CN109613429B - Pressurized water reactor steam generator model time constant testing system and method - Google Patents
Pressurized water reactor steam generator model time constant testing system and method Download PDFInfo
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- CN109613429B CN109613429B CN201811544354.7A CN201811544354A CN109613429B CN 109613429 B CN109613429 B CN 109613429B CN 201811544354 A CN201811544354 A CN 201811544354A CN 109613429 B CN109613429 B CN 109613429B
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Abstract
The invention relates to the field of power grid stability simulation modeling analysis, in particular to a pressurized water reactor steam generator model time constant testing system and a method. The method can effectively obtain the time constant of the pressurized water reactor steam generator model, can guide the regulation rhythm of the nuclear power turbine on the power of the generator set, and prevents the phenomena of over regulation, insufficient regulation quantity, unstable regulation process and the like of the power of the generator.
Description
Technical Field
The invention relates to the field of power grid stability simulation modeling analysis, in particular to a system and a method for testing a time constant of a pressurized water reactor steam generator model.
Background
A steam generator of a nuclear power unit is a bridge for transferring heat of a primary loop to a secondary loop. The steam generator model consists of a primary loop coolant average temperature model, a U-shaped metal pipe heat transfer model and a secondary loop steam pressure model. The time constants of the steam generator model reflect the response condition of the energy conversion of the reactor into the primary loop steam energy, the time constants are accurately obtained, and the method has an important effect on the power regulator section of the generator set. However, no system and method for testing the time constant of the steam generator model is reported at present.
The nuclear power generating set has large capacity and low temperature and pressure of parameters of the main steam of the two loops, the reactor energy is transmitted to the main steam of the two loops through the steam generator, the time constant of a steam generator model is accurately obtained, the regulation rhythm of the nuclear power generating set power by the nuclear power turbine can be guided, and the phenomena of overshoot, insufficient regulation quantity, unstable regulation process and the like of the power of the generator are prevented. The steam generator model time constant test improves the accuracy of power grid stability simulation calculation on one hand, and can be used as a reference basis for a power regulation mode of a generator of a nuclear power unit on the other hand.
Disclosure of Invention
In order to solve the problems, the invention provides a system and a method for testing a time constant of a pressurized water reactor steam generator model, and the specific technical scheme is as follows:
a pressurized water reactor steam generator model time constant test system is characterized in that a steam generator model is formed by sequentially connecting a primary loop coolant average temperature module, a U-shaped metal tube heat transfer module and a secondary loop steam pressure module and comprises a primary loop coolant inlet temperature measuring device of a pressurized water reactor steam generator, a primary loop coolant outlet temperature measuring device of the pressurized water reactor steam generator, a steam turbine valve opening degree measuring device, a generator power measuring device, a steam generator outlet secondary loop steam temperature measuring device, a steam generator outlet secondary loop steam pressure measuring device and a data acquisition device;
the device for measuring the temperature of the coolant inlet of the primary loop of the pressurized water reactor steam generator is used for measuring the temperature of the coolant inlet of the primary loop of the pressurized water reactor steam generator and is arranged on a coolant inlet pipeline of the steam generator;
the device for measuring the temperature of the coolant outlet of the primary loop of the pressurized water reactor steam generator is used for measuring the temperature of the coolant outlet of the primary loop of the pressurized water reactor steam generator and is arranged on a coolant outlet pipeline of the steam generator;
the steam turbine regulating valve opening degree measuring device is used for measuring the steam turbine regulating valve opening degree and is arranged on a steam turbine regulating valve servomotor valve rod;
the generator power measuring device is used for measuring the power of the generator;
the steam temperature measuring device of the second loop of the outlet of the steam generator is used for measuring the temperature of the steam of the second loop of the outlet of the steam generator and is arranged on a steam outlet pipeline of the second loop of the steam generator;
the steam generator outlet secondary loop steam pressure measuring device is used for measuring the pressure of steam at the outlet of the secondary loop of the steam generator and is arranged on an outlet pipeline of the secondary loop of the steam generator;
the data acquisition device is used for acquiring measurement data of a primary loop coolant inlet temperature measurement device of the pressurized water reactor steam generator, a primary loop coolant outlet temperature measurement device of the pressurized water reactor steam generator, a steam turbine valve opening measurement device, a generator power measurement device, a steam generator outlet secondary loop steam temperature measurement device and a steam generator outlet secondary loop steam pressure measurement device; the data acquisition device is respectively connected with a primary loop coolant inlet temperature measuring device of the pressurized water reactor steam generator, a primary loop coolant outlet temperature measuring device of the pressurized water reactor steam generator, a steam turbine valve opening degree measuring device, a generator power measuring device, a steam generator outlet secondary loop steam temperature measuring device and a steam generator outlet secondary loop steam pressure measuring device.
