Nuclear reactor coolant loss accident reactor core key parameter analysis method

Abstract

The invention discloses a nuclear reactor coolant loss accident reactor core key parameter analysis method. The method comprises the following specific steps: 1, determining initial factors such as an accident analysis model modeling range and a crevasse position and size when a pressurized water reactor has a coolant loss accident; 2, establishing an accident analysis model based on a nuclear reactor system analysis program; 3, carrying out calculation on a full-power operation steady-state working condition to obtain key operation parameters; 4, sampling and establishing an initial transient condition example set required by Monte Carlo analysis; 5, solving key thermal hydraulic parameter changes under the typical pressurized water reactor coolant loss accident condition; and 6, all examples are calculated, and the optimal estimated values of the liquid level of the reactor core, the highest temperature of the fuel cladding and the pressure in the containment in the coolant loss accident are obtained through statistics. According to the method disclosed by the invention, the optimal estimated value calculation of the reactor core liquid level, the highest temperature of the fuel cladding and the containment pressure under the nuclear reactor coolant loss accident is completed by adopting a nuclear reactor system analysis program under a Monte Carlo theory framework.

Description

Technical Field

[0001] The present invention relates to the technical field of research on the calculation of optimal estimated values ​​of key parameters that affect the safety of a pressurized water reactor nuclear power plant, such as a drop in core liquid level, an increase in fuel cladding temperature, and an increase in containment pressure, after a design basis accident such as a loss of coolant occurs in a pressurized water reactor nuclear power plant. The present invention specifically relates to a method for analyzing key core parameters of a nuclear reaction loss of coolant accident. Background Art

[0002] At present, 303 of the 412 nuclear power units in operation worldwide use pressurized water reactor technology. At the same time, pressurized water reactor technology also dominates the 58 nuclear power units under construction, reaching 49, and the proportion of both exceeds 70%. This situation highlights the important position of pressurized water reactor nuclear power technology in the field of nuclear power generation. In a pressurized water reactor nuclear power plant, the coolant continuously removes the heat of the core to the steam generator to generate steam to drive the steam turbine to generate power. When a nuclear power plant fails and the coolant pipeline breaks, causing coolant leakage, the core liquid level gradually decreases, and the temperature of the fuel cladding continues to rise due to insufficient cooling. In addition, the leaked high-temperature and high-pressure coolant enters the containment, causing the pressure and temperature in the containment to gradually increase, which poses a serious challenge to the safety of the reactor; if the accident develops further, the fuel cladding and the coolant react under high temperature conditions to produce hydrogen, thereby generating the same hydrogen explosion risk as the Fukushima nuclear accident in Japan. In addition, the high-temperature and high-pressure water vapor in the containment will seriously threaten the integrity of the containment. Therefore, the key parameters of the nuclear power plant during the accident are of great significance to the safety assessment of the nuclear power plant, and also play an important role in accident mitigation. At present, the calculation of key parameters mainly relies on the nuclear reactor system analysis program, but the results based on system analysis and calculation can only reflect a specific operating condition and do not have high universality. When taking measures for accidents in nuclear power plants, the uncertainty of initial conditions cannot be considered, and it is difficult to obtain effective thermal parameter values ​​as a reference for the selection of accident emergency strategies. In summary, it is necessary to conduct research on the calculation and analysis method of the best estimated values ​​of key parameters when a loss of coolant accident occurs in a pressurized water reactor nuclear power plant. This method will be able to provide effective tools and technical support for the safety analysis of nuclear power plants in my country, assist in the design and risk analysis of nuclear power plants, and provide a reference for the prevention and mitigation of nuclear power plant accidents. Summary of the invention

[0003] In order to overcome the problems existing in the above-mentioned prior art, the purpose of the present invention is to provide a method for analyzing key core parameters of a nuclear reactor loss of coolant accident, which can calculate the best estimated values ​​of the core liquid level, the maximum temperature of the fuel cladding, and the containment pressure under a nuclear reactor loss of coolant accident based on the idea of ​​combining probability theory and determinism.

[0004] In order to achieve the above object, the present invention adopts the following technical solution:

[0005] A method for analyzing key core parameters of a nuclear reactor loss of coolant accident comprises the following steps:

[0006] Step 1: Determine the scope of the accident analysis model and the accident process; Carry out the analysis and calculation of the loss of coolant accident for a typical pressurized water reactor nuclear power plant, first determine the accident process and the response of each system in the nuclear power plant at different stages, and establish the selection principle of the nuclear power plant systems that need to be covered by the accident analysis model: It is of great significance to the safety of nuclear power plants and accident prevention and mitigation;

[0007] Since the four key equipment in the reactor coolant primary and secondary loops, namely the reactor pressure vessel, steam generator, pressurizer, and reactor coolant pump, play an irreplaceable role in the reading and determination of key parameters of nuclear power plants, and the safety injection system, auxiliary water supply system, steam safety valve, and bypass circuit have an important impact on the process of coolant loss accidents, based on the established selection principles, the above key equipment and systems are covered, analyzed, and simplified during the construction of the accident analysis model;

[0008] In a nuclear power plant coolant loss accident, the location and size of the coolant pipeline rupture are significantly more relevant to the nuclear power plant operating status and accident response strategy than other initial state parameters. This is analyzed to determine the sampling range of this parameter and provide an accident starting point for accident analysis and calculation.

