CN113239539B - Whole plant outage accident process prediction method, system and computer readable storage medium - Google Patents
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Abstract
The invention relates to the technical field of nuclear power plant accident prediction, in particular to a whole plant outage accident process prediction method, a whole plant outage accident process prediction system and a computer readable storage medium, which comprise the following steps: step 1: on the premise of not considering mass exchange between steam and water in the pressure stabilizer, calculating the volume of the steam space of the pressure stabilizer before the tripping of a safety valve of the pressure stabilizer, and obtaining the volume expansion of liquid phase water of a loop; step 2: before the tripping of a safety valve of the voltage stabilizer is calculated, the enthalpy of the coolant in a loop is increased; step 3: calculating the tripping time of a safety valve of the voltage stabilizer; 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 liquid phase water mass of a loop and the gas space volume of the voltage stabilizer after the safety valve of the voltage stabilizer is tripped; the method is used for solving the technical problem of how to rapidly predict the power failure of the whole nuclear power plant.
Description
Technical Field
The invention relates to the technical field of nuclear power plant accident prediction, in particular to a whole plant outage accident process prediction method, a whole plant outage accident process prediction system and a computer readable storage medium.
Background
The whole plant outage (SBO) accident is the most typical design benchmark accident of a pressurized water reactor nuclear power plant, and the accident process is generally that the whole plant of the nuclear power plant is powered off, a host computer is tripped, a reactor is shut down, a loop system is heated and boosted, a safety valve of a pressure stabilizer is cycled and tripped-a seat is depressurized, the water level of the pressure stabilizer and a pressure container is rapidly lowered, a reactor core is exposed, and a zirconium water reaction occurs on the surface of a fuel element, and the reactor core is fused. The Japanese Fudao nuclear accident is the whole plant outage accident. Therefore, the nuclear power plant is very concerned with SBO accident process simulation prediction and analysis work. Currently, the prediction analysis of the accident analysis research process of the nuclear power plants at home and abroad aiming at SBO and other design references is mainly based on the development of large-scale thermal hydraulic programs of the nuclear power plants such as RELAP5, MELCOR and MAAP 5. The large-scale thermal hydraulic program can accurately solve two-fluid (multiphase flow) energy, mass and momentum equations (two-fluid six-equation) of each pipeline or equipment on each time step by finely simulating the flow 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, by developing a targeted simulation analysis model or a simulation system aiming at a specific heap, the simulation analysis of design benchmark accidents such as SBO and the like is realized. The technical scheme has the following distinct characteristics: firstly, modeling is complex, 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 system fluid network, equipment structure and operation characteristics, nuclear reaction heat generation and heat transfer, material physical properties, instruments, control and the like of a nuclear power plant reactor, a primary loop system, a secondary loop system, special safety facilities and the like, and the model dimension is generally large. In general, a thermal hydraulic control body (including equipment components such as a pump valve) of a simulation model of a nuclear power plant is generally 100 nodes or more. When accident simulation is carried out, the energy, mass and momentum conservation equation matrix of multiple fluids is needed to be solved for all control bodies, the real-time or overtime simulation of the accident process is difficult to realize, and the interruption of the simulation process is often caused by the abnormality of the water property parameters such as the instantaneous temperature, pressure, gas content and the like of the control bodies. Secondly, the requirements on the professional ability of the user are very high, and the use difficulty of a loop operator is very high. Generally, accident analysis model developers not only need a certain expertise, but also can be proficient 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 needs to be further packaged into a simulator or a simulator to be put into use. At this time, the simulation analysis system often requires multiple operators to run smoothly. The operation difficulty of single person analysis is high.
Accordingly, it is desirable to develop a method, system, and computer-readable storage medium for predicting a process of a power outage incident in a whole plant, for solving at least one of the above-mentioned problems.
Disclosure of Invention
The invention provides a whole-plant outage accident process prediction method, a whole-plant outage accident process prediction system and a computer readable storage medium, aiming at solving the technical problem of how to rapidly predict whole-plant outage of a nuclear power plant.
