Method and device for measuring critical heat flux of external cooling of pressure vessel

Abstract

A method for measuring the critical heat flux of the outer wall of a pressure vessel is provided. The method is designed by modeling with reference to the prototype ERVC flow channel of an actual nuclear power plant, and a full-height pressure vessel external cooling outer wall critical heat flux test device is established; the reference heat flux distribution of the reactor molten pool along the lower head at each angle is calculated according to the model conditions, and then the heating heat flux distribution and the normalized heat flux distribution factor at a specific operating angle are obtained by power shaping; the heating power is gradually increased in a manner that the normalized heat flux distribution factor is maintained unchanged until a critical state is reached, and then the CHF value and uncertainty of each azimuth of the outer wall surface of the pressure vessel external cooling are obtained. The present invention is suitable for measuring the critical heat flux at each inclination position of the outer wall surface of the lower head of the pressure vessel within the range of 0-98° when the pressure vessel is externally cooled.

Description

Technical Field

[0001] The invention relates to a technology in the field of nuclear engineering, in particular to a method and a device for measuring the boundary heat flux of the outer wall of a lower head when a reactor pressure vessel is externally cooled. Background Art

[0002] When an accident occurs in a pressurized water reactor nuclear power plant, the reactor has been shut down, but the residual fission heat and decay heat will continue to be generated in the core for a long time. The heat needs to be removed to ensure the integrity of the lower head of the pressure vessel. Water injection into the reactor cavity is an effective accident mitigation measure to achieve external cooling of the pressure vessel (ERVC). The heat flux density at each angle of the lower head of the pressure vessel is less than the critical heat flux (CHF) at the corresponding position of the outer wall, which is the key to ensuring the effectiveness of the external cooling measures of the pressure vessel. How to measure the CHF at each angle of the outer wall of the lower head of the pressure vessel is an urgent problem to be solved. However, the working environment of the reactor pressure vessel in the actual nuclear power plant is complex, and it is difficult to directly measure the critical heat flux. At present, simulation tests are used for measurement at home and abroad. The critical heat flux of the lower head of the reactor pressure vessel has technical difficulties such as thick wall non-uniform heating, curved flow channel, flow boiling, many engineering influencing factors, and complex mechanism. The existing CHF measurement method and prediction formula have some shortcomings.

[0003] The existing technologies include prediction formulas, which generally only consider the influence of a single or two factors, resulting in large prediction errors. In actual engineering design, experimental measurement data is still used as the basis, and the prediction object is not CHF in IVR-ERVC; or CHF is estimated by using equilibrium heat loss, which is suitable for approximate measurement of CHF when the test body is uniformly heated, but cannot be used for CHF measurement under ERVC conditions; or only a CHF correction method combining heat leakage measurement and numerical simulation under non-uniform heating is given, and the CHF measurement method and device under ERVC conditions are not involved. Summary of the invention

[0004] In view of the above problems existing in the prior art, the present invention proposes a method and device for measuring the critical heat flux of the outer wall of a pressure vessel when the pressure vessel is externally cooled, which is suitable for measuring the critical heat flux at various inclination positions of the outer wall surface of the lower head of the pressure vessel within the range of 0-98° when the pressure vessel is externally cooled.

[0005] The present invention is achieved through the following technical solutions:

[0006] The present invention relates to a method for measuring the critical heat flux of external cooling of a pressure vessel. A model design is performed with reference to the prototype ERVC flow channel of an actual nuclear power plant, and a full-height pressure vessel external cooling outer wall face critical heat flux test device is established; according to the model conditions, the reference heat flux distribution q of the reactor inner molten pool along the lower head at each angle θ is calculated. p (θ), and then the specific working angle θ is obtained through power shapingm Heating heat flux distribution q e (θ) and the normalized heat flux distribution factor; the heating power is gradually increased in a way that the normalized heat flux distribution factor is maintained unchanged until the critical state is reached, and then the CHF values ​​and their uncertainties of the various azimuths of the external cooling outer wall of the pressure vessel are obtained.

[0007] The modular design is a modular design with a height ratio of 1:1 and a cross-sectional area ratio of 1:100.

[0008] The model conditions include: reactor prototype geometry, core material types and proportions, and accident sequences.

[0009] The reference heat flux distribution is the input condition for subsequent heating power shaping.

[0010] The power shaping is performed by selecting the test operating angle θ m (0-98°), the heating heat flux distribution q is calculated according to the following method e (θ), specifically:

[0011] i) When 0≤θ m When the temperature is less than 74.5°,

[0012]

[0013] ii) When 74.5°≤θ m <88°,

[0014]

[0015] iii) When 88°≤θ m <98°,

[0016]

[0017] Where: θ m is the position angle of the critical heat flux measurement point; q p (θ) is the reference heat flux distribution of the prototype.

