CN113239539A - Method and system for predicting process of power failure accident of whole plant and computer readable storage medium - Google Patents

Method and system for predicting process of power failure accident of whole plant and computer readable storage medium Download PDF

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CN113239539A
CN113239539A CN202110512434.XA CN202110512434A CN113239539A CN 113239539 A CN113239539 A CN 113239539A CN 202110512434 A CN202110512434 A CN 202110512434A CN 113239539 A CN113239539 A CN 113239539A
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杨磊
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Abstract

The invention relates to the technical field of accident prediction of nuclear power plants, in particular to a method and a system for predicting the process of a whole plant outage accident and a computer readable storage medium, comprising the following steps: step 1: on the premise of not considering the mass exchange between the steam and the water in the voltage stabilizer, calculating the steam space volume of the voltage stabilizer before the safety valve of the voltage stabilizer jumps, and obtaining the volume expansion amount of the liquid phase water of the primary loop; step 2: calculating the enthalpy rise of the coolant in the loop before the tripping of the safety valve of the voltage stabilizer; and step 3: calculating the tripping time of the safety valve of the voltage stabilizer; and 4, step 4: updating the liquid phase water quality instant acquisition data of a loop and the space volume instant acquisition data of the voltage stabilizer; calculating the mass of liquid phase water of a loop and the volume of steam space of the voltage stabilizer after the related safety valve of the voltage stabilizer jumps; the method is used for solving the technical problem of how to quickly predict the whole plant outage of the nuclear power plant.

Description

Method and system for predicting process of power failure accident of whole plant and computer readable storage medium
Technical Field
The invention relates to the technical field of accident prediction of nuclear power plants, in particular to a method and a system for predicting the process of a whole plant outage accident and a computer readable storage medium.
Background
The whole plant power failure (SBO) accident is the most typical design benchmark accident of a pressurized water reactor nuclear power plant, and the accident process generally comprises the whole plant power failure of the nuclear power plant, host tripping, reactor shutdown, primary circuit system temperature and pressure rise, voltage stabilizer safety valve circulation tripping-seat returning pressure relief, rapid water level reduction of a voltage stabilizer and a pressure vessel, reactor core exposure of the reactor, zirconium water reaction on the surface of a fuel element and reactor core melt-down. The nature of the Japanese Fudao nuclear accident is the factory outage accident. Therefore, the nuclear power plant is very concerned with the simulation prediction and analysis work of the SBO accident process. At present, the forecast analysis of the research process of the design benchmark accident analysis aiming at SBO and the like of domestic and foreign nuclear power plants is mainly developed based on the large-scale thermal hydraulic programs of the nuclear power plants such as RELAP5, MELCOR, MAAP5 and the like. The large-scale thermal hydraulic programs accurately solve the energy, mass and momentum equations (six equations of two fluids) of two fluids (multiphase flow) of each pipeline or equipment on each time step by finely simulating the flowing heat exchange process in the fluid system of the nuclear power plant, so as to realize the simulation of the accident process. In the prior art, a specific simulation analysis model or a simulation system is developed for a specific heap type, so that simulation analysis of design basis accidents such as SBO (boundary layer operation) is realized. The technical scheme has the following distinct characteristics: firstly, the modeling is complex, the real-time or overtime simulation is not easy to realize, and the simulation process is easy to interrupt. The scheme generally needs to carry out high-precision modeling on a nuclear power plant reactor, a system fluid network of a primary loop system, a secondary loop system, a special safety facility and the like, equipment structure and operating characteristics, nuclear reaction heat generation and heat transfer, material physical properties, instruments, control and the like, and the model scale is generally large. Generally, a thermal hydraulic control body (including equipment components such as pump valves) of a nuclear power plant simulation model is generally above 100 nodes. When accident simulation is carried out, the multi-fluid energy, mass and momentum conservation equation matrix solution needs to be carried out on all control bodies, the real-time or overtime simulation of the accident process is difficult to realize, and the simulation process is often interrupted due to the fact that water physical parameters such as instantaneous temperature, pressure and gas content of the control bodies are abnormal. Secondly, the requirement on professional ability of users is high, and the use difficulty of a primary circuit operator is high. Generally speaking, the accident analysis model developer not only needs a certain professional knowledge, but also can be skilled in using large-scale thermal hydraulic programs of nuclear power plants such as RELAP5, MELCOR, MAAP5 and the like. After the accident analysis model is developed, the accident analysis model can be used only by further packaging into a simulator or a simulator. At this time, the simulation analysis system often needs a plurality of operators to run smoothly. The operation difficulty of single analysis is higher.