[0009] Step 2: Construct an accident analysis model based on the nuclear reactor system analysis program; based on the modeling scope analysis completed in step 1 and the call of each system of the nuclear power plant during the accident process, construct an accident analysis model for the pressurized water reactor nuclear power plant; in the modeling of the reactor coolant system, keep the number of equipment, flow area, flow channel length, and heat exchange area the same as the key parameters in the real equipment, and partially simplify the modeling of the four key equipments of the reactor pressure vessel, steam generator, pressurizer and reactor coolant pump to meet the needs of reading key parameters of the nuclear power plant; at the same time, the modeling of the auxiliary system and safety system that have an important impact on the accident process is also A simplified method is adopted, and the specific operations are as follows: in the modeling process of the auxiliary water supply system, the steam-driven auxiliary water supply pump and the electric auxiliary water supply pump are considered together, and the relevant pipelines and valves are simplified while retaining two auxiliary water supply pipelines and the water supply ports to the three steam generators. In the modeling process of the safety injection system, both the medium-pressure safety injection system and the low-pressure safety injection system are modeled, and the safety injection box is modeled using exclusive components to realize the non-passive safety injection function. The relevant model can realize the simulation of activating different types of circuit equipment in response to different types of accidents during the accident process. In addition, the safety valves and connecting pipelines of the regulator and steam generator are also arranged accordingly;

[0010] For the coolant loss accident starting point, i.e., the coolant pipeline rupture, the model uses a valve with the same equivalent diameter as the rupture to simulate it. The valve is connected to a control body with a volume equal to the total volume in the containment to achieve synchronous simulation of the coolant discharge process into the containment.

[0011] Step 3: Steady-state calculation of accident analysis model and acquisition of key operating parameters; Full power is the normal operating state with the highest thermal power of the reactor. At this time, the heat load and flow rate of the coolant are at the maximum level of normal operation. This state is selected as the initial condition to ensure that the impact of the coolant loss accident is evaluated under the most unfavorable thermal conditions. At the same time, the full power operating state is selected to carry out safety analysis to ensure the conservatism of the analysis results, so as to ensure that the safety system and containment of the nuclear power plant can meet the design requirements under the most stringent working conditions;

[0012] Based on the built typical PWR nuclear power plant accident analysis model, steady-state calculations at full power are carried out. The calculations are considered to be convergent when the coolant mass flow rate, core coolant inlet and outlet temperatures, main steam mass flow rate and pressure change with time by no more than 0.1% per hour. After that, the above parameters are compared with the design values ​​of the nuclear power plant. When the errors between the steady-state simulation values ​​and the design values ​​of the nuclear power plant are within 1%, the steady-state calculation results are considered to meet the transient analysis calculation requirements.

[0013] Step 4: Sampling is carried out based on the Holden sequence to establish a Monte Carlo simulation case set; according to the Wilkes theory, the number of safe calculations is confirmed to be 93 times using a bilateral 95% confidence interval, so 93 cumulative probability points need to be obtained. The Holden sequence is used to obtain the cumulative probability points between [0,1]. Through the probability distribution curve of the coolant pipeline rupture location and rupture size, the cumulative probability points are mapped to obtain the coolant pipeline rupture location and rupture size values, and the Monte Carlo simulation case set is obtained by random combination. For the bilateral tolerance interval, the Wilkes formula is:

[0014] β=1-γ N -N(1-γ)γ N-1 Formula (1)

[0015] In the formula

[0016] β——confidence;

[0017] γ——probability share;

[0018] N——number of calculations;

[0019] Step 5: Solve the key thermal-hydraulic parameters under the condition of loss of coolant in the pressurized water reactor nuclear power plant; for the accident analysis model constructed for the pressurized water reactor nuclear power plant, the nuclear reactor system analysis program is used to calculate and solve, and the time-varying characteristics of the key parameters under steady-state and transient conditions are obtained; the nuclear reactor system analysis program uses a two-fluid model to accurately simulate the thermal-hydraulic phenomena of the reactor system, especially the calculation of the rupture accident and the steam generator heat transfer tube rupture accident, which are violent phase change phenomena in the reactor, and realizes the accurate simulation of the phenomena and key parameters when various design basis accidents occur in the reactor;

[0020] The nuclear reactor system analysis program is based on a two-fluid, six-equation analysis model, and establishes three conservation equations of mass, momentum, and energy for the gas phase and liquid phase respectively as follows:

[0021] The mass conservation equation:

[0022]

[0023] The specific physical meanings of each item in the formula are as follows:

[0024] α g ——Volume fraction of vapor phase, %;

[0025] ρ g ——Fluid density of vapor phase, kg / m 3 ;

[0026] ν g ——fluid velocity of vapor phase, m / s;

[0027] Γ g ——Source term of vapor phase, kg / m 3 ·s;

[0028] A——circulation area, m 2 ;

[0029] t——time, s;

[0030] α f ——volume fraction of liquid phase, %;

[0031] ρ f ——Fluid density of liquid phase, kg / m 3 ;

[0032] ν f ——fluid velocity of liquid phase, m / s;