In order to achieve the above purpose, the present invention adopts the following technical scheme:
in one aspect, the invention provides a method for rapidly predicting the process of a power failure accident of a whole nuclear power plant, which comprises the following operation steps:
step 1: on the premise of not considering mass exchange between steam and water in the pressure stabilizer, calculating the volume of the steam space of the pressure stabilizer before the tripping of a safety valve of the pressure stabilizer, and obtaining the volume expansion of liquid phase water of a loop;
step 2: before the safety valve of the voltage stabilizer is lifted, the enthalpy of the coolant of the first loop is calculated;
step 3: calculating the tripping time of the safety valve of the voltage stabilizer; suppose that the regulator relief valve was previously tripped at t 1 Returning to the seat at moment, and circularly calculating t 1 To t 2 Relative error of effective instantaneous thermal power integral value of reactor and enthalpy rise of coolant at moment, when the relative error is smaller than energy convergence error, t 2 The starting time of the safety valve of the voltage stabilizer is; the effective instantaneous thermal power of the reactor is the difference between the instantaneous decay heat of the reactor and the heat dissipation of the boundary of the primary loop and the heat conduction of the secondary loop;
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 liquid phase water mass of the loop and the vapor space volume of the voltage stabilizer after the safety valve of the voltage stabilizer is tripped.
Preferably, in the step 1, the formula is based on:calculating the vapor space volume of the voltage stabilizer before the tripping of the safety valve of the voltage stabilizer; wherein: p (P) 1 、P 2 Respectively two instantaneous state pressures of the gas space of the pressure stabilizer in the tripping interval time of the safety valve, V 1 、V 2 The volume, T, of the gas space of the voltage stabilizer respectively corresponding to the two states 1,sat 、T 2,sat The saturation temperatures corresponding to the two states respectively.
PreferablyIn the step 2, the enthalpy rise of the primary loop coolant is based on the pressure and specific volume parameters of the primary loop water before and after expansion, and is combined with the query H 2 And O is calculated after the water property table.
Preferably, in said step 3, t 2 ＝t 1 +i×dt, i is a positive integer.
Preferably, in the step 4, the formula is based on: Δm=Δvvap×ρvap_close or Δm=Δv vap ×ρ vap_close +(ΔV liq -ΔV vap )×ρ liq_close Calculating the liquid phase water quality of a loop after the safety valve of the voltage stabilizer is tripped; based on the formula: v (V) 2 ＝max((V 1 -ΔV liq ) 0.0) calculating the volume of the vapor space of the voltage stabilizer; wherein: Δm is the total loss of primary loop coolant, kg; deltaV vap Volume of primary circuit steam and liquid phase water discharged from safety valve, unit m 3 ；ΔV liq Is the expansion volume of liquid phase water of a loop and is in unit of m 3 ；ρ vap_close 、ρ liq_close The densities of saturated steam and saturated water are respectively the pressure of the seat returning of the safety valve; v (V) 1 、V 2 The unit m is the volume of the gas space of the voltage stabilizer before the tripping of the safety valve of the voltage stabilizer and after the seat returning 3 。
Preferably, the method further comprises the step 5: and calculating the liquid level change of the active area of the reactor core and the zirconium water reaction starting time according to the updated liquid phase water quality of the primary loop so as to further predict 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 process of the power failure accident of the whole nuclear power plant, which comprises the following steps:
the calculation module of the vapor space parameters of the voltage stabilizer: the method is used for calculating the steam space volume of the pressure stabilizer before the tripping of the safety valve of the pressure stabilizer and obtaining the volume expansion of liquid phase water of a loop on the premise of not considering the mass exchange between steam and water in the pressure stabilizer;
the coolant enthalpy rise calculation module: the method is used for calculating the enthalpy rise of the coolant in a loop before the safety valve of the voltage stabilizer is tripped;
the regulator relief valve take-off time calculation module: for reckoning the voltage stabilizerThe take-off time of the full valve; assume that the regulator relief valve was previously tripped at t 1 Returning to the seat at moment, and circularly calculating t 1 To t 2 Relative error of effective instantaneous thermal power integral value of reactor and enthalpy rise of coolant at moment, when the relative error is smaller than energy convergence error, t 2 The starting time of the safety valve of the voltage stabilizer is; the effective instantaneous thermal power of the reactor is the difference value between the instantaneous decay heat of the reactor and heat traps such as primary loop heat dissipation, secondary loop heat conduction and the like;
and the loop liquid phase water mass and pressure stabilizer vapor space volume calculation module is as follows: 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 liquid phase water mass of the loop and the vapor space volume of the voltage stabilizer after the safety valve of the voltage stabilizer is tripped.