Therefore, it is necessary to develop a plant blackout accident progress prediction method, system and computer readable storage medium for solving at least one of the above technical problems.
Disclosure of Invention
The invention aims to provide a method and a system for predicting the whole plant outage accident process of a nuclear power plant and a computer readable storage medium, and solves the technical problem of how to rapidly predict the whole plant outage of the nuclear power plant.
In order to achieve the purpose, the invention adopts the following technical scheme:
on one hand, the invention provides a method for rapidly predicting the whole plant outage accident process of a nuclear power plant, which comprises the following operation steps:
step 1: on the premise of not considering the mass exchange between the steam and the water in the voltage stabilizer, calculating the steam space volume of the voltage stabilizer before the safety valve of the voltage stabilizer jumps, and obtaining the volume expansion amount of the liquid phase water of the primary loop;
step 2: calculating the enthalpy rise of the coolant in the loop before the tripping of the safety valve of the voltage stabilizer;
and step 3: calculating the tripping time of the safety valve of the voltage stabilizer; assuming that the safety valve of the voltage stabilizer is at t after the previous take-off1Returning to the seat at moment, and circularly calculating t1To t2The relative error between the integral value of the effective instantaneous thermal power of the reactor and the enthalpy rise of the coolant at the moment is t when the relative error is smaller than the energy convergence error2Namely the tripping time of the safety valve of the voltage stabilizer; the effective instantaneous thermal power of the reactor is the difference value of the instantaneous decay heat of the reactor, the boundary heat dissipation of the first loop and the heat conduction of the second loop;
and 4, step 4: updating the liquid phase water quality instant acquisition data of the loop and the space volume instant acquisition data of the voltage stabilizer; and calculating the mass of liquid phase water in the loop and the volume of the steam space of the pressure stabilizer after the safety valve of the pressure stabilizer jumps.
Preferably, in the step 1, based on the formula:
Figure BDA0003060803520000021
calculating the steam space volume of the voltage stabilizer before the tripping of the safety valve of the voltage stabilizer; in the formula: p1、P2Respectively two instantaneous state pressures of the pressure stabilizer steam space in the time interval of the take-off of the safety valve, V1、V2The volume of the steam space of the pressure stabilizer, T, corresponding to two states respectively1,sat、T2,satThe saturation temperatures for the two states are respectively.
Preferably, in step 2, the enthalpy rise of the primary loop coolant is based on the pressure and specific volume parameters of the primary loop water at the front and rear points of expansion, and is combined with the query H2And calculating to obtain the water physical property table.
Preferably, in said step 3, t2＝t1+ i × dt, i is a positive integer.
Preferably, in the step 4, based on the formula: Δ M ═ Δ Vvap × ρ vap _ close or Δ M ═ Δ Vvap×ρvap_close+(ΔVliq-ΔVvap)×ρliq_closeCalculating the quality of the liquid phase water of the loop after the tripping of the safety valve of the voltage stabilizer; based on the formula: v2＝max((V1-ΔVliq) 0.0) calculating the steam space volume of the voltage stabilizer; in the formula: Δ M is the total loss of primary coolant, kg; Δ VvapVolume of primary loop steam and liquid phase water discharged from safety valve, unit m3；ΔVliqIs the expansion volume of liquid phase water in a loop, and has unit m3；ρvap_close、ρliq_closeThe densities of saturated steam and saturated water are respectively the recoil pressure of the safety valve; v1、V2The volume of the pressure stabilizer steam space before the pressure stabilizer safety valve takes off and after the pressure stabilizer safety valve returns to the seat respectively is unit m3。
Preferably, the method further comprises the step 5: and calculating the liquid level change of the core active area and the zirconium water reaction starting time according to the updated primary loop liquid phase water mass, and further predicting the damage and melting time of the nuclear fuel elements of the reactor.