[0033] Γ f ——Source term of liquid phase, kg / m 3 ·s;

[0034] Momentum conservation equation:

[0035]

[0036] The specific physical meanings of each item in the formula are as follows:

[0037] α g ——Volume fraction of vapor phase, %;

[0038] ρ g ——Fluid density of vapor phase, kg / m 3 ;

[0039] ν g ——fluid velocity of vapor phase, m / s;

[0040] ν gi ——Initial fluid velocity of vapor phase, m / s;

[0041] Γ g ——Source term of vapor phase, kg / m 3 ·s;

[0042] V g ——Volume of vapor phase, m 3 ;

[0043] α f ——volume fraction of liquid phase, %;

[0044] ρ f ——Fluid density of liquid phase, kg / m 3 ;

[0045] ν f ——fluid velocity of liquid phase, m / s;

[0046] ν fi ——Initial fluid velocity of liquid phase, m / s;

[0047] ρ m ——Density of the mixture, kg / m 3 ;

[0048] Γ f ——Source term of liquid phase, kg / m 3 ·s;

[0049] V f ——volume of liquid phase, m 3 ;

[0050] A——circulation area, m 2 ;

[0051] C - interaction coefficient between liquid and vapor phases;

[0052] P——fluid pressure, Pa;

[0053] B x ——Volume force coefficient of the fluid in the x direction, m / s 2 ;

[0054] FWG – virtual mass force coefficient of vapor or liquid phase;

[0055] t——time, s;

[0056] Energy conservation equation:

[0057]

[0058] The specific physical meanings of each item in the formula are as follows:

[0059] α g ——Volume fraction of vapor phase, %;

[0060] ρ g ——Fluid density of vapor phase, kg / m 3 ;

[0061] U g ——Specific internal energy of vapor phase, J / kg;

[0062] ν g ——fluid velocity of vapor phase, m / s;

[0063] Q wg ——heat conduction rate between vapor phase and wall, J / s;

[0064] Q ig ——heat transfer rate between vapor phase and other items, J / s;

[0065] ——Enthalpy value of vapor phase change, J / kg;

[0066] DISS g ——vapor phase energy dissipation function;

[0067] h g ——Specific enthalpy of vapor phase, J / kg;

[0068] V g ——Volume of vapor phase, m 3 ;

[0069] α f ——volume fraction of liquid phase, %;

[0070] ρ f ——Fluid density of liquid phase, kg / m 3 ;

[0071] U f ——Specific internal energy of liquid phase, J / kg;

[0072] ν f ——fluid velocity of liquid phase, m / s;

[0073] Q wf ——heat conduction rate between liquid phase and wall, J / s;

[0074] Q if ——heat transfer rate between the liquid phase and other items, J / s;

[0075] DISS f ——Liquid phase energy dissipation function;

[0076] ——Enthalpy value of liquid phase change, J / kg;

[0077] h g ——Specific enthalpy of vapor phase, J / kg;

[0078] Γ ig ——Mass transfer rate between liquid phase and vapor phase, kg / m 3 ·s;

[0079] Γ w ——Mass transfer rate between liquid phase and wall, kg / m 3 ·s;

[0080] A——circulation area, m 2 ;

[0081] P——fluid pressure, Pa;

[0082] t——time, s;

[0083] Semi-implicit difference is used to discretize the conservation equations of mass, momentum and energy, and numerical calculations are carried out based on the discretized equations; four flow types are considered for water working fluid, including horizontal flow, vertical flow in tubular, annular and rod-shaped areas, mixed flow in the pump and mixed flow in the injection area of ​​the emergency core cooling system; at the same time, critical flow model, horizontal stratified flow model, flow channel cross-section mutation model, gas-liquid two-phase critical flow flow model, thermal stratification effect model, lateral flow model, non-condensable gas model and air gap heat conduction model are added to realize auxiliary calculation; a point reactor dynamics model is used for the core power of the nuclear reactor, and the reactor power is assumed to be an equation of time and space. In the point reactor dynamics model, it is assumed that the core power distribution is close to a constant;

[0084] Step 6: Obtain key parameters of the coolant loss accident; For the coolant loss accident of a nuclear reactor, the description of the extreme accident state is: a coolant main pipe undergoes brittle fracture and the pipe is completely disconnected and dislocated; in this state, the coolant is simultaneously ejected from the two disconnected ports, that is, from openings equivalent to twice the cross-section of the main pipe; in the large-break coolant loss accident, the continuous loss of coolant exposes the core, reduces the heat transfer capacity, and damages the fuel elements. At the same time, the zirconium cladding of the fuel elements will undergo the following series of violent chemical reactions with water vapor at high temperatures:

[0085] Zr+2H 2 0=ZrO 2 +2H 2 +6.51[MJ / kg(Zr)] Formula (8)

[0086] In addition, high-temperature and high-pressure coolant is sprayed into the containment space, which increases the gas pressure and temperature in the containment, endangering the integrity of the containment. Based on the key results of the above accident process evolution, three parameters, namely, reactor core liquid level, fuel cladding maximum temperature, and pressure in the containment, are selected as key parameters that directly reflect whether the nuclear power plant is safe. The relatively more frequent medium-break coolant loss and small-break coolant loss in nuclear power plants cause similar accident consequences to large-break coolant loss, but their severity is relatively lower. Therefore, the same key parameters as large-break coolant loss are used to evaluate the accident consequences. The sequence of events in a coolant loss accident is defined as four consecutive stages, namely, blowdown, refilling, reflooding and long-term cooling. When the reactor undergoes the reflooding stage, the temperature of the fuel core and the cladding can be kept within a controllable range. The reactor can then be kept in a safe state by deriving decay heat through long-term cooling. The accident calculation reaches the core reflooding as the calculation completion condition. For the extremely small break coolant loss condition where the core is not exposed, the moment when the pressure of the first and second circuits of the reactor are consistent is selected as the calculation termination time. At this time, the reactor decay heat is continuously removed through the steam generator to achieve long-term cooling.