Preferably, the formula is based on:
calculating the vapor space volume of the voltage stabilizer before the tripping of the safety valve of the voltage stabilizer; wherein: p (P) 1 、P 2 Respectively two instantaneous state pressures of the gas space of the pressure stabilizer in the tripping interval time of the safety valve, V 1 、V 2 The volume, T, of the gas space of the voltage stabilizer respectively corresponding to the two states 1,sat 、T 2,sat The saturation temperatures corresponding to the two states respectively.
Preferably, 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 inquiry H 2 And O is calculated after the water property table.
Preferably, t 2 ＝t 1 +i×dt, i is a positive integer; based on the formula:
ΔM＝ΔV vap ×ρ vap_close
or (b)
ΔM＝ΔV vap ×ρ vap_close +(ΔV liq -ΔV vap )×ρ liq_close
Calculating the liquid phase water quality of a loop after the safety valve of the voltage stabilizer is tripped;
based on the formula:
V 2 ＝max((V 1 -ΔV liq ),0.0)
calculating the volume of the gas space of the voltage stabilizer; wherein: Δm is the total loss of primary loop coolant, kg;
wherein DeltaV vap Volume of primary circuit steam and liquid phase water discharged from safety valve, unit m 3 ；ΔV liq Is the expansion volume of liquid phase water of a loop and is in unit of m 3 ；ρ vap_c l ose 、ρ liq_close The densities of saturated steam and saturated water are respectively the pressure of the seat returning of the safety valve; v (V) 1 、V 2 The unit m is the volume of the gas space of the voltage stabilizer before the tripping of the safety valve of the voltage stabilizer and after the seat returning 3 。
Furthermore, the invention also provides a computer readable storage medium storing a computer program which is used for realizing the rapid prediction method of the whole plant power failure accident progress of the nuclear power plant when being executed by a processor.
In summary, the beneficial effects of the technical proposal are as follows:
1. the method can solve the problems of large difficulty in real-time or overtime analysis and prediction of SBO accident analysis and easy interruption of a computing process in the prior art scheme, and realize the ultra-real-time stable analysis of the SBO accident process of the nuclear power plant; furthermore, the invention can also reduce the technical difficulty of analysis of SBO accidents of the nuclear power plant and meet the use requirements of common operation operators of the nuclear power plant; in addition, the invention introduces a new technical method for analyzing the SBO accident of the nuclear power plant, and can be used for carrying out contrast verification with the prior art scheme.
2. According to the method provided by the invention, an analysis model is built by only needing a small amount of accident parameters and macroscopic design parameters of the nuclear power plant, and the design parameters related to the size structure and the flow heat exchange of the system equipment are not required to be accurately mastered. Modeling data requirements are about 10% of the best existing solutions. The labor cost of modeling work is about 1% of the existing best technical scheme.
3. When the method provided by the invention is used for analyzing the model to predict the accident progress, only the mass and energy equations are required to be continuously searched progressively along the time scale, iterative calculation is not required to be carried out on parameters such as two fluids, six equations and the like on each time step, and the calculation time can be greatly reduced. When SBO accident process simulation with the total time scale of 24 hours is carried out, and the single-core CPU with the main frequency of 2.0GHz is operated, the accident process prediction time is generally 1-100 seconds, and the accident process prediction time of the best technical scheme at present is generally 1-12 hours.
Drawings
FIG. 1 is a schematic diagram of an analysis flow of SBO accident processes of a safety valve action cycle of a voltage regulator according to an embodiment of the present invention;
Detailed Description
The following description of the embodiments of the present invention will be made clearly and completely with reference to the accompanying drawings, in which it is apparent that the embodiments described are only some embodiments of the present invention, but not all embodiments. All other embodiments, which can be made by those skilled in the art based on the embodiments of the invention without making any inventive effort, are intended to be within the scope of the invention.
Embodiment one:
the technical principle involved in the method for rapidly predicting the process of the power failure accident of the whole nuclear power plant in the embodiment is as follows:
dividing a loop coolant into two parts of vapor space steam of a pressure stabilizer and liquid phase water of the loop, and respectively establishing the following equations:
1. based on an ideal gas state equation, a voltage stabilizer steam space state approximation equation (formula (1)) before the safety valve is opened is established to solve the volume-pressure change of the voltage stabilizer steam, and further solve the volume-pressure change of the coolant of a loop.