On the other hand, the invention also provides a system for rapidly predicting the whole plant outage accident process of the nuclear power plant, which comprises the following steps:
the steam space parameter calculation module of the voltage stabilizer comprises: the method is used for calculating the steam space volume of the pressure stabilizer before the safety valve of the pressure stabilizer jumps on the premise of not considering the mass exchange between the steam and the water in the pressure stabilizer, and obtaining the volume expansion amount of the liquid phase water of the primary loop;
a coolant enthalpy rise calculation module: the system is used for calculating the enthalpy rise of the coolant of a loop before the tripping of the safety valve of the voltage stabilizer;
the voltage stabilizer safety valve take-off time calculation module: the device is used for calculating the tripping time of the safety valve of the voltage stabilizer; assuming that the pressurizer safety valve is at t after the previous take-off1Returning to the seat at moment, and circularly calculating t1To t2The relative error between the integral value of the effective instantaneous thermal power of the reactor and the enthalpy rise of the coolant at the moment is t when the relative error is smaller than the energy convergence error2Namely the tripping time of the safety valve of the voltage stabilizer; the effective instantaneous thermal power of the reactor is the difference value of the instantaneous decay heat of the reactor and heat traps for heat dissipation of the first loop, heat conduction of the second loop and the like;
a loop liquid phase water quality and pressure stabilizer steam space volume calculation module: the system is used for updating the liquid phase water quality instant acquisition data of a loop and the space volume instant acquisition data of the voltage stabilizer; and calculating the mass of liquid phase water of the loop and the volume of steam space of the voltage stabilizer after the related safety valve of the voltage stabilizer jumps.
Preferably, based on the formula:
Figure BDA0003060803520000031
calculating the steam space volume of the voltage stabilizer before the tripping of the safety valve of the voltage stabilizer; in the formula: p1、P2Respectively two instantaneous state pressures of the pressure stabilizer steam space in the time interval of the take-off of the safety valve, V1、V2The volume of the steam space of the pressure stabilizer, T, corresponding to two states respectively1,sat、T2,satThe saturation temperatures for the two states are respectively.
Preferably, the enthalpy rise of the primary loop coolant is based on the pressure and specific volume parameters of the front and rear points of the primary loop water expansion, and is combined with the query H2And calculating to obtain the water physical property table.
Preferably, t is2＝t1+ i × dt, i is a positive integer; based on the formula:
ΔM＝ΔVvap×ρvap_close
or
ΔM＝ΔVvap×ρvap_close+(ΔVliq-ΔVvap)×ρliq_close
Calculating the quality of the liquid phase water of the loop after the tripping of the safety valve of the voltage stabilizer;
based on the formula:
V2＝max((V1-ΔVliq),0.0)
calculating the steam space volume of the voltage stabilizer; in the formula: Δ M is the total loss of primary coolant, kg;
wherein, is Δ VvapVolume of primary loop steam and liquid phase water discharged from safety valve, unit m3；ΔVliqIs the expansion volume of liquid phase water in a loop, and has unit m3；ρvap_close、ρliq_closeThe densities of saturated steam and saturated water are respectively the recoil pressure of the safety valve; v1、V2The volume of the pressure stabilizer steam space before the pressure stabilizer safety valve takes off and after the pressure stabilizer safety valve returns to the seat respectively is unit m3。
Further, the present invention also provides a computer readable storage medium storing a computer program, which when executed by a processor, is used to implement the method for rapidly predicting the progress of the nuclear power plant blackout accident.
In summary, the technical scheme has the beneficial effects that:
1. the method can solve the problems of high difficulty in real-time or overtime analysis and prediction of the SBO accident and easy interruption of the calculation process in the prior art, and realize the ultra-real-time stable analysis of the SBO accident process of the nuclear power plant; furthermore, the technical difficulty of SBO accident analysis of the nuclear power plant can be reduced, and the use requirements of common operation operators of the nuclear power plant can be met; in addition, the invention introduces a new technical method for the nuclear power plant SBO accident analysis, and can be used for carrying out comparison and verification with the prior technical scheme.