Preferably, the primary loop coolant inlet temperature measuring device of the pressurized water reactor steam generator and the primary loop coolant outlet temperature measuring device of the pressurized water reactor steam generator are respectively measured by a narrow-range direct immersion resistance thermometer; the narrow-range direct immersion resistance thermometer is vertically arranged on the corresponding pipeline.
Preferably, the steam generator outlet two-loop steam temperature measuring device comprises a temperature sleeve and a temperature measuring element respectively; the temperature sleeve is vertically arranged on the corresponding pipeline, the temperature measuring element is arranged in the temperature sleeve, and the temperature measuring element adopts an E-type thermocouple.
Preferably, the steam turbine throttle opening measuring device measures the steam turbine throttle opening using a displacement sensor.
Preferably, the generator power measuring device comprises a current transformer, a voltage transformer and a power transmitter; the current transformer is used for converting the current output by the stator of the generator and then transmitting the converted current to the power transmitter; the voltage transformer is used for converting the voltage output by the stator of the generator and then transmitting the voltage to the power transmitter; and the power transmitter is used for calculating the power of the generator according to the current and voltage data transmitted by the current transformer and the voltage transformer.
Preferably, the steam generator outlet secondary loop steam pressure measuring device comprises a pressure sampling tube and a pressure transmitter; the pressure sampling tube is arranged on an outlet pipeline of the secondary loop of the steam generator and is used for transmitting a pressure signal of steam at the outlet of the secondary loop of the steam generator to the pressure transmitter; the pressure transmitter is used for converting a pressure signal of steam at the outlet of the second loop of the steam generator into a measurable direct current signal of-5- +5V or + 1- + 5V.
Preferably, the data acquisition device adopts an EIC-02L type data acquisition system.
A test method of a pressurized water reactor steam generator model time constant test system comprises the following steps:
(1) the system comprises a reactor, a measuring system, a debugging instrument and a steam generator, wherein the measuring system is installed and the debugging instrument is normal when the reactor is empty and the unit is in a shutdown state, and the measuring system specifically comprises a primary loop coolant inlet temperature measuring device of a pressurized water reactor steam generator, a primary loop coolant outlet temperature measuring device of the pressurized water reactor steam generator, a steam turbine opening regulating measuring device, a generator power measuring device, a secondary loop steam temperature measuring device of a steam generator outlet, and a secondary loop steam pressure measuring device of a steam generator outlet, wherein the primary loop coolant inlet temperature measuring device is installed on a coolant inlet pipeline of the steam generator;
(2) after the reactor is filled, the nuclear power unit is started and is loaded, the high-pressure regulating valve of the steam turbine is in a stable state, a step instruction is given to the high-pressure regulating valve, the opening of the high-pressure regulating valve is suddenly changed, the power of the power unit is also suddenly changed, and the time of 100 milliseconds is kept after the sudden change of the opening of the high-pressure regulating valve; in the test process, the strokes of the R rod and the G rod are kept unchanged, and the thermal power of the reactor is also unchanged;
(3) starting a data acquisition device 200 milliseconds before giving a step instruction to a high-pressure regulating valve, and recording the inlet temperature of a primary loop coolant of a pressurized water reactor steam generator, the outlet temperature of the primary loop coolant of the pressurized water reactor steam generator, the opening degree of a steam turbine regulating valve, the power of a generator, the steam temperature of a secondary loop of a steam generator outlet and the steam pressure of a secondary loop of the steam generator outlet;
(4) the sudden change of the high-pressure regulating valve causes the sudden change of the main steam pressure of the two loops to cause the change of the inlet temperature and the outlet temperature of the reactor coolant, the inlet temperature and the outlet temperature of the steam generator coolant correspondingly change, and the simulation curve is consistent with the measured data through the least square method, so that the time constant of the steam generator model is determined.