2. Based on energy conservation, a relation (formula (2)) of heat loss and coolant enthalpy rise, such as primary loop accumulated decay heat, secondary loop heat conduction, primary loop heat dissipation and the like, is established.
3. Based on the ideal gas state equation, an ideal gas state equation (formula (3) and formula (4)) of two states of a tripping and a returning seat of a pressure stabilizer safety valve is established, and the discharge amount (formula (5) and formula (6)) of a returning coolant and a new pressure stabilizer steam space volume (formula (7)) can be solved; based on the conservation of mass and energy, the mass of the remaining coolant and its enthalpy of the primary loop can be solved (equations (8) and (9)).
The process is the SBO accident process analysis process of the action cycle of the safety valve of the voltage stabilizer. A flow chart of accident analysis is designed accordingly, see fig. 1.
Wherein, the calculation relation of the formulas (1) to (9) is described as follows:
and a voltage stabilizer steam space state equation established based on an ideal gas state equation:
wherein: p (P) 1 、P 2 Respectively two instantaneous state pressures of the gas space of the pressure stabilizer in the tripping interval time of the safety valve, V 1 、V 2 The volume, T, of the gas space of the voltage stabilizer respectively corresponding to the two states 1,sat 、T 2,sat The saturation temperatures corresponding to the two states respectively.
The heat absorbed by the enthalpy rise of the coolant in one circuit is calculated as follows:
wherein: t is t 0 、t 1 The time of occurrence of the two states, s; ΔH is t 0 To t 1 Enthalpy difference, J/kg, of the coolant of the first circuit at the moment; q (t) is loop transient decay heat, w; s (t) is the power of the heat trap of the two loops, i.e. the heat conduction of the two loops, the heat dissipation of the one loop, and the like.
When the volume expansion of the primary loop coolant is not considered, the jump of the primary safety valve and the transient state of the pressure stabilizer after returning to the seat are expressed as follows:
the correction relation after the expansion of the volume of the coolant of the primary circuit is considered as follows:
wherein: p (P) open 、P close The pressure of the safety valve is respectively the lifting and returning seat pressure of the safety valve, and the unit Pa; v (V) open The volume of the gas space of the pressure stabilizer is unit m when the safety valve is lifted 3 ；ΔV vap The volume expansion amount of the pressure stabilizer gas space at the tripping pressure and the seat returning pressure of the safety valve is unit m 3 ；T sat,open 、T sat,close The water saturation temperature corresponding to the tripping and seat returning pressure of the safety valve is respectively set. DeltaV liq For the relief valve to jump to the two pressure states of the seat, the volume expansion of the coolant in the loop is as unit m 3 。
When DeltaV liq Less than DeltaV vap The mass of the primary circuit coolant discharged by the primary safety valve action is as follows:
ΔM＝ΔV vap ×ρ vap_close (5)
When DeltaV liq Greater than DeltaV vap The primary safety valve action-loop lost coolant mass is as follows:
ΔM＝ΔV vap ×ρ vap_close +(ΔV liq -ΔV vap )×ρ liq_close (6)
Wherein: Δm is the total loss of primary loop coolant, kg; deltaV vap Volume of primary circuit steam and liquid phase water discharged from safety valve, unit m 3 ；ΔV liq Is the expansion volume of liquid phase water of a loop and is in unit of m 3 ；ρ vap_close 、ρ liq_close The densities of saturated steam and saturated water are respectively the pressure of the seat returning of the safety valve.
After the primary safety valve is tripped and returns to the seat, the volume of the steam space of the pressure stabilizer is as follows:
V 2 ＝max((V 1 -ΔV liq ) 0.0) formula (7)
After the primary safety valve is lifted-returned to the seat, the mass of the liquid phase water of the primary loop is as follows:
M 2 ＝M 1 -max((ΔV liq -ΔV vap )×ρ liq_close ) 0.0) type (8)
After the primary relief valve is tripped and returned to the seat, the enthalpy of the liquid phase water of the primary loop is as follows:
wherein: v (V) 1 、V 2 The unit m is the volume of the gas space of the voltage stabilizer before the tripping of the safety valve of the voltage stabilizer and after the seat returning 3 ；M 1 、M 2 、H 1 、H 2 The mass and enthalpy of liquid phase water of a loop before the tripping of a safety valve of the voltage stabilizer and after the tripping of a loop seat are respectively in m units 3 And J/kg.
