CN116842752A - System-level simplified nonlinear modeling method for pressurized water reactor nuclear power plant - Google Patents
System-level simplified nonlinear modeling method for pressurized water reactor nuclear power plant Download PDFInfo
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Abstract
A simplified nonlinear modeling method for a pressurized water reactor nuclear power plant system level comprises the following steps: 1 production of nuclear energy in a nuclear reactor and conversion thereof into thermal energy; 2 energy exchange between the first loop and the second loop of the nuclear power plant; and 3, expanding and working the saturated steam in the second loop in the steam turbine system. The invention determines the average temperature of a loop coolant and the power of a turbine as output variables of a model, determines the effective length of a reactor core control rod and the opening degree of a regulating valve of the turbine as input variables of the model, and is determined by mechanism analysis, wherein a model frame is obtained based on conservation equations of various processes; model parameters are determined by analysis, derivation, fitting and nonlinear least squares identification based on pressurized water reactor nuclear power system large-scale operational data. On the premise of ensuring the simple structure of the model, the invention can comprehensively capture the system-level operation characteristics of the pressurized water reactor nuclear power plant and the nonlinear characteristics thereof under the condition of large-scale operation, and can meet the design requirements of the analysis and coordination control of the system-level characteristics of the pressurized water reactor nuclear power plant.
Description
Technical Field
The invention belongs to the field of nuclear energy control, and particularly relates to a system-level simplified nonlinear modeling method of a pressurized water reactor nuclear power plant.
Background
Nuclear energy is widely paid attention to the field of ships by the characteristics of high energy density, strong cruising ability and no dependence on air. The development of the nuclear power system for the ship has strategic significance for national defense and national economy of China. However, due to the mobility of the vessel, it requires frequent, rapid, widely varying loads to meet its movement needs, which puts higher demands on the coordinated operation and control of the system.
The related control of the nuclear power system still takes traditional PID control as a main means at present. While PID has the characteristics of simple design and easy implementation, it is more adept at dealing with the control problem of a linear, single variable process. The coupling between each equipment and parameter of the marine nuclear power system is serious, and the coupling shows stronger nonlinearity in the process of changing the load in a large range, so the PID can not meet the system-level coordination control and large-range flexible operation requirements of the marine nuclear power device. Model Predictive Control (MPC) is an excellent choice for solving the problems of multivariable, coupled, nonlinear process control. But MPC is a model-based control algorithm, the establishment of which predictive models is critical to the design of MPC controllers.
The existing dynamic mechanism models related to the marine nuclear power system mainly comprise 2 types: (1) a model of a single device built based on local links; (2) A system-wide mechanism model of a nuclear power plant comprising a plurality of devices. The local equipment model can only describe the operation characteristics of local links and cannot meet the system-level simulation and control design requirements of the nuclear power plant; the full system mechanism model of the nuclear power plant containing a plurality of devices can comprehensively reflect the operation characteristics of the system, but the model involves excessive variables and complex structures and is not suitable for control design. In addition, modeling ideas that can reflect the large-scale operating characteristics of a nuclear power plant include pure data-driven black-box models, such as neural network models, and multiple models weighted based on multiple linear models. However, the internal structure of the system is unknown and is only supported by limited data, and the system cannot cope with the unknown situation in the running process of the system, so that the safety problem is brought to the running of the system; the latter is only obtained by combining individual linear models and has limited capability of describing a wide range of operating characteristics of the system.
The prior art is as follows:
compared with the technology of the patent CN 109783936A' calculation method of variable working condition heat accumulation increment of nuclear island of pressurized water reactor nuclear power station:
patent CN109783936A is used for realizing calculation of variable-working-condition heat accumulation increment of a nuclear island of a nuclear power unit under the condition that the nuclear power unit does not have a true machine test, and the method aims at providing a system-level simplified nonlinear model of the nuclear power unit so as to promote application of a control algorithm based on the model in the field of nuclear energy control.
The calculation model of the nuclear power unit nuclear island variable working condition heat accumulation increment in the patent CN109783936A is a steady-state model, reflects a steady-state relation based on an energy conservation equation among main variables, and the system-level simplified nonlinear model of the pressurized water reactor nuclear power device is a dynamic model, and can comprehensively reflect the dynamic operation process of the nuclear power device.