Preferably, the step (4) further comprises the steps of establishing a steam generator model on the ADPSS software, setting the working condition the same as that of the actual nuclear power generating unit, and enabling a simulation curve of the steam generator model to be consistent with measured data of the nuclear power generating unit through a least square method, so that a time constant of the steam generator model is determined.
Preferably, the determining of the steam generator model time constant is specifically: and calculating the time constant of the steam generator model by performing Laplace transform on the mathematical expression of the steam generator model.
The invention has the beneficial effects that: by the test system and the test method, the time constant of the pressurized water reactor steam generator model can be effectively obtained, the regulation rhythm of the nuclear turbine on the power of the generator set can be guided, and the phenomena of over regulation, insufficient regulation quantity, unstable regulation process and the like of the power of the generator are prevented.
Drawings
FIG. 1 is a schematic diagram of a test system according to the present invention;
FIG. 2 is a comparative schematic of actual, simulated generator power for a steam generator model in an embodiment of the invention;
FIG. 3 is a schematic comparison of actual, simulated primary loop coolant outlet temperatures of a pressurized water reactor steam generator for a steam generator model according to an embodiment of the present invention;
FIG. 4 is a schematic comparison of actual, simulated primary loop coolant inlet temperatures of a pressurized water reactor steam generator for a steam generator model according to an embodiment of the present invention;
FIG. 5 is a schematic comparison of actual, simulated turbine damper opening for a steam generator model in an embodiment of the present invention;
FIG. 6 is a comparative schematic of actual, simulated steam generator outlet two-circuit steam pressure for a steam generator model in an embodiment of the present invention.
Detailed Description
For a better understanding of the present invention, reference is made to the following detailed description taken in conjunction with the accompanying drawings in which:
as shown in fig. 1, the steam generator model is formed by sequentially connecting a primary circuit coolant average temperature module, a U-shaped metal pipe heat transfer module and a secondary circuit steam pressure module, wherein the primary circuit coolant average temperature module comprises a nuclear reactor, a voltage stabilizer and a coolant pump; the U-shaped metal tube heat transfer module comprises a steam generator; the two-loop steam pressure module comprises a high-pressure cylinder, a reheater, a low-pressure cylinder, a condenser, a condensate pump, a low-pressure heater, a deaerator, a feed pump and a high-pressure heater; the coolant pump is respectively connected with the nuclear reactor, the voltage stabilizer and the steam generator through pipelines; the voltage stabilizer is respectively connected with the nuclear reactor and the steam generator through pipelines; the steam generator is respectively connected with the high-pressure cylinder and the high-pressure heater through pipelines; the high-pressure cylinder, the reheater and the low-pressure cylinder are mutually connected through pipelines; the low-pressure cylinder is respectively connected with the condenser and the generator. The condenser, the condensate pump, the low-pressure heater, the deaerator, the water feed pump and the high-pressure heater are sequentially connected; and a high-pressure regulating valve is arranged on a connecting pipeline of the high-pressure cylinder and the steam generator.
As shown in fig. 1, a system for testing a time constant of a model of a pressurized water reactor steam generator includes a primary loop coolant inlet temperature measuring device 1 of the pressurized water reactor steam generator, a primary loop coolant outlet temperature measuring device 2 of the pressurized water reactor steam generator, a steam turbine regulating valve opening degree measuring device 5, a generator power measuring device 6, a steam generator outlet secondary loop steam temperature measuring device 3, a steam generator outlet secondary loop steam pressure measuring device 4, and a data collecting device 10.