In this embodiment, the SBO accident process analysis model requires 15 classes of input (hypothetical) parameters:
1. macroscopic design (operating) parameters of a nuclear power plant:
(1) Water content of a loop system;
(2) The safety valve of the pressure stabilizer takes off, returns the seat pressure and rated displacement;
(3) A regulator volume;
(4) Uranium loading of the reactor;
(5) A loop average temperature-heat dissipation power meter.
2. Accident-related parameters:
(6) The pre-accident circuit pressure;
(7) Reactor inlet and outlet temperatures before an accident;
(8) The volume of the water of the voltage stabilizer before an accident;
(9) Reactor power operation history (power and time) before accident;
(10) Reactor shutdown time;
(11) The power-off time of the whole plant;
(12) The steam flow-time table for the two loops after the whole plant is powered off;
(13) And (5) a two-loop pressure curve after the whole plant is powered off.
3. Analysis of hypothesis-related parameters
(14) A time step;
(15) Allowable convergence errors, including energy convergence errors and quality convergence errors;
the specific analysis flow of this embodiment is described in detail as follows:
assuming that the core flow of the SBO incident process analysis of the nuclear power plant of this patent includes 4 stages, as shown in FIG. 1:
1. calculating the volume of the vapor space of the pressure stabilizer before the tripping of the safety valve:
the volume of the vapor space of the pressure stabilizer before the tripping of the safety valve is calculated based on the formula (1) without considering the mass exchange between vapor and water in the pressure stabilizer, so that the volume expansion of liquid phase water of a loop can be calculated.
2. Calculating the enthalpy rise of the coolant of the primary loop before the jump of the safety valve:
the enthalpy rise of the returned coolant can be calculated by inquiring the H2O water property table based on the (pressure and specific volume) parameters of the front and rear points of the primary loop water expansion.
3. Calculating the tripping time of the safety valve:
suppose that the regulator relief valve was previously tripped at t 1 Returning to the seat at moment, and circularly calculating t 1 To t 2 (t 2 ＝t 1 +i x dt, i=1,..n) relative error of the reactor effective instantaneous thermal power integrated value and the coolant enthalpy rise, when the relative error is smaller than the energy convergence error, t 2 The trip time of the safety valve is obtained.
The effective instantaneous thermal power of the reactor is the difference between the instantaneous decay heat of the reactor and the heat traps such as primary heat dissipation and secondary loop heat conduction.
4. Updating the liquid phase water quality of the primary loop and the vapor space volume of the voltage stabilizer;
and calculating the liquid phase water mass of the first loop after the tripping of the safety valve of the voltage stabilizer based on the formula (5) or the formula (6). The volume of the stabilizer vapor space is calculated based on equation (7).
For a specific nuclear power plant, parameters such as the liquid level of the reactor core active area can be calculated according to the updated liquid phase water quality of the primary loop.
Example 2:
the embodiment provides a method for rapidly predicting the process of a power failure accident of a whole nuclear power plant on the basis of embodiment 1, which can rapidly predict the power failure of the whole nuclear power plant, including but not limited to; the method comprises the following operation steps:
step 1: on the premise of not considering mass exchange between steam and water in the pressure stabilizer, calculating the volume of the steam space of the pressure stabilizer before the tripping of the safety valve, and obtaining the volume expansion of liquid phase water of a loop;
step 2: calculating the enthalpy rise of the coolant of the first loop before the tripping of the safety valve;
step 3: calculating the tripping time of the safety valve; suppose that the regulator relief valve was previously tripped at t 1 Returning to the seat at moment, and circularly calculating t 1 To t 2 Relative error of effective instantaneous thermal power integral value of reactor and enthalpy rise of coolant at moment, when the relative error is smaller than energy convergence error, t 2 Namely 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 dissipation of the boundary of the primary loop and the heat conduction of the secondary loop;
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 liquid phase water mass of the loop and the vapor space volume of the voltage stabilizer after the safety valve of the voltage stabilizer is tripped.