2. The method for constructing the analysis model only needs a small number of accident parameters and macroscopic design parameters of the nuclear power plant, and does not need to accurately master the design parameters related to the size structure and the flow heat exchange of the system equipment. The modeling data requirements are about 10% of the best existing solutions. The labor cost of modeling work is about 1% of the best technical scheme in the prior art.
3. When the method is adopted to analyze the model and predict the accident process, only the mass equation and the energy equation are required to be continuously and progressively searched along the time scale, iterative calculation on parameters such as two-fluid six equations and the like at each time step is not required, and the calculation time can be greatly reduced. When the SBO accident process with the total time scale of 24 hours is simulated and runs on a single-core CPU with the dominant frequency of 2.0GHz, the accident process prediction time is generally 1-100 seconds, and the accident process prediction time is generally 1-12 hours according to the best technical scheme at present.
Drawings
FIG. 1 is a schematic diagram illustrating an SBO accident process analysis of a safety valve operation cycle of a pressure stabilizer according to an embodiment of the present invention;
Detailed Description
The technical solutions in the embodiments of the present invention will be clearly and completely described below with reference to the drawings in the embodiments of the present invention, and it is obvious that the described embodiments are only a part of the embodiments of the present invention, and not all of the embodiments. All other embodiments, which can be derived by a person skilled in the art from the embodiments given herein without making any creative effort, shall fall within the protection scope of the present invention.
The first embodiment is as follows:
the technical principle of the rapid prediction method for the whole plant outage accident process of the nuclear power plant in the embodiment is as follows:
dividing a primary loop coolant area into two parts of pressure stabilizer steam space steam and primary loop liquid phase water, and respectively establishing the following equations:
1. based on the ideal gas state equation, a pressurizer steam space state approximate equation (1)) before the safety valve is opened is established, so that the volume-pressure change of the pressurizer steam is solved, and the volume-pressure change of the primary loop coolant is further solved.
2. Based on energy conservation, a relation (expression (2)) between heat loss and coolant enthalpy rise, such as accumulated decay heat of a primary circuit, heat conduction of a secondary circuit, heat dissipation of the primary circuit and the like, is established.
3. Establishing ideal gas state equations (formula (3) and formula (4)) of a safety valve jump-seat return state of the voltage stabilizer based on the ideal gas state equations, and solving the discharge amount (formula (5) and formula (6)) of a return coolant and a new space volume (formula (7)) of the voltage stabilizer; based on the conservation of mass and conservation of energy, the primary loop residual coolant mass and its enthalpy (equations (8) and (9)) can be solved.
The process is an SBO accident process analysis process of the action period of the safety valve of the voltage stabilizer. The accident analysis flow chart is designed according to the method and is shown in figure 1.
The calculation of the relational expressions by the expressions (1) to (9) is described below:
the regulator gas space state equation is established based on an ideal gas state equation:
Figure BDA0003060803520000061
in the formula: p1、P2Respectively two instantaneous state pressures of the pressure stabilizer steam space in the time interval of the take-off of the safety valve, V1、V2The volume of the steam space of the pressure stabilizer, T, corresponding to two states respectively1,sat、T2,satThe saturation temperatures for the two states are respectively.
The enthalpy rise absorbed heat calculation relation of the primary loop coolant is as follows:
Figure BDA0003060803520000062
in the formula: t is t0、t1The times of occurrence of the two states, s, respectively; Δ H is t0To t1Enthalpy difference of primary loop coolant at the moment, J/kg; q (t) is a loop transient decay heat, w; s (t) is the power of n kinds of loop heat traps, w, such as two loop heat conduction and one loop heat dissipation.