[0087] Step 7: Statistical analysis is performed to obtain the best estimate of the key parameters of the core of the nuclear reactor in the event of a loss of coolant accident; the number of safe calculations is confirmed to be 93 according to the double-sided 95% confidence interval, and the corresponding core water level {h 1 ,h 2 ,……,h 92 ,h 93}、Maximum fuel cladding temperature {t 1 ,t 2 ,……,t 92 ,t 93}、Maximum pressure in containment {p 1 ,p 2,……,p 92 ,p 93}, perform statistical analysis on these parameters, calculate the best estimated value of each parameter, and use formula (9) to calculate the best estimated value;

[0088]

[0089] In the formula

[0090] - best estimates of key parameters of pressurized water reactor nuclear power plants;

[0091] x i ——The key parameter values ​​of the pressurized water reactor nuclear power plant calculated in the i-th example.

[0092] Compared with the prior art, the present invention has the following advantages:

[0093] The method of the present invention is developed based on a nuclear reactor system analysis program, which has significant advantages in the calculation and analysis of loss of coolant accidents after a rupture in a coolant pipeline of a nuclear power plant, and the calculation accuracy has been widely verified by experiments; the method of the present invention adopts an analysis method that combines probability theory and determinism, and calculates the best estimated values ​​of the core water level, the maximum temperature of the fuel cladding, and the maximum pressure in the containment after a loss of coolant accident in a nuclear power plant from a statistical perspective, fully combining the advantages of the two methods, and can provide a numerical basis for the formulation of accident management procedures and severe accident management guidelines for nuclear power plants. BRIEF DESCRIPTION OF THE DRAWINGS

[0094] Figure 1 The present invention is a flow chart for calculating the best estimated values ​​of key core parameters of a nuclear reactor in a loss of coolant accident. DETAILED DESCRIPTION

[0095] The present invention is further described in detail below in conjunction with the accompanying drawings and specific embodiments.

[0096] like Figure 1 As shown, a method for analyzing key core parameters of a nuclear reactor loss of coolant accident includes the following steps:

[0097] Step 1: Determine the scope of the accident analysis model and the accident process; Carry out the analysis and calculation of the loss of coolant accident (LOCA) for a typical pressurized water reactor nuclear power plant, first determine the accident process and the response of each system in the nuclear power plant at different stages, and establish the selection principle of the nuclear power plant systems that need to be covered by the accident analysis model: It is of great significance to the safety of nuclear power plants and accident prevention and mitigation;

[0098] Since the four key equipment in the reactor coolant primary and secondary loops, namely the reactor pressure vessel, steam generator, pressurizer, and reactor coolant pump, play an irreplaceable role in reading and determining key parameters of nuclear power plants, and the safety injection system, auxiliary water supply system, steam safety valve, and bypass circuit have an important impact on the LOCA accident process, based on the established selection principles, the above key equipment and systems are covered, analyzed, and simplified during the construction of the accident analysis model;

[0099] In a nuclear power plant loss of coolant accident (LOCA accident, also known as a loss of coolant accident), the location and size of the coolant pipe rupture are significantly more relevant to the nuclear power plant's operating status and accident response strategy than other initial state parameters. This is analyzed to determine the sampling range of this parameter, providing an accident starting point for accident analysis and calculation;

[0100] Step 2: Construct an accident analysis model based on the nuclear reactor system analysis program; based on the modeling scope analysis completed in step 1 and the call of each system of the nuclear power plant during the accident process, construct an accident analysis model for the pressurized water reactor nuclear power plant; in the modeling of the reactor coolant system, keep the number of equipment, flow area, flow channel length, and heat exchange area the same as the key parameters in the real equipment, and partially simplify the modeling of the four key equipments of the reactor pressure vessel, steam generator, pressurizer and reactor coolant pump to meet the needs of reading key parameters of the nuclear power plant; at the same time, the modeling of the auxiliary system and safety system that have an important impact on the accident process is also A simplified method is adopted, and the specific operations are as follows: in the modeling process of the auxiliary water supply system, the steam-driven auxiliary water supply pump and the electric auxiliary water supply pump are considered together, and the relevant pipelines and valves are simplified while retaining two auxiliary water supply pipelines and the water supply ports to the three steam generators. In the modeling process of the safety injection system, both the medium-pressure safety injection system and the low-pressure safety injection system are modeled, and the safety injection box is modeled using exclusive components to realize the non-passive safety injection function. The relevant model can realize the simulation of activating different types of circuit equipment in response to different types of accidents during the accident process. In addition, the safety valves and connecting pipelines of the pressurizer and steam generator are also arranged accordingly;

[0101] For the starting point of the LOCA accident - the rupture of the coolant pipeline, a valve with the same equivalent diameter as the rupture is used in the model to simulate it. The valve is connected to a control body with a volume equal to the total volume in the containment to realize the synchronous simulation of the coolant discharge process into the containment.