Embodiment III:
the embodiment further provides a rapid prediction system for the process of the power failure accident of the whole nuclear power plant on the basis of embodiment 2, which comprises:
the calculation module of the vapor space parameters of the voltage stabilizer: the method is used for calculating the steam space volume of the pressure stabilizer before the tripping of the safety valve of the pressure stabilizer and obtaining the volume expansion of liquid phase water of a loop on the premise of not considering the mass exchange between steam and water in the pressure stabilizer;
the coolant enthalpy rise calculation module: the method is used for calculating the enthalpy rise of the coolant in a loop before the safety valve of the voltage stabilizer is tripped;
the regulator relief valve take-off time calculation module: the method is used for calculating the tripping time of the safety valve of the voltage stabilizer; assume that the regulator relief valve was previously tripped at t 1 Returning to the seat at moment, and circularly calculating t 1 To t 2 Relative error of effective instantaneous thermal power integral value of reactor and enthalpy rise of coolant at moment, when the relative error is smaller than energy convergence error, t 2 The starting time of the safety valve of the voltage stabilizer is; the effective instantaneous thermal power of the reactor is the difference value between the instantaneous decay heat of the reactor and heat traps such as primary loop heat dissipation, secondary loop heat conduction and the like;
and the loop liquid phase water mass and pressure stabilizer vapor space volume calculation module is as follows: 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 liquid phase water mass of the loop and the vapor space volume of the voltage stabilizer after the safety valve of the voltage stabilizer is tripped.
Embodiment four:
the present embodiment provides a computer-readable storage medium storing a computer program for implementing the method in the foregoing embodiment when executed by a processor.
In summary, the beneficial effects of the technical proposal are as follows:
1. the method can solve the problems of large difficulty in real-time or overtime analysis and prediction of SBO accident analysis and easy interruption of a computing process in the prior art scheme, and realize the ultra-real-time stable analysis of the SBO accident process of the nuclear power plant; furthermore, the invention can also reduce the technical difficulty of analysis of SBO accidents of the nuclear power plant and meet the use requirements of common operation operators of the nuclear power plant; in addition, the invention introduces a new technical method for analyzing the SBO accident of the nuclear power plant, and can be used for carrying out contrast verification with the prior art scheme.
2. According to the method provided by the invention, an analysis model is built by only needing a small amount of accident parameters and macroscopic design parameters of the nuclear power plant, and the design parameters related to the size structure and the flow heat exchange of the system equipment are not required to be accurately mastered. Modeling data requirements are about 10% of the best existing solutions. The labor cost of modeling work is about 1% of the existing best technical scheme.
3. When the method provided by the invention is used for analyzing the model to predict the accident progress, only the mass and energy equations are required to be continuously searched progressively along the time scale, iterative calculation is not required to be carried out on parameters such as two fluids, six equations and the like on each time step, and the calculation time can be greatly reduced. When SBO accident process simulation with the total time scale of 24 hours is carried out, and the single-core CPU with the main frequency of 2.0GHz is operated, the accident process prediction time is generally 1-100 seconds, and the accident process prediction time of the best technical scheme at present is generally 1-12 hours.
The foregoing is only a preferred embodiment of the present invention, but the scope of the present invention is not limited thereto, and any person skilled in the art, who is within the scope of the present invention, should make equivalent substitutions or modifications according to the technical scheme of the present invention and the inventive concept thereof, and should be covered by the scope of the present invention.