When the volume expansion of the coolant in the primary loop is not considered, the gas space state parameter relation of the primary safety valve take-off and return seat transient voltage stabilizer is as follows:
Figure BDA0003060803520000063
the corrected relation after considering the volume expansion of the primary circuit coolant is as follows:
Figure BDA0003060803520000064
in the formula: popen、PcloseRespectively the take-off pressure and the recoil pressure of the safety valve in unit Pa; vopenThe volume of the steam space of the pressure stabilizer in unit of m when the safety valve jumps3；ΔVvapThe volume expansion of the pressure stabilizer in the steam space at the take-off pressure and the recoil pressure of the safety valve is unit m3；Tsat,open、Tsat,closeThe saturated temperature of the water is respectively corresponding to the tripping pressure and the recoil pressure of the safety valve. Δ VliqVolume expansion of coolant in unit m for primary circuit in two pressure states from tripping of safety valve to return seat3。
When Δ VliqLess than Δ VvapAt this time, the quality of the primary loop coolant discharged by the primary safety valve actuation is as follows:
ΔM＝ΔVvap×ρvap_closeformula (5)
When Δ VliqGreater than Δ VvapIn time, the primary safety valve action loop loses coolant mass as follows:
ΔM＝ΔVvap×ρvap_close+(ΔVliq-ΔVvap)×ρliq_closeformula (6)
In the formula: Δ M is the total loss of primary coolant, kg; Δ VvapVolume of primary loop steam and liquid phase water discharged from safety valve, unit m3；ΔVliqIs the expansion volume of liquid phase water in a loop, and has unit m3；ρvap_close、ρliq_closeThe densities of saturated steam and saturated water are respectively the pressure of the safety valve during the recoil.
After the primary safety valve jumps and returns to the seat, the steam space volume of the pressure stabilizer is as follows:
V2＝max((V1-ΔVliq) 0.0) formula (7)
After the primary safety valve jumps and returns to the seat, the quality of primary loop liquid phase water is as follows:
M2＝M1-max((ΔVliq-ΔVvap)×ρliq_close) 0.0) formula (8)
After the primary safety valve is tripped and returns to the seat, the enthalpy of primary loop liquid phase water is as follows:
Figure BDA0003060803520000071
in the formula: v1、V2The volume of the pressure stabilizer steam space before the pressure stabilizer safety valve takes off and after the pressure stabilizer safety valve returns to the seat respectively is unit m3；M1、M2、H1、H2Respectively the mass and enthalpy of the loop liquid phase water before take-off and after return seat of the safety valve of the voltage stabilizer, and the unit is m3And J/kg.
In this embodiment, the SBO accident process analysis model requires 15 types of input (assumed) parameters:
1. nuclear power plant macro design (operating) parameters:
(1) the water content of a loop system;
(2) the safety valve of the voltage stabilizer has the pressure of jumping and returning to the base and rated discharge capacity;
(3) a potentiostat volume;
(4) loading uranium in a reactor;
(5) a loop average temperature-heat dissipation power meter.
2. Accident-related parameters:
(6) loop pressure before an accident;
(7) reactor inlet and outlet temperatures before an accident;
(8) pressurizer water volume before accident;
(9) pre-accident reactor power operating history (power and time);
(10) reactor shut-down time;
(11) the power-off time of the whole plant;
(12) the steam flow for the second loop-the time table after the whole plant is powered off;
(13) and (5) a pressure curve of the two loops after the whole plant is powered off.
3. Analyzing hypothesis-related parameters
(14) A time step;
(15) allowable convergence errors, including energy convergence errors and mass convergence errors;
the detailed analysis procedure of this example is as follows:
suppose that the core flow of the nuclear power plant SBO accident progress analysis of this patent includes 4 stages, as shown in fig. 1:
1. calculating the gas space volume of the pressure stabilizer before the tripping of the safety valve:
and (3) calculating the space volume of the steam of the pressure stabilizer before the tripping of the safety valve based on the formula (1) without considering the mass exchange between the steam and the water in the pressure stabilizer, and further calculating the volume expansion amount of the liquid phase water in the primary loop.
2. Calculating the enthalpy rise of the primary loop coolant before the tripping of the safety valve:
based on the parameters (pressure, specific volume) of the point before and after the expansion of the primary water, the enthalpy rise of the return coolant can be calculated by inquiring the H2O water physical table.
3. Calculating the tripping time of the safety valve:
assuming that the safety valve of the voltage stabilizer is at t after the previous take-off1Returning to the seat at moment, and circularly calculating t1To t2(t2＝t1+ i × dt, i ═ 1,.., n) the relative error between the integral value of the instantaneous thermal power of the reactor and the enthalpy rise of the coolant, when the relative error is smaller than the energy convergence error, t2Namely the tripping time of the safety valve.