The primary loop coolant inlet temperature measuring device 1 of the pressurized water reactor steam generator is used for measuring the temperature of a primary loop coolant inlet of the pressurized water reactor steam generator, and the primary loop coolant inlet temperature measuring device 1 of the pressurized water reactor steam generator adopts a narrow-range direct immersion resistance thermometer for measurement; the narrow-range direct immersion resistance thermometer is vertically arranged on the corresponding pipeline. The narrow-range direct immersion resistance thermometer adopts one of an N9355S-1B thermometer and a KN615-S41CB-276S-S1 thermometer.
The primary loop coolant outlet temperature measuring device 2 of the pressurized water reactor steam generator is used for measuring the temperature of a primary loop coolant outlet of the pressurized water reactor steam generator; a primary loop coolant outlet temperature measuring device 2 of a pressurized water reactor steam generator adopts a narrow-range direct immersion resistance thermometer for measurement; the narrow-range direct immersion resistance thermometer is vertically arranged on the corresponding pipeline. The narrow-range direct immersion resistance thermometer adopts one of an N9355S-1B thermometer and a KN615-S41CB-276S-S1 thermometer.
The steam turbine opening degree measuring device 5 is used for measuring the steam turbine opening degree; the steam turbine throttle opening measuring device 5 comprises a displacement sensor, and the displacement sensor is arranged on a valve rod of a steam turbine throttle servomotor. Wherein, the range of the displacement sensor is 0-500mm, and the output voltage is-5V- +5V or + 1- + 5V.
The generator power measuring device 6 is used for measuring the power of the generator; the generator power measuring device 6 comprises a current transformer 7, a voltage transformer 8 and a power transmitter 9; the current transformer 7 is used for converting the current output by the stator of the generator and then transmitting the converted current to the power transmitter; the voltage transformer 8 is used for converting the voltage output by the stator of the generator and then transmitting the voltage to the power transmitter; the power transmitter 9 is used for calculating the power of the generator according to the current and voltage data transmitted by the current transformer 7 and the voltage transformer 8, and the power transmitter 9 is an FPWT-201 type power transmitter.
The steam temperature measuring device 3 of the second loop of the outlet of the steam generator is used for measuring the temperature of the steam of the second loop of the outlet of the steam generator; the steam temperature measuring device 3 of the steam generator outlet secondary loop comprises a temperature sleeve and a temperature measuring element; the temperature sleeve is vertically arranged on a steam outlet pipeline of a second loop of the steam generator, the temperature measuring element is arranged in the temperature sleeve, and the temperature measuring element adopts an E-type thermocouple.
The steam generator outlet secondary loop steam pressure measuring device 4 is used for measuring the pressure of steam at the outlet of the steam generator secondary loop; the steam generator outlet secondary loop steam pressure measuring device 4 comprises a pressure sampling tube and a pressure transmitter; the pressure sampling tube is arranged on an outlet pipeline of the secondary loop of the steam generator and is used for transmitting a pressure signal of steam at the outlet of the secondary loop of the steam generator to the pressure transmitter; the pressure transmitter is used for converting a pressure signal of steam at the outlet of the second loop of the steam generator into a measurable direct current signal of-5- +5V or + 1- + 5V. The pressure transmitter adopts EJA series pressure transmitter.
The data acquisition device 10 is used for acquiring measurement data of a primary loop coolant inlet temperature measurement device 1 of the pressurized water reactor steam generator, a primary loop coolant outlet temperature measurement device 2 of the pressurized water reactor steam generator, a steam turbine valve opening measurement device 5, a generator power measurement device 6, a secondary loop steam temperature measurement device 3 of the steam generator outlet and a secondary loop steam pressure measurement device 4 of the steam generator outlet, and the data acquisition device 10 adopts an EIC-02L type data acquisition system.
The data acquisition device 10 is respectively connected with a primary loop coolant inlet temperature measuring device 1 of the pressurized water reactor steam generator, a primary loop coolant outlet temperature measuring device 2 of the pressurized water reactor steam generator, a steam turbine valve opening degree measuring device 5, a generator power measuring device 6, a secondary loop steam temperature measuring device 3 of the steam generator outlet and a secondary loop steam pressure measuring device 4 of the steam generator outlet in a wired or wireless mode.