Claims (3)
1. The method for rapidly predicting the process of the power failure accident of the whole nuclear power plant is characterized by comprising the following operation steps:
step 1: on the premise of not considering mass exchange between steam and water in the pressure stabilizer, calculating the volume of the steam space of the pressure stabilizer before the tripping of a safety valve of the pressure stabilizer, and obtaining the volume expansion of liquid phase water of a loop;
step 2: before the safety valve of the voltage stabilizer is lifted, the enthalpy of the coolant in a loop is calculated;
step 3: calculating the tripping time of the safety valve of the voltage stabilizer; assuming that the safety valve of the voltage stabilizer returns to a seat at the moment t1 after the previous jump, circularly calculating the relative error of the effective instantaneous thermal power integral value of the reactor and the enthalpy rise of the coolant at the moment t1 to the moment t2, and when the relative error is smaller than the energy convergence error, t2 is the jump time of the safety valve of the voltage stabilizer; the effective instantaneous thermal power of the reactor is the difference between the instantaneous decay heat of the reactor and the heat dissipation of the boundary of the primary loop and the heat conduction of the secondary loop;
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 liquid phase water mass of a loop and the gas space volume of the voltage stabilizer after the safety valve of the voltage stabilizer is tripped;
in the step 1, based on the formula:
；
calculating the vapor space volume of the voltage stabilizer before the tripping of the safety valve of the voltage stabilizer; wherein: p1 and P2 are respectively the pressure of two instantaneous states of the gas space of the pressure stabilizer in the time interval of the jump of the safety valve, T1, sat and T2, sat are respectively the saturation temperatures corresponding to the two states;
in the step 2, the enthalpy rise of the primary loop coolant is calculated based on the pressure and specific volume parameters of the primary loop water before and after expansion, and by combining with inquiring the H2O water property table;
in said step 4, based on the formula:
or->；
Calculating the liquid phase water quality of a loop after the safety valve of the voltage stabilizer is tripped;
based on the formula:
；
calculating the volume of the gas space of the voltage stabilizer; wherein: Δm is the total loss of primary loop coolant, kg;
wherein DeltaV vap Volume of primary circuit steam and liquid phase water discharged from safety valve, unit m 3 ；ΔV liq Is the expansion volume of liquid phase water of a loop and is in unit of m 3 ；ρ vap_close 、ρ liq_close The densities of saturated steam and saturated water are respectively the pressure of the seat returning of the safety valve; v (V) 1 、V 2 The unit m is the volume of the gas space of the voltage stabilizer before the tripping of the safety valve of the voltage stabilizer and after the seat returning 3 。
2. The utility model provides a quick prediction system of nuclear power plant whole plant outage accident process which characterized in that includes: the calculation module of the vapor space parameters of the voltage stabilizer: for irrespective of the mass between steam and water in the stabiliser
On the premise of exchange, calculating the volume of the steam space of the voltage stabilizer before the safety valve of the voltage stabilizer jumps, and obtaining the volume expansion quantity of liquid phase water of a loop;
the coolant enthalpy rise calculation module: the method is used for calculating the enthalpy rise of the coolant in a loop before the safety valve of the voltage stabilizer is tripped;
the regulator relief valve take-off time calculation module: the method is used for calculating the tripping time of the safety valve of the voltage stabilizer; assuming that the safety valve of the voltage stabilizer returns to a seat at the moment t1 after the previous jump, circularly calculating the relative error of the effective instantaneous thermal power integral value of the reactor and the enthalpy rise of the coolant at the moment t1 to the moment t2, and when the relative error is smaller than the energy convergence error, t2 is the jump time of the safety valve of the voltage stabilizer; the effective instantaneous thermal power of the reactor is the difference between the instantaneous decay heat of the reactor and the primary loop heat dissipation and secondary loop heat conduction heat sink;
and the loop liquid phase water mass and pressure stabilizer vapor space volume calculation module is as follows: 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; calculating the liquid phase water mass of a loop and the gas space volume of the voltage stabilizer after the safety valve of the voltage stabilizer is tripped;
based on the formula:
；
calculating the vapor space volume of the voltage stabilizer before the tripping of the safety valve of the voltage stabilizer; wherein: p1 and P2 are respectively the pressure of two instantaneous states of the gas space of the pressure stabilizer in the time interval of the jump of the safety valve, T1, sat and T2, sat are respectively the saturation temperatures corresponding to the two states;
the enthalpy rise of the primary loop coolant is calculated based on pressure and specific volume parameters of the primary loop water before and after expansion and by combining with the inquiry of an H2O water property table;
based on the formula:or->Calculating the liquid phase water quality of a loop after the safety valve of the voltage stabilizer is tripped;
based on the formula:；
calculating the volume of the gas space of the voltage stabilizer; wherein: d (D)M The total loss of the primary loop coolant is kg;
wherein DeltaV vap Volume of primary circuit steam and liquid phase water discharged from safety valve, unit m 3 ；ΔV liq Is the expansion volume of liquid phase water of a loop and is in unit of m 3 ；ρ vap_close 、ρ liq_close The densities of saturated steam and saturated water are respectively the pressure of the seat returning of the safety valve; v (V) 1 、V 2 The unit m is the volume of the gas space of the voltage stabilizer before the tripping of the safety valve of the voltage stabilizer and after the seat returning 3 。
3. 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 fast predicting a process of a plant-wide outage event in a nuclear power plant as set forth in claim 1 or 2.
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