The effective instantaneous thermal power of the reactor is the difference between the instantaneous decay heat of the reactor and the heat traps for heat dissipation and heat conduction of the two loops.
4. Updating the liquid phase water mass of the primary loop and the steam space volume of the pressure stabilizer;
and calculating the quality of the liquid phase water in the loop after the tripping of the safety valve of the voltage stabilizer based on the formula (5) or the formula (6). The pressurizer vapor space volume was calculated based on equation (7).
And for a specific nuclear power plant, parameters such as the liquid level of the core active area and the like can be calculated according to the updated primary loop liquid phase water mass.
Example 2:
the embodiment provides a method for quickly predicting the whole plant outage accident process of a nuclear power plant on the basis of the embodiment 1, which can quickly predict the whole plant outage of the nuclear power plant, including but not limited to; the method comprises the following operation steps:
step 1: on the premise of not considering the mass exchange between the steam and the water in the pressure stabilizer, calculating the steam space volume of the pressure stabilizer before the safety valve jumps, and obtaining the volume expansion amount of the liquid phase water in the primary loop;
step 2: calculating the enthalpy rise of the coolant of the circuit before the tripping of the safety valve;
and step 3: calculating the take-off time of the safety valve; assuming that the safety valve of the voltage stabilizer is at t after the previous take-off1Returning to the seat at moment, and circularly calculating t1To t2The relative error between the integral value of the effective instantaneous thermal power of the reactor and the enthalpy rise of the coolant at the moment is t when the relative error is smaller than the energy convergence error2Namely the tripping time of the safety valve; the effective instantaneous thermal power of the reactor is the difference value of the instantaneous decay heat of the reactor, the boundary heat dissipation of the first loop and the heat conduction of the second loop;
and 4, step 4: updating the liquid phase water quality instant acquisition data of the loop and the space volume instant acquisition data of the voltage stabilizer; and calculating the mass of liquid phase water in the loop and the volume of the steam space of the pressure stabilizer after the safety valve of the pressure stabilizer jumps.
Example three:
the embodiment further provides a system for rapidly predicting the whole plant outage accident process of a nuclear power plant on the basis of the embodiment 2, which includes:
the steam space parameter calculation module of the voltage stabilizer comprises: the method is used for calculating the steam space volume of the pressure stabilizer before the safety valve of the pressure stabilizer jumps on the premise of not considering the mass exchange between the steam and the water in the pressure stabilizer, and obtaining the volume expansion amount of the liquid phase water of the primary loop;
a coolant enthalpy rise calculation module: the system is used for calculating the enthalpy rise of the coolant of a loop before the tripping of the safety valve of the voltage stabilizer;
the voltage stabilizer safety valve take-off time calculation module: the device is used for calculating the tripping time of the safety valve of the voltage stabilizer; assuming that the pressurizer safety valve is at t after the previous take-off1Returning to the seat at moment, and circularly calculating t1To t2The relative error between the integral value of the effective instantaneous thermal power of the reactor and the enthalpy rise of the coolant at the moment is t when the relative error is smaller than the energy convergence error2Namely the tripping time of the safety valve of the voltage stabilizer; the effective instantaneous thermal power of the reactor is the difference value of the instantaneous decay heat of the reactor and heat traps for heat dissipation of the first loop, heat conduction of the second loop and the like;
a loop liquid phase water quality and pressure stabilizer steam space volume calculation module: the system is used for updating the liquid phase water quality instant acquisition data of a loop and the space volume instant acquisition data of the voltage stabilizer; and calculating the mass of liquid phase water of the loop and the volume of steam space of the voltage stabilizer after the related safety valve of the voltage stabilizer jumps.
Example four:
the present embodiment provides a computer-readable storage medium on the basis of the above-mentioned embodiments, and stores a computer program, which is used for implementing the method in the foregoing embodiments when being executed by a processor.