[0018] The normalized heat flux distribution factor is obtained by comparing the reference heat flux distribution q of the prototype e (θ) is normalized to determine the heating power distribution factor of each heating rod group.

[0019] The stepwise increase in heating power means: determining the input power shape according to the normalized heat flux distribution factor, and gradually increasing the heating power in proportion to the shape distribution until the measuring point θ=θ is reached. m Critical heat flux.

[0020] The gradual increase in heating power is achieved by heating in groups of heating power rods, specifically: the more groups there are, the closer the test heating heat flux distribution is to the actual heat flux distribution on the outer wall of the lower head of the reactor pressure vessel, and the complexity of the test heating power control and measurement system is correspondingly increased; a total of 33 groups are arranged in this device, corresponding to 0-98° of the lower head of the pressure vessel, one group for about 3° area; the heating power control systems of all heating rod groups are designed to be in synchronous control mode, to ensure that the power is increased synchronously according to the above-mentioned heat flux distribution during the test and the heat flux distribution shape factor is kept unchanged until the CHF value is reached.

[0021] The step length of the gradual increase in heating power is determined according to the control accuracy of the heating power and the overall heating time. Generally, the step length can be set higher in the low power stage, and the step length should be reduced when approaching the boiling crisis to reduce the uncertainty of CHF measurement. m The heating rod group is used as a reference. When the power is low, the step length is 100kW / m. 2 When the boiling crisis is approaching, the step size is reduced to 20kW / m 2 .

[0022] The critical state, according to specific identification characteristics, includes: the heating power does not change, but the wall temperature at this position continues to increase, indicating that heat transfer deteriorates and reaches the CHF state; the order of the sudden increase in wall temperature extends from the outer wall to the inside; the typical characteristic of the temperature rise of thick-wall CHF is that the heating rate is not fast but continues to rise, and the heating rate is about 0.5-1.5℃ / s.

[0023] Technical Effects

[0024] The present invention calculates the heating power distribution on the arc-shaped metal heating thick wall according to the heat flux distribution of the reactor molten pool and the power shaping method, which can accurately simulate the heat flux distribution in the actual reactor engineering environment, and the CHF value obtained by experiment is closer to the engineering condition; the system proposes a CHF measurement and identification and uncertainty analysis method for the outer wall of the head under the external cooling condition of the pressure vessel, and the CHF values ​​of different azimuths in the range of 0 to 98° can be obtained through experiments. BRIEF DESCRIPTION OF THE DRAWINGS

[0025] Figure 1 The principle diagram of the method for measuring the critical heat flux density on the outer wall of the head under external cooling of the pressure vessel;

[0026] Figure 2 It is a schematic diagram of the test body structure;

[0027] Figure 3 Schematic diagram of heating a thick wall for curved metal;

[0028] Figure 4This is a schematic diagram of the temperature rising when the boiling crisis occurs;

[0029] In the figure: arc-shaped metal heating thick wall 1, cooling water flow channel 2, visualization window 3, metal insulation layer 4, heater 5, multi-layer temperature measuring point 6. DETAILED DESCRIPTION

[0030] like Figure 1 As shown, this embodiment relates to a full-height pressure vessel external cooling outer wall facing boundary heat flux test device, including: a host computer, a heating power control cabinet, a thermal parameter and power measurement module, a test body and a high-speed camera and system as an auxiliary criterion for CHF occurrence identification, wherein: the host computer calculates and transmits the heating power signal to the heating power control cabinet after power input, the heating power control cabinet controls the heating power distribution at various azimuths on the test body, the thermal parameter and power measurement module collects the wall temperature, flow channel temperature and power signal on the test body and outputs them to the host computer, and the host computer performs CHF identification according to the input signal.

[0031] The power input calculation is to calculate the reference heat flux distribution q of the reactor pool along the lower head at each angle θ according to the geometric dimensions of the reactor prototype, the type and proportion of the core material, the accident sequence and other conditions. p (θ); according to the actual reactor geometry and the reference heat flux distribution q p (θ) Calculate the heating heat flux distribution q e (θ); according to the predetermined test angle, heating power grouping and heating heat flux distribution q e (θ) The normalized heat flux distribution factor of this test is calculated; the heating power signals of each group are calculated according to the normalized heat flux distribution factor.

[0032] The CHF identification is to calculate the actual heating heat flow q of each group according to the feedback signal a (θ); As the heating power gradually increases, when the temperature at the target azimuth position rises, it indicates that the heat transfer at that position deteriorates and reaches the CHF state.

[0033] Preferably, when the CHF state is reached, the host computer gives a partial power cut-off instruction to the heating power control cabinet, and the heating power control cabinet executes the partial power cut-off instruction, and the CHF phenomenon disappears.

