system in the maintenance process is avoided, and the hydrogen removal efficiency is improved.

The invention provides a shell-and-tube phase change heat storage system for a nuclear reactor containment and a nuclear accident heat energy management method, and belongs to the technical field of nuclear safety engineering and heat energy storage. The problems that an existing shell-and-tube heat reservoir is designed for industrial waste heat recovery and is not suitable for nuclear accidents, and the internal space structure of a containment is complex and limited are solved. The heat storage units are distributed in parallel, and each heat storage unit comprises a shell and a plurality of heat exchange tube bundles arranged in the shell; the heat storage unit has an interchangeable heat storage medium arrangement mode which comprises the first mode that the space in the heat exchange tube bundle is filled with a phase change material, and accident steam flows through a shell side channel between the shell and the heat exchange tube bundle; in the second mode, the shell side space between the shell and the heat exchange tube bundle is filled with the phase change material, and accident steam flows through an in-tube channel of the heat exchange tube bundle; and the accident steam passively flows through the channel which is not filled with the phase change material under the driving of the pressure difference. The method is used in the phase-change heat storage field.

A controlled reactor comprises a reactor core thermally coupled to one or more heat pipes and an active cooling loop. A fluid may be circulated through the active cooling loop. A heat exchanger is thermally coupled to the active cooling loop and extracts heat from the fluid as the fluid is circulated through the active cooling loop. A heating system may be provided to deliver the heat extracted by the heat exchanger to a community. A thermoelectric generator may be provided to convert heat extracted by the heat pipes to electricity for delivery to the community.

The invention belongs to the technical field of heat pipe reactors, and discloses a hexagonal annular heat pipe cooling reactor, which comprises a pipe wall of a heat pipe, a silk screen wick, a steam cavity structure of the heat pipe, uranium-molybdenum alloy fuel, an outer pipe wall of the heat pipe, a wick attached to the inner wall surface of the heat pipe, and uranium-molybdenum alloy. The device is formed by coupling a hexagonal annular heat pipe with a hexagonal fuel element. Firstly, in the aspect of the structure, a framework structure composed of graphite is not needed to fix the heat pipe and the fuel element, an assembly composed of the heat pipe and the fuel element can be directly adopted for expansion to form the fuel element, and gap thermal resistance brought by the framework is eliminated. In addition, in terms of heat transfer performance, the heat transfer performance of the hexagonal annular heat pipe is superior to that of a traditional cylindrical heat pipe. The same heat transfer area is adopted, and under the condition of the same heat transfer power, the temperature of the outer surface of the pipe wall of the hexagonal annular heat pipe is lower than that of a cylindrical heat pipe. Therefore, the material stress requirement in the heat pipe pile and the thermoelectric conversion form of the condensation section can be richer.

A heat pipe heat exchanger device for use with cooling water in nuclear power plants is provided. [Solution] The heat pipe heat exchanger device includes a treatment device, an anti-clogging assembly, a drive assembly, and an ash cleaning assembly. The anti-clogging assembly includes a filter screen installed inside a housing included in the treatment device, a groove on the outer wall of the filter screen, a partition plate installed inside the housing, and an ash drop groove opening through the outer wall of the partition plate. The ash cleaning assembly includes a storage chamber installed inside the housing, a guide rail on the inner wall of the storage chamber, an ash collection box movably installed inside the storage chamber, and a guide groove on the outer wall of the ash collection box. When the motor is started, the moving column moves up and down to push the filter screen and shake off impurities on the filter screen. As the mounting plate moves, the cleaning brush completes the cleaning work, and as the mounting plate moves, a moving door opens to make it easier for operators to remove dust.

An integrated energy system including a power plant is discussed herein. In some examples, the integrated energy system may include at least one nuclear reactor and electrical power generation system configured to generate steam and electricity, a water treatment plant configured to produce Sodium Hydroxide (NaOH) from salt water, a Sodium Formate (HCOONa) production plant configured to receive the Sodium Hydroxide (NaOH) to produce Sodium Formate (HCOONa), a Thermal Decomposition reactor configured to receive the Sodium Formate (HCOONa) and configured to receive at least a first portion of the steam or at least a second portion of the electricity from the power plant to indirectly heat the Thermal Decomposition reactor to produce Hydrogen (H2), Carbon Dioxide (CO2), and Carbon Monoxide (CO) from the Sodium Formate (HCOONa), and a Methanol (CH3OH) reaction chamber configured to receive the Hydrogen (H2), the Carbon Dioxide (CO2), and the Carbon Monoxide (CO) to produce Methanol (CH3OH).