In summary, the technical scheme has the beneficial effects that:
1. the method can solve the problems of high difficulty in real-time or overtime analysis and prediction of the SBO accident and easy interruption of the calculation process in the prior art, and realize the ultra-real-time stable analysis of the SBO accident process of the nuclear power plant; furthermore, the technical difficulty of SBO accident analysis of the nuclear power plant can be reduced, and the use requirements of common operation operators of the nuclear power plant can be met; in addition, the invention introduces a new technical method for the nuclear power plant SBO accident analysis, and can be used for carrying out comparison and verification with the prior technical scheme.
2. The method for constructing the analysis model only needs a small number of accident parameters and macroscopic design parameters of the nuclear power plant, and does not need to accurately master the design parameters related to the size structure and the flow heat exchange of the system equipment. The modeling data requirements are about 10% of the best existing solutions. The labor cost of modeling work is about 1% of the best technical scheme in the prior art.
3. When the method is adopted to analyze the model and predict the accident process, only the mass equation and the energy equation are required to be continuously and progressively searched along the time scale, iterative calculation on parameters such as two-fluid six equations and the like at each time step is not required, and the calculation time can be greatly reduced. When the SBO accident process with the total time scale of 24 hours is simulated and runs on a single-core CPU with the dominant frequency of 2.0GHz, the accident process prediction time is generally 1-100 seconds, and the accident process prediction time is generally 1-12 hours according to the best technical scheme at present.
The above description is only for the preferred embodiment of the present invention, but the scope of the present invention is not limited thereto, and any person skilled in the art should be considered to be within the technical scope of the present invention, and the technical solutions and the inventive concepts thereof according to the present invention should be equivalent or changed within the scope of the present invention.

Claims (10)

1. A method for rapidly predicting the whole plant outage accident process of a nuclear power plant is characterized by comprising the following operation steps:
step 1: on the premise of not considering the mass exchange between the steam and the water in the voltage stabilizer, calculating the steam space volume of the voltage stabilizer before the safety valve of the voltage stabilizer jumps, and obtaining the volume expansion amount of the liquid phase water of the primary loop;
step 2: calculating the enthalpy rise of the coolant in the loop before the tripping of the safety valve of the voltage stabilizer;
and step 3: calculating the tripping time of the safety valve of the voltage stabilizer; suppose thatThe safety valve of the voltage stabilizer is at t after the previous take-off1Returning to the seat at moment, and circularly calculating t1To t2The relative error between the integral value of the effective instantaneous thermal power of the reactor and the enthalpy rise of the coolant at the moment is t when the relative error is smaller than the energy convergence error2Namely the tripping time of the safety valve of the voltage stabilizer; the effective instantaneous thermal power of the reactor is the difference value of the instantaneous decay heat of the reactor, the boundary heat dissipation of the first loop and the heat conduction of the second loop;
and 4, step 4: updating the liquid phase water quality instant acquisition data of a loop and the space volume instant acquisition data of the voltage stabilizer; and calculating the mass of liquid phase water of the loop and the volume of steam space of the voltage stabilizer after the related safety valve of the voltage stabilizer jumps.
2. The method according to claim 1, wherein in step 1, based on the formula:
Figure FDA0003060803510000011
calculating the steam space volume of the voltage stabilizer before the tripping of the safety valve of the voltage stabilizer; in the formula: p1、P2Respectively two instantaneous state pressures of the pressure stabilizer steam space in the time interval of the take-off of the safety valve, V1、V2The volume of the steam space of the pressure stabilizer, T, corresponding to two states respectively1,sat、T2,satThe saturation temperatures for the two states are respectively.
3. The method of claim 1, wherein in step 2, the enthalpy rise of the primary loop coolant is based on pressure and specific volume parameters of a point before and after expansion of the primary loop water, in combination with a query H2And calculating to obtain the water physical property table.