[0034] Preferably, the host computer further calculates the actual critical heat flux and uncertainty when CHF occurs based on the test data, specifically:

[0035] The uncertainty when CHF occurs is: Δq = Δq 1 +Δq 2 , where Δq 1is the heat flux uncertainty calculated from the last power step before CHF occurs; Δq 2 is the uncertainty of heat flow calculated based on the uncertainty of the measuring instrument; the sum of the two is the uncertainty of this CHF measurement.

[0036] like Figure 2 As shown, the test body is provided with an arc-shaped metal heating thick wall 1, and the metal heating wall 1 and the metal insulation layer 4 are connected by a metal enclosure to form a cooling water channel 2 with a rectangular cross-section, wherein: the fluid flows from bottom to top or from top to bottom, and the heat conduction direction of the heated heat flow is from inside to outside, and the heat is taken away by the coolant in the cooling water channel 2 to achieve outer wall cooling.

[0037] The metal enclosure is provided with a plurality of visualization windows 3 for high-speed camera shooting. The high-speed camera transmits the bubble dynamic information on the arc-shaped metal heating thick wall 1 to the host computer to assist CHF identification.

[0038] The auxiliary CHF identification refers to: observing the dynamic characteristics of bubbles such as bubble generation, aggregation, sliding, detachment, and aggregation on the heating wall through the visualization windows on the front and rear panels on both sides of the flow channel, and observing whether the wall surface of the arc-shaped metal heating thick wall 1 is completely covered by bubbles, resulting in deterioration of heat transfer and boiling crisis.

[0039] like Figure 3 As shown, an array heater 5 is arranged inside the arc-shaped metal heating thick wall 1, and the array heater 5 grouped according to the azimuth position is controlled by a heating power control cabinet to realize heating power distribution control.

[0040] The arc-shaped metal heating thick wall 1 is provided with multiple layers of temperature measuring points 6 along the normal direction of the arc to monitor the wall temperature change of the arc-shaped metal heating thick wall 1.

[0041] This embodiment relates to a CHF measurement method based on the above device, including:

[0042] Step 1) Calculate the reference heat flux distribution q according to the actual engineering conditions of the target reactor p (θ), according to the reference heat flux distribution q p (θ) and the reactor geometry are used to perform power shaping to obtain the heat flux distribution q of the critical heat flux test. e (θ);

[0043] The purpose of the power shaping is to obtain upstream conditions and downstream gas phase superficial velocity that are more consistent with those in the actual reactor cavity prototype. The CHF obtained by experimental measurement is closer to the actual nuclear power plant prototype.

[0044] Step 2) determining the normalized heat flux distribution factor of this test according to the target test condition azimuth angle θ and the grouping of the test body heating rods; during the test, gradually increasing the test body heating power according to the normalized heat flux distribution factor, and keeping the heat flux distribution factor unchanged during the heating process;

[0045] Step 3) As the heating power gradually increases, when a temperature spike occurs at the target azimuth position, it indicates that the heat transfer at that position has deteriorated and CHF has occurred. Specifically, the heating power has not changed, but the wall temperature at that position continues to increase, indicating that heat transfer has deteriorated and CHF has occurred; the order of the sudden increase in wall temperature is from the outer wall to the inside; the typical feature of the temperature spike of thick-wall CHF is that the heating rate is not fast but continues to rise, and the heating rate is about 0.5 to 1.5 ° C / s.

[0046] Step 4) After CHF occurs, the host computer gives a partial power cut-off instruction to the heating power cabinet, and the CHF phenomenon disappears. The CHF at this location is preliminarily estimated based on the various groups of heating power and wall temperature data fed back to the host computer in real time during the test, and the actual CHF and the uncertainty of this critical heat flux measurement can be further calculated based on the measurement parameters in the surrounding area and the flow channel. At this point, a critical heat flux test under external cooling conditions of a pressure vessel is completed. By changing the target azimuth angle and repeating the above process, the CHF value and uncertainty of the outer wall of the pressure vessel external cooling RPV at an azimuth angle of 0 to 98° can be obtained.

[0047] The heating power and wall temperature data of each group are preliminarily estimated using Fourier's law, the formula is: Where: q CHF1 is the preliminary estimated critical heat flux at the boiling crisis location, k is the thermal conductivity of the metal, ΔT is the temperature difference between the wall temperature measuring points at the same azimuth when the temperature field is stable after the power is increased before CHF occurs, and ΔL is the distance between the corresponding wall temperature measuring points.

[0048] The power increase step is determined according to the control accuracy of the heating power and the overall heating time. Generally, the step can be set higher in the low power stage, and the step should be reduced when approaching CHF to reduce the uncertainty of CHF measurement. m The heating rod group is used as a reference. When the power is low, the step length is 100kW / m. 2 When the boiling crisis is approaching, the step size is reduced to 20kW / m 2 .