[0102] Step 3: Steady-state calculation of accident analysis model and acquisition of key operating parameters; Full power is the normal operating state with the highest thermal power of the reactor. At this time, the heat load and flow rate of the coolant are at the maximum level of normal operation. Selecting this state as the initial condition can ensure that the impact of LOCA accidents can be evaluated under the most unfavorable thermal conditions. At the same time, selecting the full power operating state for safety analysis can ensure the conservatism of the analysis results, so as to ensure that the safety system and containment of the nuclear power plant can meet the design requirements under the most stringent working conditions;

[0103] Based on the built typical PWR nuclear power plant accident analysis model, steady-state calculations at full power are carried out. The calculations are considered to be convergent when the coolant mass flow rate, core coolant inlet and outlet temperatures, main steam mass flow rate and pressure change with time by no more than 0.1% per hour. After that, the above parameters are compared with the design values ​​of the nuclear power plant. When the errors between the steady-state simulation values ​​and the design values ​​of the nuclear power plant are within 1%, the steady-state calculation results are considered to meet the transient analysis calculation requirements.

[0104] Step 4: Sampling is carried out based on the Holden sequence to establish a Monte Carlo simulation case set; according to the Wilkes theory, the number of safe calculations using a bilateral 95% confidence interval is 93 times, so 93 cumulative probability points need to be obtained. The Holden sequence is used in the method of the present invention to obtain the cumulative probability points between [0,1], and the cumulative probability points are mapped to obtain the coolant pipeline rupture location and rupture size values ​​through the probability distribution curve of the coolant pipeline rupture location and rupture size, and the Monte Carlo simulation case set is obtained by random combination. For the bilateral tolerance interval, the Wilkes formula is:

[0105] β=1-γ N -N(1-γ)γ N-1 Formula (1)

[0106] In the formula

[0107] β——confidence;

[0108] γ——probability share;

[0109] N——number of calculations;

[0110] Step 5: Solve the key thermal-hydraulic parameters under the LOCA condition of the PWR nuclear power plant; for the accident analysis model constructed for the PWR nuclear power plant, the nuclear reactor system analysis program is used to calculate and solve, and the time-varying characteristics of the key parameters under steady-state and transient conditions are obtained; the nuclear reactor system analysis program uses a two-fluid model to accurately simulate the thermal-hydraulic phenomena of the reactor system, especially the calculation of the rupture accident and the steam generator heat transfer tube rupture accident and other violent phase change accidents in the reactor, realizing the accurate simulation of the phenomena and key parameters when various design basis accidents occur in the reactor;

[0111] The nuclear reactor system analysis program is based on a two-fluid, six-equation analysis model. The three conservation equations of mass, momentum and energy are established for the gas phase and liquid phase respectively as follows: Mass conservation equation:

[0112]

[0113] The specific physical meanings of each item in the formula are as follows:

[0114] α g ——Volume fraction of vapor phase, %;

[0115] ρ g ——Fluid density of vapor phase, kg / m 3 ;

[0116] ν g ——fluid velocity of vapor phase, m / s;

[0117] Γ g ——Source term of vapor phase, kg / m 3 ·s;

[0118] A——circulation area, m 2 ;

[0119] t——time, s;

[0120] α f ——volume fraction of liquid phase, %;

[0121] ρ f ——Fluid density of liquid phase, kg / m 3 ;

[0122] ν f ——fluid velocity of liquid phase, m / s;

[0123] Γ f ——Source term of liquid phase, kg / m 3 ·s;

[0124] Momentum conservation equation:

[0125]

[0126]

[0127] The specific physical meanings of each item in the formula are as follows:

[0128] α g ——Volume fraction of vapor phase, %;

[0129] ρ g ——Fluid density of vapor phase, kg / m 3 ;

[0130] ν g ——fluid velocity of vapor phase, m / s;

[0131] ν gi ——Initial fluid velocity of vapor phase, m / s;

[0132] Γ g ——Source term of vapor phase, kg / m 3 ·s;

[0133] V g ——Volume of vapor phase, m 3 ;

[0134] α f ——volume fraction of liquid phase, %;

[0135] ρ f ——Fluid density of liquid phase, kg / m 3 ;

[0136] ν f ——fluid velocity of liquid phase, m / s;

[0137] ν fi ——Initial fluid velocity of liquid phase, m / s;

[0138] ρ m ——Density of the mixture, kg / m 3 ;

[0139] Γ f ——Source term of liquid phase, kg / m 3 ·s;

[0140] V f ——volume of liquid phase, m 3 ;

[0141] A——circulation area, m 2 ;

[0142] C - interaction coefficient between liquid and vapor phases;

[0143] P——fluid pressure, Pa;

[0144] B x ——Volume force coefficient of the fluid in the x direction, m / s 2 ;

[0145] FWG – virtual mass force coefficient of vapor or liquid phase;