This invention discloses an in-core component with highly uniform core outlet flow rate and its design method, relating to the field of nuclear reactor design technology. It solves the problem of poor uniformity in coolant flow distribution at the core outlet in existing reactors, balances flow resistance at various points at the core outlet, and reduces lateral flow between different channels. The specific solution is as follows: it includes a core top plate and a support column assembly mounted above the core top plate. The core top plate is provided with several water flow holes and mounting holes. The water flow holes consist of circular and square water flow holes. The circular water flow holes are divided into first and second circular water flow holes with different functions. The area ratio of the square water flow holes to the circular water flow holes is 1.1:1 to 1.3:1.

The invention discloses an assembly method of a heat pipe type reactor and the heat pipe type reactor, and belongs to the field of nuclear power. The assembly method of the heat pipe type reactor comprises the following steps: closing one end of each mounting hole of the grid base body, vertically arranging the grid base body to enable the closed end to be located at the bottom, injecting filling slurry containing heat-conducting powder and a dispersing agent into the mounting holes, then inserting pipe fittings into the mounting holes until the filling slurry overflows, and finally assembling the grid base body into the mounting holes. The gap between the pipe fitting and the mounting hole is uniformly filled with the filling slurry in a forced flowing mode; and heating to cure and shape the filling slurry. The method can effectively eliminate the assembly gap between the grid substrate and the pipe fitting of the heat pipe type reactor, and improves the heat transfer efficiency and the safety margin of the reactor.

The invention discloses a passive reactor core and containment cooling and heat extraction system which comprises a steam generator, an air cooling heat exchanger, an exhaust unit and a water cooling heat exchanger, the primary side of the steam generator is communicated with the reactor pressure vessel; a secondary side outlet of the steam generator communicates with a heat source channel inlet of the air-cooling heat exchanger through a first pipeline; a heat source channel outlet of the air-cooling heat exchanger communicates with a heat source channel inlet of the water-cooling heat exchanger; a heat source channel of the air-cooling heat exchanger communicates with an inlet of the exhaust unit; a heat source channel outlet of the water-cooling heat exchanger is communicated with a secondary side inlet of the steam generator through a second pipeline; the interior of the containment is communicated with the first pipeline through a third pipeline; and the interior of the containment is communicated with the second pipeline through a fourth pipeline. The invention also discloses a method for cooling and heat removal of the reactor core and the containment. The system has the beneficial effects that the reactor core heat removal and the containment cooling are integrated, so that the number of system equipment is greatly reduced, and the occupation of total resources is reduced.

A core fuel element (10) and a heat pipe cooled reactor having radial extraction heat pipes (2). The core fuel element (10) is used for forming a core fuel (1), and the core fuel (1) is used for forming a heat pipe cooled reactor; the core fuel element (10) is divided into a core fuel element segments (11) along the axis; an axial heat pipe channel is formed in each core fuel element segment (11); and a radial heat pipe extraction channel is further formed in the radial direction in each core fuel element segment (11) other than the core fuel element segments (11) located at the head and the tail. The heat pipes (2) are accommodated in both the axial heat pipe channel and the radial heat pipe extraction channel. Straight heat pipes (21) and bent heat pipes (22) are arranged in cooperation with the axial heat pipe channels and the radial heat pipe extraction channels, so that a plurality of heat pipes (2) are arranged in each axial heat pipe channel in the core fuel element (10), and thus the heat exchange effect of a single core fuel element (10) is improved, thereby improving the overall heat exchange efficiency of the heat pipe cooled reactor.

The invention provides a reactor core structure capable of improving an assembly clearance and a manufacturing method thereof. The reactor core structure comprises a reactor core heat management element and a filling material; wherein the base body is provided with a plurality of assembly holes; the reactor core heat management element comprises a fuel rod or a heat pipe; the reactor core heat management element is assembled in the assembly hole of the base body; the filling material is filled in an assembly gap between the reactor core heat management element and the assembly hole; the filling material is a foam porous material, and the pore diameter of the foam porous material is micron-sized. The reactor core structure can improve the safety margin of the heat pipe reactor.

A nuclear engine system includes: a pump configured to pump a propellant; a fuel element including a set of nuclear fuel particles; a moderator configured to surround the fuel element and defining a set of moderator coolant channels configured to cool the moderator; a reflector including a neutron-reflecting material and a reflector coolant channel arranged within the reflector to cool the reflector, the reflector configured to at a first time, operate in a closed configuration to reflect neutrons to the fuel element to increase an energy flux, and at a second time, operate in an open configured to leak neutrons out of the engine system to decrease the energy flux; and a thrust nozzle configured to outlet propellant from the fuel element to produce thrust including a nozzle coolant channel arranged within a wall of the thrust nozzle configured to cool the nozzle.