A test method of a pressurized water reactor steam generator model time constant test system comprises the following steps:
(1) the system comprises a reactor, a measuring system, a debugging instrument and a steam generator, wherein the measuring system is installed and the debugging instrument is normal when the reactor is empty and the unit is in a shutdown state, and the system specifically comprises a primary loop coolant inlet temperature measuring device 1 of a pressurized water reactor steam generator, which is installed on a coolant inlet pipeline of the steam generator, a primary loop coolant outlet temperature measuring device 2 of the pressurized water reactor steam generator, which is installed on a coolant outlet pipeline of the steam generator, a steam turbine opening regulating measuring device 3, which is installed on a valve rod of a steam turbine opening regulating oil-operated machine, a generator power measuring device 6, a secondary loop steam temperature measuring device 3 of a steam generator outlet, which is installed on a steam outlet pipeline of a secondary loop of the steam generator, and a secondary loop;
(2) after the reactor is filled, the nuclear power unit is started and is loaded, the high-pressure regulating valve of the steam turbine is in a stable state, a step instruction is given to the high-pressure regulating valve, the opening of the high-pressure regulating valve is suddenly changed, the power of the power unit is also suddenly changed, and the time of 100 milliseconds is kept after the sudden change of the opening of the high-pressure regulating valve; in the test process, the strokes of the R rod and the G rod are kept unchanged, and the thermal power of the reactor is also unchanged;
(3) starting the data acquisition device 10 200 milliseconds before giving a step instruction to the high-pressure regulating valve, and recording the inlet temperature of the primary loop coolant of the pressurized water reactor steam generator, the outlet temperature of the primary loop coolant of the pressurized water reactor steam generator, the opening degree of a steam turbine regulating valve, the power of a generator, the steam temperature of a second loop of the outlet of the steam generator and the steam pressure of a second loop of the outlet of the steam generator;
(4) the sudden change of the high-pressure regulating valve causes the sudden change of the main steam pressure of the two loops to cause the change of the inlet temperature and the outlet temperature of the reactor coolant, the inlet temperature and the outlet temperature of the steam generator coolant correspondingly change, and the simulation curve is consistent with the measured data through the least square method, so that the time constant of the steam generator model is determined.
And (4) establishing a steam generator model on ADPSS software, setting the working condition the same as that of the actual nuclear power unit, and enabling the simulation curve of the steam generator model to be consistent with the actually measured data of the nuclear power unit by a least square method so as to determine the time constant of the steam generator model. The specific determination of the steam generator model time constant is as follows: and calculating the time constant of the steam generator model by performing Laplace transform on the mathematical expression of the steam generator model.
For a certain nuclear power unit field test, the output power of the nuclear power unit is subjected to two steps at 200ms and 300ms respectively, and meanwhile, the reaction rate of a reactor in the experimental process, namely the output thermal power is kept constant. The rated power of the nuclear power unit is 1086MW, and the two steps are respectively from 100 percent of rated power to 95 percent of rated power and then back to 100 percent of rated power. The actual measurement data of the nuclear power plant is shown by solid lines in fig. 2-6, and the solid lines in fig. 2-6 respectively show the generator power, the outlet temperature of the primary loop coolant of the steam generator, the inlet temperature of the primary loop coolant of the steam generator, the opening degree of the steam turbine regulating valve and the steam pressure of the secondary loop of the outlet of the steam generator actually tested by the nuclear power plant.
A nuclear power unit model is established on the ADPSS, and a steam generator is a bridge for transferring heat of a primary loop to a secondary loop. The steam generator consists of a primary loop coolant average temperature module, a U-shaped metal pipe heat transfer module and a secondary loop steam pressure module. The coolant of the primary loop transfers heat to the secondary loop through the metal tube wall of the steam generator, and the working medium of the secondary loop is continuously heated to be changed into a steam-water mixture. The following assumptions are made for the model to be built:
(1) the fluid of the first loop and the second loop in the steam generator flows only in one direction;
(2) the density and specific heat of the fluid in the first loop and the second loop in the steam generator are constant;
(3) the thermal conductivity of the metal tube wall is constant;
(4) the thermodynamic characteristics of the saturated water and the saturated steam of the second loop are in linear relation with the pressure;
(5) the enthalpy and mass characteristics of the steam-water mixture of the two loops are in linear relation with the position of the heat conducting path;
(6) except for the primary loop coolant and the metal pipe wall, and the metal pipe wall and the secondary loop working medium, no other heat exchange is generated.