[0146] t——time, s;

[0147] Energy conservation equation:

[0148]

[0149] The specific physical meanings of each item in the formula are as follows:

[0150] α g ——Volume fraction of vapor phase, %;

[0151] ρ g ——Fluid density of vapor phase, kg / m 3 ;

[0152] U g ——Specific internal energy of vapor phase, J / kg;

[0153] ν g ——fluid velocity of vapor phase, m / s;

[0154] Q wg ——heat conduction rate between vapor phase and wall, J / s;

[0155] Q ig ——heat transfer rate between vapor phase and other items, J / s;

[0156] ——Enthalpy value of vapor phase change, J / kg;

[0157] DISS g ——vapor phase energy dissipation function;

[0158] h g ——Specific enthalpy of vapor phase, J / kg;

[0159] V g ——Volume of vapor phase, m 3 ;

[0160] α f ——volume fraction of liquid phase, %;

[0161] ρ f ——Fluid density of liquid phase, kg / m 3 ;

[0162] U f ——Specific internal energy of liquid phase, J / kg;

[0163] ν f ——fluid velocity of liquid phase, m / s;

[0164] Q wf ——heat conduction rate between liquid phase and wall, J / s;

[0165] Q if ——heat transfer rate between the liquid phase and other items, J / s;

[0166] DISS f ——Liquid phase energy dissipation function;

[0167] ——Enthalpy value of liquid phase change, J / kg;

[0168] h g ——Specific enthalpy of vapor phase, J / kg;

[0169] Γ ig ——Mass transfer rate between liquid phase and vapor phase, kg / m 3 ·s;

[0170] Γ w ——Mass transfer rate between liquid phase and wall, kg / m 3 ·s;

[0171] A——circulation area, m 2 ;

[0172] P——fluid pressure, Pa;

[0173] t——time, s;

[0174] Semi-implicit difference is used to discretize the conservation equations of mass, momentum and energy, and numerical calculations are carried out based on the discretized equations; four flow types are considered for water working fluid, including horizontal flow, vertical flow in tubular, annular and rod-shaped areas, mixed flow in the pump and mixed flow in the injection area of ​​the emergency core cooling system; at the same time, critical flow model, horizontal stratified flow model, flow channel cross-section mutation model, gas-liquid two-phase critical flow flow model, thermal stratification effect model, lateral flow model, non-condensable gas model and air gap heat conduction model are added to realize auxiliary calculation; a point reactor dynamics model is used for the core power of the nuclear reactor, and the reactor power is assumed to be an equation of time and space. In the point reactor dynamics model, it is assumed that the core power distribution is close to a constant;

[0175] Step 6: Obtain key parameters of the loss of coolant accident; for the nuclear reactor loss of coolant accident, the description of the extreme accident state is: a coolant main pipeline undergoes brittle fracture and the pipeline is completely disconnected and dislocated; in this state, the coolant is simultaneously ejected from the two disconnected ports, that is, from openings equivalent to twice the cross-section of the main pipeline; the continuous loss of coolant in the large breach LOCA exposes the core, greatly reduces the heat transfer capacity, and damages the fuel elements. At the same time, the zirconium cladding of the fuel elements will undergo the following series of violent chemical reactions with water vapor at high temperatures:

[0176] Zr+2H 2 0=ZrO 2 +2H 2 +6.51[MJ / kg(Zr)] Formula (8)

[0177] In addition, high-temperature and high-pressure coolant is sprayed into the containment space, which increases the gas pressure and temperature in the containment, endangering the integrity of the containment. Based on the key results of the above accident process evolution, three parameters, namely, reactor core liquid level, fuel cladding maximum temperature, and pressure in the containment, are selected as key parameters that directly reflect whether the nuclear power plant is safe. The relatively more frequent medium-break LOCA and small-break LOCA accidents in nuclear power plants also cause similar consequences to large-break LOCA, but their severity is relatively lower. The method of the present invention uses the same key parameters as large-break LOCA to evaluate the consequences of the accident. Large-break coolant loss The sequence of events occurring in the accident is defined as four consecutive stages, namely, blowdown, refilling, reflooding and long-term cooling. When the reactor undergoes the reflooding stage, the temperature of the fuel core and the cladding can be guaranteed to be within a controllable range. The reactor can then be guaranteed to be always in a safe state by deriving the decay heat through long-term cooling. The method of the present invention uses the accident calculation to reach the core reflooding as the calculation completion condition. For working conditions such as the extremely small breach LOCA where the core is not exposed, the moment when the pressure of the first and second circuits of the reactor are consistent is selected as the calculation termination time. At this time, the decay heat of the reactor is continuously removed through the steam generator to achieve long-term cooling.