[0049] The actual CHF is specifically: Where: q CHF2 is the actual critical heat flux at the boiling crisis position, P is the heating power of the rod group corresponding to the CHF position, A is the effective heating area on the heating wall corresponding to the rod group at the CHF position, and C is the correction factor calculated based on heat loss.

[0050] This method deduces the reference heat flux distribution and the heat flux distribution loaded after power shaping from the reactor prototype, and uses the host computer and the test body heating power control system to gradually increase the power under the condition of unchanged normalized heat flux distribution factor until CHF occurs; the CHF values ​​and their uncertainties at various azimuths are calculated based on the power and thermal parameters fed back by the test body.

[0051] After specific actual experiments, based on the above method and device, by selecting different target azimuth angles θ, the critical heat flux curve CHF(θ) covering the full angle of the outer wall of the lower head of the pressure vessel (θ is 0-98°) can be obtained; assisted by other auxiliary and supporting systems to change the test conditions (flow rate, pressure, inlet subcooling, outer wall surface characteristics, etc.), based on the test method and device proposed in this patent, the critical heat flux curve CHF(θ) under the corresponding sensitivity test conditions can be obtained.

[0052] Compared with the existing technology, this method solves the problem of difficulty in directly measuring the critical heat flux on the outer wall of the lower head of the nuclear power plant pressure vessel, can truly simulate the heat flux distribution in the nuclear power plant prototype, and realize the critical heat flux measurement device and method of the outer wall of the pressure vessel at all angles. The measured critical heat flux curve CHF(θ) is closer to the prototype. The critical heat flux data obtained by this device and method can be directly used for the safety review and optimization design of the IVR strategy of the nuclear power plant.

[0053] The above-mentioned specific implementation can be partially adjusted in different ways by those skilled in the art without departing from the principle and purpose of the present invention. The protection scope of the present invention shall be based on the claims and shall not be limited by the above-mentioned specific implementation. Each implementation scheme within its scope shall be subject to the constraints of the present invention.

Claims

1. A method for measuring the boundary heat flux of the outer wall of a pressure vessel, It is characterized in that The simulation design is carried out with reference to the prototype ERVC flow channel of the actual nuclear power plant, and a full-height pressure vessel external cooling outer wall boundary heat flux test device is established; the reference heat flux distribution q of the reactor molten pool along the lower head at each angle θ is calculated according to the model conditions. p (θ), and then the specific working angle θ is obtained through power shaping m Heating heat flux distribution q e (θ) and normalized heat flux distribution factor; gradually increase the heating power in a manner that maintains the normalized heat flux distribution factor unchanged until a critical state is reached, thereby obtaining the CHF value and uncertainty of each azimuth angle of the external cooling outer wall of the pressure vessel; The power shaping is performed by selecting the test operating angle θ in the range of 0-98°. m , the heating heat flux distribution q is calculated according to the following method e (θ), specifically: i) When 0≤θ m When the temperature is less than 74.5°, ii) When 74.5°≤θ m <88°, iii) When 88°≤θ m <98°, Where: θ m is the position angle of the critical heat flux measurement point; q p (θ) is the reference heat flux distribution of the prototype; The normalized heat flux distribution factor is obtained by comparing the reference heat flux distribution q of the prototype e (θ) is normalized to determine the heating power distribution factor of each heating rod group; The stepwise increase in heating power means: determining the input power shape according to the normalized heat flux distribution factor, and gradually increasing the heating power in proportion to the shape distribution until the measuring point θ=θ is reached. m At critical heat flux; The stepwise increase in heating power is achieved by heating with heating power rods in groups, specifically: the more groups there are, the closer the test heating heat flux distribution is to the actual heat flux distribution on the outer wall of the lower head of the reactor pressure vessel, and the complexity of the test heating power control and measurement system is increased accordingly; a total of 33 groups are arranged in this device, corresponding to one group of 0-98° and 3° areas of the lower head of the pressure vessel; the heating power control systems of all heating rod groups are designed to be in synchronous control mode, ensuring that the power is increased synchronously according to the above heat flux distribution during the test and the heat flux distribution shape factor is kept unchanged until the CHF value is reached; The step size of the stepwise increase in heating power is determined according to the control accuracy of the heating power and the overall heating time.

2. The method for measuring the critical heat flux of the outer wall of a pressure vessel according to claim 1, Its characteristics are: The modular design is a modular design with a height ratio of 1:1 and a cross-sectional area ratio of 1:

100.

3. The method for measuring the critical heat flux of the outer wall of a pressure vessel according to claim 1, Its characteristics are: The model conditions include: reactor prototype geometry, core material types and proportions, and accident sequences.

4. The method for measuring the critical heat flux of the outer wall of a pressure vessel according to claim 1, Its characteristics are: The reference heat flux distribution is the input condition for subsequent heating power shaping.

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