The coolant average temperature-nuclear power calculation model in patent CN109783936a is obtained by a unitary linear regression method based on the micro-perturbation linear theory around the equilibrium point, which is essentially a linear model, whereas in the modeling method proposed by us, the model structure is obtained by system mechanism analysis, and the model parameters are obtained by an identification method based on the system operation data under a wide range of variable conditions, which is essentially a nonlinear model.
The key parameters related to the calculation method of the variable-working-condition heat accumulation increment of the nuclear island of the nuclear power unit proposed by the patent CN109783936A are only related to a nuclear reactor, a primary loop coolant and an evaporator in the nuclear power unit, the whole system characteristic of the nuclear power unit cannot be reflected, and the parameters related to the system-level simplified nonlinear model of the pressurized water reactor nuclear power unit disclosed by the patent CN109783936A cover all main equipment of the nuclear power unit, so that the whole system characteristic of the nuclear power unit can be comprehensively reflected.
The calculation method proposed by the patent CN109783936A can only be used for calculating the heat accumulation increment of the nuclear island of the nuclear power unit, and the model proposed by the patent CN109783936A is obtained based on the energy, mass, momentum and volume conservation relation of each link, so that the analysis and calculation of the thermodynamic and hydraulic processes of the nuclear power system can be comprehensively supported.
In contrast to the technology of patent CN102279901a "a modeling method for a third generation pressurized water reactor nuclear power unit":
patent CN102279901a aims to provide modeling ideas for the analysis of nuclear power plant systems, the proposed model is used as a simulation platform and controlled object, while the proposed pressurized water reactor nuclear power plant system level nonlinear simplified model aims to serve control, the function of which is used as a control model.
The modeling method disclosed by the patent CN102279901A relates to numerous parameters, has a complex model structure, can not support the application of a control algorithm based on a model in the coordination control of a nuclear power system reactor, and the simplified nonlinear model disclosed by the patent CN102279901A only relates to four key state variables, has a simple model structure and can promote the application of the control algorithm based on the model in the field of nuclear power control.
The modeling method proposed by patent CN102279901a does not consider turbine extraction, whereas the modeling concept proposed by us considers the effect of extraction from a turbine on feedwater temperature.
The pressurized water reactor nuclear power unit model established by the patent CN102279901A does not consider the influence of the dynamic process of the regenerative heater and the deaerator on the system time constant, but the modeling method provided by the inventor considers the influence of the high-pressure regenerative heater and the low-pressure regenerative heater and the deaerator on the whole dynamic process of the system, and the consideration is more practical.
In summary, in the field of marine nuclear power control, there is a lack of a high-quality and simple-structure dynamic model of a nuclear power plant system level to promote the application of a model-based control algorithm in the field of nuclear power control.
Disclosure of Invention
In order to solve the technical problems, the invention provides a system-level simplified nonlinear modeling method for a pressurized water reactor nuclear power plant, which not only can comprehensively describe the system-level operation characteristics of the pressurized water reactor nuclear power plant under a large-scale operation condition, but also has the characteristics of simple structure, safety and reliability, and can support the application of a control algorithm based on the model in the system-level coordinated operation design of the marine nuclear power plant.
In order to achieve the above purpose, the technical scheme adopted by the invention is as follows:
a pressurized water reactor nuclear power plant system level simplified nonlinear modeling method, comprising:
1) The actual operation process of the pressurized water reactor nuclear power plant is simplified into 3 processes:
the generation and conversion of nuclear energy in a loop, the energy exchange of the first loop side and the second loop side in a steam generator and the expansion and working of saturated steam in a steam turbine system;
2) Extracting nuclear reactor power, core fuel temperature, primary loop coolant average temperature, steam generator steam pressure and turbine power which can characterize the pressurized water reactor nuclear power plant process as key elements;
3) Determining a model framework based on conservation principles and mechanism analysis of each process;
4) And obtaining model parameters by adopting data analysis, deduction, fitting and identification methods.
As a further improvement of the invention, the model simplification in the actual operation process of the pressurized water reactor nuclear power plant in the step 1) is based on the following assumption:
(1) The pressure stabilizer in the first loop is provided with an independent controller to ensure the stability of the internal pressure and the water level, and the steam generator, the condenser, the high-pressure heater, the deaerator and the low-pressure heater in the second loop are provided with independent controllers to ensure the stability of the internal water level, so that the two-loop water supply flow, the steam generator steam flow and the steam flow entering the steam turbine are considered to be equal;
(2) The working medium in the steam generator is always in a saturated state;
(3) The process of expanding saturated steam in a turbine system is considered a static process.