4. The method according to claim 1, wherein in step 3, t is2＝t1+ i × dt, i is a positive integer.
5. The method according to claim 1, wherein in step 4, based on the formula:
ΔM＝ΔVvap×ρvap_close
or
ΔM＝ΔVvap×ρvap_close+(ΔVliq-ΔVvap)×ρliq_close
Calculating the quality of the liquid phase water of the loop after the tripping of the safety valve of the voltage stabilizer;
based on the formula:
V2＝max((V1-ΔVliq),0.0)
calculating the steam space volume of the voltage stabilizer; in the formula: Δ M is the total loss of primary coolant, kg;
wherein, is Δ VvapVolume of primary loop steam and liquid phase water discharged from safety valve, unit m3；ΔVliqIs the expansion volume of liquid phase water in a loop, and has unit m3；ρvap_close、ρliq_closeThe densities of saturated steam and saturated water are respectively the recoil pressure of the safety valve; v1、V2The volume of the pressure stabilizer steam space before the pressure stabilizer safety valve takes off and after the pressure stabilizer safety valve returns to the seat respectively is unit m3。
6. The utility model provides a nuclear power plant's outage accident process rapid prediction system which characterized in that includes:
the steam space parameter calculation module of the voltage stabilizer comprises: the method is used for calculating the steam space volume of the pressure stabilizer before the safety valve of the pressure stabilizer jumps on the premise of not considering the mass exchange between the steam and the water in the pressure stabilizer, and obtaining the volume expansion amount of the liquid phase water of the primary loop;
a coolant enthalpy rise calculation module: the system is used for calculating the enthalpy rise of the coolant of a loop before the tripping of the safety valve of the voltage stabilizer;
the voltage stabilizer safety valve take-off time calculation module: the device is used for calculating the tripping time of the safety valve of the voltage stabilizer; assuming that the pressurizer safety valve is at t after the previous take-off1Returning to the seat at moment, and circularly calculating t1To t2The relative error between the integral value of the effective instantaneous thermal power of the reactor and the enthalpy rise of the coolant at the moment is smaller than the energyWhen the amount converges on the error, t2Namely the tripping time of the safety valve of the voltage stabilizer; the effective instantaneous thermal power of the reactor is the difference value of the instantaneous decay heat of the reactor and heat traps for heat dissipation of the first loop, heat conduction of the second loop and the like;
a loop liquid phase water quality and pressure stabilizer steam space volume calculation module: the system is used for updating the liquid phase water quality instant acquisition data of a loop and the space volume instant acquisition data of the voltage stabilizer; and calculating the mass of liquid phase water of the loop and the volume of steam space of the voltage stabilizer after the related safety valve of the voltage stabilizer jumps.
7. The system for rapidly predicting the progress of the nuclear power plant blackout accident according to claim 6, wherein the system is based on a formula:
Figure FDA0003060803510000021
calculating the steam space volume of the voltage stabilizer before the tripping of the safety valve of the voltage stabilizer; in the formula: p1、P2Respectively two instantaneous state pressures of the pressure stabilizer steam space in the time interval of the take-off of the safety valve, V1、V2The volume of the steam space of the pressure stabilizer, T, corresponding to two states respectively1,sat、T2,satThe saturation temperatures for the two states are respectively.
8. The system of claim 6, wherein the enthalpy rise of the primary loop coolant is based on the pressure and specific volume parameters of the front and rear points of the primary loop water expansion, and is combined with the query H2And calculating to obtain the water physical property table.
9. The system of claim 6, wherein t is a measure of the progress of the nuclear power plant outage event2＝t1+ i × dt, i is a positive integer; based on the formula:
ΔM＝ΔVvap×ρvap_close
or
ΔM＝ΔVvap×ρvap_close+(ΔVliq-ΔVvap)×ρliq_close
Calculating the quality of the liquid phase water of the loop after the tripping of the safety valve of the voltage stabilizer;
based on the formula:
V2＝max((V1-ΔVliq),0.0)
calculating the steam space volume of the voltage stabilizer; in the formula: Δ M is the total loss of primary coolant, kg;
wherein, is Δ VvapVolume of primary loop steam and liquid phase water discharged from safety valve, unit m3；ΔVliqIs the expansion volume of liquid phase water in a loop, and has unit m3；ρvap_close、ρliq_closeThe densities of saturated steam and saturated water are respectively the recoil pressure of the safety valve; v1、V2The volume of the pressure stabilizer steam space before the pressure stabilizer safety valve takes off and after the pressure stabilizer safety valve returns to the seat respectively is unit m3。
10. A computer-readable storage medium, characterized in that a computer program is stored, which, when being executed by a processor, is adapted to implement the method for rapid prediction of the progress of a nuclear power plant blackout accident according to any one of claims 1 to 5.
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