The mathematical models of the parameters of the steam generator set can be obtained as the following formulas (1) to (3):
in the formula:
△TC: steam generator primary circuit coolant outlet temperature deviation; delta TH: steam generator primary circuit coolant inlet temperature deviation; delta Tm: temperature deviation of the metal tube; delta PS: two-circuit steam pressure deviation; Δ y: deviation of opening degree of a steam turbine regulating valve; t is tSteaming food: time constant of coolant in steam generator; t is tU: a U-tube time constant; t is tPress and press: a pressure time constant; lambda [ alpha ]1: temperature coefficient of the metal tube; lambda [ alpha ]2: hot wire temperature coefficient; lambda [ alpha ]3: a temperature coefficient of coolant; lambda [ alpha ]4: a pressure temperature coefficient; lambda [ alpha ]5: the metal tube pressure coefficient; lambda [ alpha ]6: valve pressure coefficient.
And (3) performing Laplace transform on the formulas (1), (2) and (3) to obtain a transfer function of a steam generator centralized parameter mathematical model as the following formulas (4) to (6):
in the modeling process, the thermal power output by the reactor is kept constant at a rated value, so that a reactor neutron dynamic model is selectively not considered, and a thermodynamic system of coolant circulation and a steam generator model are considered.
The same working conditions are set, and the thermodynamic parameters of a nuclear power unit loop are adjusted, so that the built nuclear power unit model shows the same response characteristics as an actual nuclear power unit. The simulation curve is consistent with the measured data through a least square method, the simulation data are shown as dotted lines in fig. 2-6, and the dotted lines in fig. 2-6 respectively represent the generator power, the primary loop coolant outlet temperature of the steam generator, the primary loop coolant inlet temperature of the steam generator, the steam turbine throttle opening and the steam pressure of the secondary loop of the steam generator outlet of the nuclear power unit simulation.
The response of the power instant step of the nuclear power unit can completely express the characteristics of internal thermodynamic links of the nuclear power unit, and parameter identification can be performed on the nuclear power unit modeling data based on the ADPSS through simulation of the parameters through experiments, so that time parameters reflecting the actual unit response performance are determined. The relevant parameter settings of the mathematical model of the steam generator lumped parameters are shown in table 1 below:
TABLE 1 parameters of steam generator model settings
| λ1 | Temperature coefficient of metal tube | 0.7912 |
| λ2 | Temperature coefficient of hot wire | 0.2088 |
| λ3 | Temperature coefficient of coolant | 0.4339 |
| λ4 | Temperature coefficient of pressure | 178.3125 |
| λ5 | Pressure coefficient of metal tube | 0.003 |
| λ6 | Coefficient of pressure of valve | 0.0284 |
The time constants calculated from the experimental data are shown in table 2 below:
TABLE 2 time constants of steam generator models calculated from experimental data
| tSteaming food | Time constant(s) of coolant in steam generator | 0.002 |
| tU | Time constant(s) of U-tube | 0.00936 |
| tPress and press | Time constant of pressure(s) | 0.01 |
The present invention is not limited to the above-described embodiments, which are merely preferred embodiments of the present invention, and the present invention is not limited thereto, and any modification, equivalent replacement, and improvement made within the spirit and principle of the present invention should be included in the protection scope of the present invention.