As a further improvement of the invention, the nuclear reactor power, the reactor core fuel temperature, the average temperature of the primary loop coolant and the steam pressure of the secondary loop are selected as the state variables of the model, the average temperature of the primary loop coolant and the steam power of the steam turbine are selected as the output variables of the model, and the effective length of the reactor core control rod and the valve opening of the regulating valve of the steam turbine are selected as the input variables of the model;
the simplified nonlinear model of the pressurized water reactor nuclear power plant system level is as follows:
W st ＝f 5 (G st )·(G st h g -G fw h fw -Q chc ) (2)
T sg ＝f 2 (p sg ) (3)
h g ＝f 3 (p sg ) (4)
G st ＝f 4 (RV)·p sg (5)
Q chc ＝f 6 (G st ) (6)
h fw ＝f 7 (G st ) (7)
the model contains 2 input variables, 4 state variables, and 2 output variables:
wherein u represents an input vector, the effective length L of the control rod is Bu, and the valve opening RV of the regulating valve of the steam turbine is; x represents a state vector comprising nuclear reactor power W rp The unit is kW, the core fuel temperature T rf In degrees Celsius, average temperature T of the primary coolant ave Steam pressure p of steam generator at a temperature of sg The unit is bar; y is the output vector, comprising a loop coolant average temperature T ave Unit of turbine power W at a temperature of st The unit is kW; rou (L) represents the induced reactivity of the control rod; alpha rf A negative feedback coefficient representing core fuel temperature versus core reactivity; alpha ave A negative feedback coefficient representing the average temperature of the coolant of the primary loop versus the reactivity of the core; delta T rf The unit of change in core fuel temperature is expressed in deg.c; delta T ave Representing the average temperature of the coolant in a loopThe unit of change is deg.c; (kA) fc Indicating that the heat exchange unit between the core fuel and the primary loop coolant is kW/K; (kA) sgh The heat exchange unit between the first loop and the second loop of the steam generator is kW/K; t (T) sg The unit of the saturation temperature of the steam at the secondary side of the steam generator is represented as the temperature; p is p sg Indicating the steam pressure unit in the steam generator as bar; g fw The unit of feed water flow is kg/s; h is a fw Represents the unit of specific enthalpy of water supply as kJ/kg; g sg The steam flow unit of the steam generator is kg/s; h is a g The unit of specific enthalpy of saturated steam is kJ/kg; h is a f Represents that the specific enthalpy unit of saturated water is kJ/kg; ρ f Represents the saturated water density unit is kg/m3; ρ g Represents the saturated steam density unit is kg/m3; v (V) sgf Representing the volume unit of the liquid phase zone in the steam generator as m 3 ；V sgg Indicating the volume unit of vapor phase zone in the steam generator is m 3 ；V sg Representing the total volume unit of the steam generator as m 3 The method comprises the steps of carrying out a first treatment on the surface of the Psi is the unit conversion coefficient; g st Indicating the steam flow unit entering the turbine as kg/s; q (Q) chc The unit of the heat absorbed by the secondary loop working medium in the condenser is kW; d, d 1 -d 4 The dynamic parameters of the model to be determined are; f (f) 1 -f 7 For the function to be determined.
Compared with the prior art, the invention has the following advantages and beneficial effects:
the invention has the advantages that a novel simplified nonlinear modeling method of the pressurized water reactor nuclear power plant system level is provided, in the method, the model structure is determined by the mechanism analysis of the system, and the model can be ensured to comprehensively extract the operation characteristics of the pressurized water reactor nuclear power plant system level; the model parameters are obtained by analyzing, deducing, fitting and identifying the large-range nonlinear operation data of the pressurized water reactor nuclear power plant, so that the model can capture the nonlinear characteristics of the pressurized water reactor nuclear power plant caused by load change in the large-range operation process. Compared with the existing nuclear power plant model, the simplified nonlinear modeling method of the pressurized water reactor nuclear power plant system level can simultaneously meet the requirements of simple model structure and comprehensive description of the operation characteristics of the plant system level, and the proposed model can be used for control design and can support the application of a control algorithm based on the model in the nuclear energy field so as to realize and improve the large-scale and system level coordinated operation design of the nuclear power plant.
Drawings
FIG. 1 is an exploded view of a system process of one embodiment of the present invention;
FIG. 2 is a graph of model accuracy versus an embodiment of the present invention.