[0178] Step 7: Statistical analysis is performed to obtain the best estimate of the key parameters of the core of the nuclear reactor in the event of a loss of coolant accident; the number of safe calculations is confirmed to be 93 according to the double-sided 95% confidence interval, and the corresponding core water level {h 1 ,h 2 ,……,h 92 ,h 93}、Maximum fuel cladding temperature {t 1 ,t 2 ,……,t 92 ,t 93}、Maximum pressure in containment {p 1 ,p 2 ,……,p 92 ,p 93}, perform statistical analysis on these parameters, calculate the best estimated value of each parameter, and use formula (9) to calculate the best estimated value;

[0179]

[0180] In the formula

[0181] - best estimates of key parameters of pressurized water reactor nuclear power plants;

[0182] x i ——The key parameter values ​​of the PWR nuclear power plant calculated in the ith case;

[0183] Through step 1, the key systems and accident processes involved in the loss of cooling accident in the pressurized water reactor nuclear power plant are analyzed to determine the modeling scope of the accident analysis model and the location and size of the breach starting factors; through step 2, an accident analysis model suitable for the pressurized water reactor nuclear power plant is established based on the nuclear reactor system analysis program; through step 3, the full-power steady-state operation calculation of the typical pressurized water reactor based on the accident analysis model is completed and the key operating parameters in the model are read; through step 4, the initial transient operating condition example set required for Monte Carlo analysis is established; based on the system analysis program, through step 5, the changes in key thermal-hydraulic parameters of the typical pressurized water reactor nuclear power plant under LOCA conditions are solved; through step 6, the core liquid level, the maximum temperature of the fuel cladding, The specific value of the pressure inside the containment; the statistics of all calculation conditions are realized through step 7, and the number of safety calculations with a double-sided 95 confidence interval that conforms to the Wilkes theory is used to obtain 93 groups of sample data of key parameters of PWR nuclear power plants under LOCA accidents, and their mean is taken as the best estimate; based on the above steps, the nuclear reactor system analysis program is used to complete the best estimate value calculation of key parameters such as core liquid level, maximum temperature of fuel cladding, and containment pressure under the Monte Carlo theory framework under the coolant loss accident of PWR nuclear power plants. The best estimate value is the value with the highest probability of occurrence of key parameters such as core liquid level, maximum temperature of fuel cladding, and containment pressure after a coolant loss accident occurs in a nuclear power plant, which provides a strong reference for the selection of accident response strategies.