Claims (10)
1. The utility model provides a pressurized water reactor steam generator model time constant test system, wherein, the steam generator model is connected gradually by a return circuit coolant average temperature module, U type tubular metal resonator heat transfer module and two return circuit steam pressure module triples and constitutes its characterized in that: the system comprises a primary loop coolant inlet temperature measuring device (1) of a pressurized water reactor steam generator, a primary loop coolant outlet temperature measuring device (2) of the pressurized water reactor steam generator, a steam turbine valve opening degree measuring device (5), a generator power measuring device (6), a secondary loop steam temperature measuring device (3) of a steam generator outlet, a secondary loop steam pressure measuring device (4) of the steam generator outlet and a data collecting device (10);
the device (1) for measuring the temperature of the primary loop coolant inlet of the pressurized water reactor steam generator is used for measuring the temperature of the primary loop coolant inlet of the pressurized water reactor steam generator and is arranged on a coolant inlet pipeline of the steam generator;
the device (2) for measuring the temperature of the coolant outlet of the primary loop of the pressurized water reactor steam generator is used for measuring the temperature of the coolant outlet of the primary loop of the pressurized water reactor steam generator and is arranged on a coolant outlet pipeline of the steam generator;
the steam turbine regulating valve opening degree measuring device (5) is used for measuring the steam turbine regulating valve opening degree and is arranged on a valve rod of a steam turbine regulating valve servomotor; the generator power measuring device (6) is used for measuring the power of the generator;
the steam temperature measuring device (3) of the second loop of the steam generator outlet is used for measuring the temperature of the steam of the second loop of the steam generator outlet and is arranged on a steam outlet pipeline of the second loop of the steam generator;
the steam generator outlet secondary loop steam pressure measuring device (4) is used for measuring the pressure of steam at the outlet of the secondary loop of the steam generator and is arranged on an outlet pipeline of the secondary loop of the steam generator;
the data acquisition device (10) is used for acquiring measurement data of a primary loop coolant inlet temperature measurement device (1) of the pressurized water reactor steam generator, a primary loop coolant outlet temperature measurement device (2) of the pressurized water reactor steam generator, a steam turbine regulating valve opening degree measurement device (5), a generator power measurement device (6), a steam generator outlet secondary loop steam temperature measurement device (3) and a steam generator outlet secondary loop steam pressure measurement device (4); the data acquisition device (10) is respectively connected with a primary loop coolant inlet temperature measuring device (1) of the pressurized water reactor steam generator, a primary loop coolant outlet temperature measuring device (2) of the pressurized water reactor steam generator, a steam turbine regulating valve opening degree measuring device (5), a generator power measuring device (6), a steam generator outlet secondary loop steam temperature measuring device (3) and a steam generator outlet secondary loop steam pressure measuring device (4).
2. The PWR steam generator model time constant testing system of claim 1, wherein: the primary loop coolant inlet temperature measuring device (1) of the pressurized water reactor steam generator and the primary loop coolant outlet temperature measuring device (2) of the pressurized water reactor steam generator are respectively measured by adopting narrow-range direct immersion resistance thermometers; the narrow-range direct immersion resistance thermometer is vertically arranged on the corresponding pipeline.
3. The PWR steam generator model time constant testing system of claim 1, wherein: the steam temperature measuring device (3) of the two loops at the outlet of the steam generator respectively comprises a temperature sleeve and a temperature measuring element; the temperature sleeve is vertically arranged on the corresponding pipeline, the temperature measuring element is arranged in the temperature sleeve, and the temperature measuring element adopts an E-type thermocouple.
4. The PWR steam generator model time constant testing system of claim 1, wherein: the steam turbine opening adjusting degree measuring device (5) adopts a displacement sensor to measure the steam turbine opening adjusting degree.
5. The PWR steam generator model time constant testing system of claim 1, wherein: the generator power measuring device (6) comprises a current transformer (7), a voltage transformer (8) and a power transmitter (9); the current transformer (7) is used for converting the current output by the stator of the generator and then transmitting the converted current to the power transmitter; the voltage transformer (8) is used for converting the voltage output by the stator of the generator and then transmitting the voltage to the power transmitter; and the power transmitter (9) is used for calculating the power of the generator according to the current and voltage data transmitted by the current transformer (7) and the voltage transformer (8).