Detailed Description
The invention is described in further detail below with reference to the attached drawings and detailed description:
the invention discloses a system-level simplified nonlinear modeling method of a pressurized water reactor nuclear power plant, which is implemented as follows:
step 1: the actual operation process of the pressurized water reactor nuclear power plant is simplified into 3 main processes: the generation and conversion of nuclear energy in a loop, the energy exchange of a first loop side and a second loop side in a steam generator and the expansion work of saturated steam in a steam turbine system are shown in figure 1;
step 2: extracting nuclear reactor power, core fuel temperature, primary loop coolant average temperature, steam generator steam pressure and turbine power which can represent the main process of the pressurized water reactor nuclear power plant as key elements;
step 3: determining a model framework based on conservation principles and mechanism analysis of each process;
step 4: building a model on a simulation software Simulink/MATLAB platform;
step 5: extracting large-scale nonlinear operation data of the pressurized water reactor nuclear power plant based on the verified complex mechanism model of the pressurized water reactor nuclear power plant for the ship;
step 6: based on the acquired data, model parameters are acquired by adopting methods of analysis, deduction, fitting and identification.
The simplification of the model is based on the following assumptions:
(1) The pressure stabilizer in the first loop is provided with an independent controller to ensure the stability of the internal pressure and the water level, and the steam generator, the condenser, the high-pressure heater, the deaerator and the low-pressure heater in the second loop are provided with independent controllers to ensure the stability of the internal water level, so that the two-loop water supply flow, the steam generator steam flow and the steam flow entering the steam turbine are considered to be equal;
(2) The working medium in the steam generator is always in a saturated state;
(3) The process of expanding saturated steam in a turbine system is considered a static process.
The nuclear reactor power can be expressed as:
wherein W is rp Indicating the nuclear reactor power unit as kW; rou (L) represents the induced reactivity of the control rod; alpha rf A negative feedback coefficient representing core fuel temperature versus core reactivity; alpha ave A negative feedback coefficient representing the average temperature of the coolant of the primary loop versus the reactivity of the core; delta T rf The unit of change in core fuel temperature is expressed in deg.c; delta T ave Mean temperature change unit of the coolant in the first circuit is expressed as the temperature; d, d 1 For the model dynamic parameters to be determined.
The core fuel temperature can be expressed as:
wherein T is rf The fuel temperature is expressed in degrees celsius; t (T) ave Mean coolant temperature in degrees celsius; (kA) fc Indicating that the heat exchange unit between the core fuel and the primary loop coolant is kW/K; d, d 2 For the model dynamic parameters to be determined.
In a steam generator, the heat exchange between the primary circuit side coolant and the secondary circuit side working medium can be expressed as:
wherein, (kA) sgh The heat exchange unit between the first loop and the second loop of the steam generator is kW/K; t (T) sg The unit of the saturation temperature of the steam at the secondary side of the steam generator is represented as the temperature; d, d 3 For the model dynamic parameters to be determined.
The steam pressure within the steam generator can be expressed as:
wherein p is sg Indicating the steam pressure unit in the steam generator as bar; g fw The unit of feed water flow is kg/s; h is a fw Represents the unit of specific enthalpy of water supply as kJ/kg; g sg The steam flow unit of the steam generator is kg/s; h is a g The unit of specific enthalpy of saturated steam is kJ/kg; h is a f Represents that the specific enthalpy unit of saturated water is kJ/kg; ρ f Represents the saturated water density unit is kg/m3; ρ g Represents the saturated steam density unit is kg/m3; v (V) sgf Representing the volume unit of the liquid phase zone in the steam generator as m 3 ；V sgg Indicating the volume unit of vapor phase zone in the steam generator is m 3 ；V sg Representing the total volume unit of the steam generator as m 3 The method comprises the steps of carrying out a first treatment on the surface of the Psi is the unit conversion coefficient; z represents the denominator on the right side of the equation. Z has the following expression form according to the heat exchange mechanism in the steam generator:
Z＝f 1 (p sg ,T ave ) (5)
wherein f 1 (-) is a pending function.
Temperature T of saturated steam in evaporator sg And specific enthalpy h g And the vapor pressure p in the evaporator sg The following relationship exists:
T sg ＝f 2 (p sg ) (6)
h g ＝f 3 (p sg ) (7)
wherein f 2 (-) and f 3 (-) is a pending function;
the calculation formula of the steam flow entering the steam turbine is as follows:
G st ＝f 4 (RV)·p sg (8)
wherein G is st Indicating the steam flow unit entering the turbine as kg/s; RV represents that the valve opening unit of the regulating valve of the steam turbine is; p is p sg Indicating the steam pressure unit in the steam generator as bar; f (f) 4 (-) is a pending function.