There is provided a method of dismantling a dousing tank of a heavy water reactor structure. The method of dismantling a dousing tank of a heavy water reactor structure includes: selecting a dousing tank of a heavy water reactor structure for dismantlement; dismantling a 1st structure inside the dousing tank; fixing the dousing tank to a dome structure portion provided above the dousing tank through a mounting port; positioning a cutting unit on a 2nd structure including a concrete structure inside the dousing tank; and performing, by the cutting unit, cutting based on a forward or reverse rotation manner on an inner peripheral surface of the dousing tank.

The invention discloses a pressurized water reactor primary loop main pump flow prediction and control method and system. The method comprises the steps that an original main pump flow signal of a pressurized water reactor primary loop is obtained; performing multi-scale decomposition on the main pump flow signal by using an HOFEM module of an NL-FCNN to obtain multi-size main pump flow; the multi-scale main pump flow component is input into an ANSA dynamic optimization network structure for feature extraction, and a multi-scale feature vector is obtained; on the basis of the extracted multi-scale feature vector, NL-FCNN is used for predicting a main pump flow signal, and an optimal main pump flow predicted value is obtained; based on a D-ADC algorithm, the optimal main pump flow predicted value is converted into an electric signal through a digital-to-analog conversion circuit, then the electric signal is input into a main pump frequency converter to drive a main pump to operate in a frequency conversion mode, the main pump flow of the pressurized water reactor primary loop is controlled, and the method and system can accurately control the main pump flow of the pressurized water reactor primary loop.

A nuclear reactor core includes a plurality of fuel elements and moderator blocks to form a nuclear reactor core inner portion and a nuclear reactor core outer portion. The nuclear reactor core inner portion includes an inner moderator matrix formed of a plurality of inner holes that include a plurality of inner fuel openings with one or more fuel elements disposed therein. The plurality of inner holes further include a plurality of inner coolant passages to flow a coolant. The nuclear reactor core outer portion includes an outer moderator matrix formed of a plurality of outer holes that include a plurality of outer fuel openings with one or more fuel elements disposed therein. The plurality of outer holes further include a plurality of outer coolant passages to flow the coolant. The outer coolant passage diameter exceeds the inner coolant passage diameter.

A nuclear power plant containment cooling system disclosed by the present invention comprises a storage box and a phase change material, the phase change material is a material with a melting point higher than the normal working condition temperature of a containment and lower than the accident working condition temperature of the containment, the phase change material is accommodated in the storage box, and a heat exchange structure is arranged on at least one side surface of the storage box. The heat exchange structure is exposed in the environment in the containment so as to transfer heat to the phase change material in the storage box under the working condition of a containment accident, so that the phase change material absorbs heat in a phase change manner. The nuclear power plant containment cooling system provided by the invention is quick and efficient in response, and can improve the accident mitigation ability and ensure the safety. The invention also provides the nuclear power plant containment.

The invention belongs to the technical field of nuclear power chemistry, and particularly relates to a novel chemical additive process for a high-temperature gas cooled reactor cabin cooling system. Lithium hydroxide 7LiOH. H2O and hydrazine hydrate are added into the primary loop cabin cooling system through a nuclear island dosing device, and the concentration of lithium hydroxide 7LiOH. H2O and the concentration of hydrazine hydrate in the primary loop cabin cooling system are controlled in a coordinated manner, so that the requirement that the pH value of cooling water of the primary loop cabin cooling system is controlled to be 8.0-9.0 is met. The characteristics of lithium hydroxide 7LiOH. H2O are fully utilized to improve the pH value of cooling water of a primary loop cabin cooling system, hydrazine hydrate is added to maintain the reducing environment in the system, reduce the corrosion rate of low-carbon steel, reduce the radioactive dose in the primary loop cabin cooling water, reduce the generation amount of radioactive wastewater, reduce the irradiation dose of personnel and improve the operation safety of a nuclear power plant.

The invention belongs to the technical field of reactor thermal hydraulic tests, and particularly relates to a device for measuring steam-water direct contact condensation. The device comprises an upper cylinder body, a lower cylinder body, an upper sealing head and a lower sealing head, the upper sealing head, the upper cylinder body, the lower cylinder body and the lower sealing head are connected from top to bottom to form a sealed cylinder body structure; a visual mirror is arranged on the lower cylinder body, and the axial height of the visual mirror covering the lower cylinder body is more than 1 / 2; an electrode wire measuring structure is arranged at the joint of the upper cylinder and the lower cylinder and is used for capturing the vapor-liquid interface contact area; a through hole is formed in the upper sealing head, a steam outlet of the steam inlet connecting pipe is connected into the barrel structure through the through hole, and the steam outlet is connected with a steam diffuser. The vapor-liquid interface contact area can be obtained directly through visual shooting and later graphic processing, the vapor-liquid effective contact area can be quantitatively measured through the electrode wire, and therefore the vapor-water direct contact condensation heat transmission intensity is quantified.