6. The PWR steam generator model time constant testing system of claim 1, wherein: the steam generator outlet secondary loop steam pressure measuring device (4) comprises a pressure sampling tube and a pressure transmitter; the pressure sampling tube is arranged on an outlet pipeline of the secondary loop of the steam generator and is used for transmitting a pressure signal of steam at the outlet of the secondary loop of the steam generator to the pressure transmitter; the pressure transmitter is used for converting a pressure signal of steam at the outlet of the second loop of the steam generator into a measurable direct current signal of-5- +5V or + 1- + 5V.
7. The PWR steam generator model time constant testing system of claim 1, wherein: the data acquisition device (10) adopts an EIC-02L type data acquisition system.
8. The method for testing the pressurized water reactor steam generator model time constant test system according to any one of claims 1 to 7, characterized in that: the method comprises the following steps:
1) the system comprises a reactor without materials, a measuring system is installed and a meter is debugged to be normal under the state that a unit is shut down, and the system specifically comprises a primary loop coolant inlet temperature measuring device (1) of a pressurized water reactor steam generator, which is installed on a coolant inlet pipeline of a steam generator, a primary loop coolant outlet temperature measuring device (2) of the pressurized water reactor steam generator, which is installed on a coolant outlet pipeline of the steam generator, a steam turbine opening regulating measuring device (5) which is installed on a valve rod of a steam turbine opening regulating servomotor, a generator power measuring device (6), a secondary loop steam temperature measuring device (3) of a steam generator outlet, which is installed on a steam outlet pipeline of a secondary loop of the steam generator, and a secondary loop steam pressure measuring device (4) of the steam generator outlet;
2) after the reactor is filled, the nuclear power unit is started and is loaded, the high-pressure regulating valve of the steam turbine is in a stable state, a step instruction is given to the high-pressure regulating valve, the opening of the high-pressure regulating valve is suddenly changed, the power of the power unit is also suddenly changed, and the time of 100 milliseconds is kept after the sudden change of the opening of the high-pressure regulating valve; in the test process, the strokes of the R rod and the G rod are kept unchanged, and the thermal power of the reactor is also unchanged;
3) starting a data acquisition device (10) 200 milliseconds before giving a step instruction to a high-pressure regulating valve, and recording the inlet temperature of a primary loop coolant of a pressurized water reactor steam generator, the outlet temperature of the primary loop coolant of the pressurized water reactor steam generator, the opening degree of a steam turbine regulating valve, the power of a generator, the steam temperature of a secondary loop of an outlet of the steam generator and the steam pressure of a secondary loop of the outlet of the steam generator;
4) the sudden change of the high-pressure regulating valve causes the sudden change of the main steam pressure of the two loops to cause the change of the inlet temperature and the outlet temperature of the reactor coolant, the inlet temperature and the outlet temperature of the steam generator coolant correspondingly change, and the simulation curve is consistent with the measured data through the least square method, so that the time constant of the steam generator model is determined.
9. The method for testing the pressurized water reactor steam generator model time constant test system according to claim 8, wherein the method comprises the following steps: and 4) establishing a steam generator model on ADPSS software, setting the working condition the same as that of the actual nuclear power unit, and enabling the simulation curve of the steam generator model to be consistent with the actually measured data of the nuclear power unit by a least square method so as to determine the time constant of the steam generator model.
10. The method for testing the pressurized water reactor steam generator model time constant test system according to claim 9, wherein the method comprises the following steps: the method for determining the steam generator model time constant specifically comprises the following steps: and calculating the time constant of the steam generator model by performing Laplace transform on the mathematical expression of the steam generator model.
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| CN110186026B (en) * | 2019-05-06 | 2020-09-08 | 中广核研究院有限公司 | Dry thermal start method for once-through steam generator |
| CN110568357A (en) * | 2019-09-10 | 2019-12-13 | 国网黑龙江省电力有限公司电力科学研究院 | nuclear power unit electric output monitoring and diagnosing system |
| CN113311729A (en) * | 2021-06-11 | 2021-08-27 | 国家工业信息安全发展研究中心 | Nuclear power control system safety test environment simulation device |
| CN115331538B (en) * | 2022-08-29 | 2024-05-28 | 中国舰船研究设计中心 | Steam generator secondary side edge simulation device for water supply system test |
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