The turbine power can be obtained by:
W st ＝f 5 (G st )·(G st h g -G fw h fw -Q chc ) (9)
wherein W is st Representing the power unit of the steam turbine as kW; q (Q) chc The unit of the heat absorbed by the secondary loop working medium in the condenser is kW; f (f) 5 (-) is a pending function.
Specific enthalpy of feed water h fw And heat Q absorbed by the two-loop working medium in the condenser chc Mainly by flow G of the turbine st The effect of (c) can be expressed as follows:
Q chc ＝f 6 (G st ) (10)
h fw ＝f 7 (G st ) (11)
wherein f 6 (-) and f 7 (-) is a pending function;
the influence of the thermal process of each device on the two loop sides of the pressurized water reactor nuclear power plant on the system time constant is used as a dynamic parameter d 4 Added to the solution equation of the steam pressure of the steam generator:
wherein d 4 For the model dynamic parameters to be determined.
The simplified nonlinear model of the pressurized water reactor nuclear power plant system level is as follows:
W st ＝f 5 (G st )·(G st h g -G fw h fw -Q chc ) (14)
T sg ＝f 2 (p sg ) (15)
h g ＝f 3 (p sg ) (16)
G st ＝f 4 (RV)·p sg (17)
Q chc ＝f 6 (G st ) (18)
h fw ＝f 7 (G st ) (19)
the model contains 2 input variables, 4 state variables, and 2 output variables:
wherein u represents an input vector, the effective length L of the control rod is Bu, and the valve opening RV of the regulating valve of the steam turbine is; x represents a state vector comprising nuclear reactor power W rp The unit is kW, the core fuel temperature T rf In degrees Celsius, average temperature T of the primary coolant ave Steam pressure p of steam generator at a temperature of sg The unit is bar; y is the output vector, comprising a loop coolant average temperature T ave Unit of turbine power W at a temperature of st The unit is kW.
The model parameters to be determined include 4 static parameters ((kA) fc ，(kA) sgh ，α rf ，α ave ) 7 functions to be determined (f 1 -f 7 ) And 4 dynamic parameters (d 1 -d 4 )。
Based on the model structure, the simplified nonlinear model of the pressurized water reactor nuclear power plant system level is built on a Simulink/MATLAB platform, and based on the verified large-scale nonlinear operation data of the pressurized water reactor nuclear power plant extracted by the marine pressurized water reactor nuclear power plant complex mechanism model, static parameters and functions to be determined of the pressurized water reactor nuclear power plant are obtained by adopting analysis, deduction and fitting methods. And based on the collected large-range nonlinear operation data of the pressurized water reactor nuclear power plant, the nonlinear least square method is adopted to identify and determine the dynamic parameters of the model. The nonlinear least squares method follows the following objective function in the solution process:
where d is the dynamic parameter to be determined, d= [ d ] 1 ,d 2 ,d 3 ,d 4 ]；Y p Simplifying the data set of the nonlinear model for the system level from the pressurized water reactor nuclear power plant to be constructed, Y P ＝[W rp ,T rf ,T ave ,p sg ]；Y M Is the large-scale nonlinear operation data of the pressurized water reactor nuclear power plant, Y M ＝[W rp ,T rf ,T ave ,p sg ]。
After the model parameters are determined, the simplified nonlinear model (simplified model) of the pressurized water reactor nuclear power plant system level provided by the invention is compared with the complex mechanism model (data) of the pressurized water reactor nuclear power plant for the ship, and the precision of the proposed model is verified, as shown in figure 2.
Therefore, the system-level simplified nonlinear modeling method of the pressurized water reactor nuclear power plant provided by the invention can accurately and comprehensively capture the system-level operation characteristics of the marine nuclear power plant under the condition of large-scale operation.
The above description is only of the preferred embodiment of the present invention, and is not intended to limit the present invention in any other way, but is intended to cover any modifications or equivalent variations according to the technical spirit of the present invention, which fall within the scope of the present invention as defined by the appended claims.
Claims (3)
1. A simplified nonlinear modeling method for a pressurized water reactor nuclear power plant system level is characterized by comprising the following steps of: comprising the following steps:
1) The actual operation process of the pressurized water reactor nuclear power plant is simplified into 3 processes:
the generation and conversion of nuclear energy in a loop, the energy exchange of the first loop side and the second loop side in a steam generator and the expansion and working of saturated steam in a steam turbine system;
2) Extracting nuclear reactor power, core fuel temperature, primary loop coolant average temperature, steam generator steam pressure and turbine power which can characterize the pressurized water reactor nuclear power plant process as key elements;
3) Determining a model framework based on conservation principles and mechanism analysis of each process;
4) And obtaining model parameters by adopting data analysis, deduction, fitting and identification methods.
2. The simplified nonlinear modeling method for a pressurized water reactor nuclear power plant system level according to claim 1, wherein the method comprises the following steps:
the model simplification in the actual operation process of the pressurized water reactor nuclear power plant in the step 1) is based on the following assumption:
(1) The pressure stabilizer in the first loop is provided with an independent controller to ensure the stability of the internal pressure and the water level, and the steam generator, the condenser, the high-pressure heater, the deaerator and the low-pressure heater in the second loop are provided with independent controllers to ensure the stability of the internal water level, so that the two-loop water supply flow, the steam generator steam flow and the steam flow entering the steam turbine are considered to be equal;
(2) The working medium in the steam generator is always in a saturated state;
(3) The process of expanding saturated steam in a turbine system is considered a static process.
3. The simplified nonlinear modeling method for a pressurized water reactor nuclear power plant system level according to claim 1, wherein the method comprises the following steps:
selecting nuclear reactor power, reactor core fuel temperature, primary loop coolant average temperature and secondary loop steam pressure as state variables of a model, selecting primary loop coolant average temperature and turbine power as output variables of the model, and selecting reactor core control rod effective length and turbine regulating valve opening as input variables of the model;
the simplified nonlinear model of the pressurized water reactor nuclear power plant system level is as follows:
W st ＝f 5 (G st )·(G st h g -G fw h fw -Q chc ) (2)
T sg ＝f 2 (p sg ) (3)
h g ＝f 3 (p sg ) (4)
G st ＝f 4 (RV)·p sg (5)
Q chc ＝f 6 (G st ) (6)
h fw ＝f 7 (G st ) (7)
the model contains 2 input variables, 4 state variables, and 2 output variables:
wherein u represents an input vector, the effective length L of the control rod is Bu, and the valve opening RV of the regulating valve of the steam turbine is; x represents a state vector comprising nuclear reactor power W rp The unit is kW, the core fuel temperature T rf The unit is at the temperature of,Average temperature T of coolant in one circuit ave Steam pressure p of steam generator at a temperature of sg The unit is bar; y is the output vector, comprising a loop coolant average temperature T ave Unit of turbine power W at a temperature of st The unit is kW; rou (L) represents the induced reactivity of the control rod; alpha rf A negative feedback coefficient representing core fuel temperature versus core reactivity; alpha ave A negative feedback coefficient representing the average temperature of the coolant of the primary loop versus the reactivity of the core; delta T rf The unit of change in core fuel temperature is expressed in deg.c; delta T ave Mean temperature change unit of the coolant in the first circuit is expressed as the temperature; (kA) fc Indicating that the heat exchange unit between the core fuel and the primary loop coolant is kW/K; (kA) sgh The heat exchange unit between the first loop and the second loop of the steam generator is kW/K; t (T) sg The unit of the saturation temperature of the steam at the secondary side of the steam generator is represented as the temperature; p is p sg Indicating the steam pressure unit in the steam generator as bar; g fw The unit of feed water flow is kg/s; h is a fw Represents the unit of specific enthalpy of water supply as kJ/kg; g sg The steam flow unit of the steam generator is kg/s; h is a g The unit of specific enthalpy of saturated steam is kJ/kg; h is a f Represents that the specific enthalpy unit of saturated water is kJ/kg; ρ f Represents the saturated water density unit is kg/m3; ρ g Represents the saturated steam density unit is kg/m3; v (V) sgf Representing the volume unit of the liquid phase zone in the steam generator as m 3 ；V sgg Indicating the volume unit of vapor phase zone in the steam generator is m 3 ；V sg Representing the total volume unit of the steam generator as m 3 The method comprises the steps of carrying out a first treatment on the surface of the Psi is the unit conversion coefficient; g st Indicating the steam flow unit entering the turbine as kg/s; q (Q) chc The unit of the heat absorbed by the secondary loop working medium in the condenser is kW; d, d 1 -d 4 The dynamic parameters of the model to be determined are; f (f) 1 -f 7 For the function to be determined.
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