Claims

1. A method for analyzing key core parameters of a nuclear reactor loss of coolant accident, characterized in that: The steps include: Step 1: Determine the scope of the accident analysis model and the accident process; Carry out the analysis and calculation of the loss of coolant accident for a typical pressurized water reactor nuclear power plant, first determine the accident process and the response of each system in the nuclear power plant at different stages, and establish the selection principle of the nuclear power plant systems that need to be covered by the accident analysis model: It is of great significance to the safety of nuclear power plants and accident prevention and mitigation; Since the four key equipment in the reactor coolant primary and secondary loops, namely the reactor pressure vessel, steam generator, pressurizer, and reactor coolant pump, play an irreplaceable role in the reading and determination of key parameters of nuclear power plants, and the safety injection system, auxiliary water supply system, steam safety valve, and bypass circuit have an important impact on the process of coolant loss accidents, based on the established selection principles, the above key equipment and systems are covered, analyzed, and simplified during the construction of the accident analysis model; In a nuclear power plant coolant loss accident, the location and size of the coolant pipeline rupture are significantly more relevant to the nuclear power plant operating status and accident response strategy than other initial state parameters. This is analyzed to determine the sampling range of this parameter and provide an accident starting point for accident analysis and calculation. Step 2: Construction of accident analysis model based on nuclear reactor system analysis program; Based on the modeling scope analysis completed in step 1 and the call of each system of the nuclear power plant during the accident process, an accident analysis model is built for the pressurized water reactor nuclear power plant; in the modeling of the reactor coolant system, the number of equipment, flow area, flow channel length, and heat exchange area are kept the same as the key parameters in the real equipment, and the four types of key equipment, namely the reactor pressure vessel, steam generator, pressurizer, and reactor coolant pump, are partially simplified modeled to meet the needs of reading key parameters of the nuclear power plant; at the same time, the simplified method is also used in the modeling of auxiliary systems and safety systems that have an important impact on the accident process. The specific operations are as follows: In the modeling process of the auxiliary water supply system, the steam-driven auxiliary water supply pump and the electric auxiliary water supply pump are combined for consideration. The relevant pipelines and valves are simplified while retaining two auxiliary water supply pipelines and the water supply ports to the three steam generators. In the modeling process of the safety injection system, both the medium-pressure safety injection system and the low-pressure safety injection system are modeled, and the safety injection box is modeled using exclusive components to realize the non-passive safety injection function. The relevant model can simulate the activation of different types of circuit equipment in response to different types of accidents during the accident process. In addition, the safety valves and connecting pipelines of the regulator and steam generator are also arranged accordingly; For the coolant loss accident starting point, i.e., the coolant pipeline rupture, the model uses a valve with the same equivalent diameter as the rupture to simulate it. The valve is connected to a control body with a volume equal to the total volume in the containment to achieve synchronous simulation of the coolant discharge process into the containment. Step 3: Steady-state calculation of accident analysis model and acquisition of key operating parameters; Full power is the normal operating state with the highest thermal power of the reactor. At this time, the heat load and flow rate of the coolant are at the maximum level of normal operation. This state is selected as the initial condition to ensure that the impact of the coolant loss accident is evaluated under the most unfavorable thermal conditions. At the same time, the full power operating state is selected to carry out safety analysis to ensure the conservatism of the analysis results, so as to ensure that the safety system and containment of the nuclear power plant can meet the design requirements under the most stringent working conditions; Based on the built typical PWR nuclear power plant accident analysis model, steady-state calculations at full power are carried out. The calculations are considered to be convergent when the coolant mass flow rate, core coolant inlet and outlet temperatures, main steam mass flow rate and pressure change with time by no more than 0.1% per hour. After that, the above parameters are compared with the design values ​​of the nuclear power plant. When the errors between the steady-state simulation values ​​and the design values ​​of the nuclear power plant are within 1%, the steady-state calculation results are considered to meet the transient analysis calculation requirements. Step 4: Sampling is carried out based on the Holden sequence to establish a Monte Carlo simulation case set; according to the Wilkes theory, the number of safe calculations is confirmed to be 93 times using a bilateral 95% confidence interval, so 93 cumulative probability points need to be obtained. The Holden sequence is used to obtain the cumulative probability points between [0,1]. Through the probability distribution curve of the coolant pipeline rupture location and rupture size, the cumulative probability points are mapped to obtain the coolant pipeline rupture location and rupture size values, and the Monte Carlo simulation case set is obtained by random combination. For the bilateral tolerance interval, the Wilkes formula is: β=1-γ N -N(1-γ)γ N-1 official(1) In the formula β——confidence; γ——probability share; N——number of calculations; Step 5: Solve the key thermal-hydraulic parameters under the condition of loss of coolant in the pressurized water reactor nuclear power plant; for the accident analysis model constructed for the pressurized water reactor nuclear power plant, the nuclear reactor system analysis program is used to calculate and solve, and the time-varying characteristics of the key parameters under steady-state and transient conditions are obtained; the nuclear reactor system analysis program uses a two-fluid model to accurately simulate the thermal-hydraulic phenomena of the reactor system, especially the calculation of the rupture accident and the steam generator heat transfer tube rupture accident, which are violent phase change phenomena in the reactor, and realizes the accurate simulation of the phenomena and key parameters when various design basis accidents occur in the reactor; The nuclear reactor system analysis program is based on a two-fluid, six-equation analysis model, and establishes three conservation equations of mass, momentum, and energy for the gas phase and liquid phase respectively as follows: The mass conservation equation: The specific physical meanings of each term in the formula are as follows: g ——Volume fraction of vapor phase, %; ρ g ——Fluid density of vapor phase, kg / m 3 ; ν g ——fluid velocity of vapor phase, m / s; Γ g ——Source term of vapor phase, kg / m 3 ·s; A——circulation area, m 2 ; t——time, s; α f ——volume fraction of liquid phase, %; ρ f ——Fluid density of liquid phase, kg / m 3 ; ν f ——fluid velocity of liquid phase, m / s; Γ f ——Source term of liquid phase, kg / m 3 ·s; Momentum conservation equation: The specific physical meanings of each term in the formula are as follows: g ——Volume fraction of vapor phase, %; ρ g ——Fluid density of vapor phase, kg / m 3 ; ν g ——fluid velocity of vapor phase, m / s; ν gi ——Initial fluid velocity of vapor phase, m / s; Γ g ——Source term of vapor phase, kg / m 3 ·s; V g ——Volume of vapor phase, m 3 ; α f ——volume fraction of liquid phase, %; ρ f ——Fluid density of liquid phase, kg / m 3 ; ν f ——fluid velocity of liquid phase, m / s; ν fi ——Initial fluid velocity of liquid phase, m / s; ρ m ——Density of the mixture, kg / m 3 ; Γ f ——Source term of liquid phase, kg / m 3 ·s; V f ——volume of liquid phase, m 3 ; A——circulation area, m 2 ; C——interaction coefficient between liquid phase and vapor phase; P——fluid pressure, Pa; B x ——Volume force coefficient of the fluid in the x direction, m / s 2 ; FWG——virtual mass force coefficient of vapor or liquid phase; t——time, s; Energy conservation equation: The specific physical meanings of each item in the formula are as follows: α g ——Volume fraction of vapor phase, %; ρ g ——Fluid density of vapor phase, kg / m 3 ; U g ——Specific internal energy of vapor phase, J / kg; ν g ——fluid velocity of vapor phase, m / s; Q wg ——heat transfer rate between vapor phase and wall, J / s; Q ig ——heat transfer rate between vapor phase and other items, J / s; ——Enthalpy value of vapor phase change, J / kg; DISS g ——vapor phase energy dissipation function; h g ——Specific enthalpy of vapor phase, J / kg; V g ——Volume of vapor phase, m 3 ; α f ——volume fraction of liquid phase, %; ρ f ——Fluid density of liquid phase, kg / m 3 ; U f ——Specific internal energy of liquid phase, J / kg; ν f ——fluid velocity of liquid phase, m / s; Q wf ——heat conduction rate between liquid phase and wall, J / s; Q if ——heat transfer rate between liquid phase and other items, J / s; DISS f ——Liquid phase energy dissipation function; ——Enthalpy value of liquid phase change, J / kg; h g ——Specific enthalpy of vapor phase, J / kg; Γ ig ——Mass transfer rate between liquid phase and vapor phase, kg / m 3 ·s; Γ w ——Mass transfer rate between liquid phase and wall, kg / m 3 ·s; A——circulation area, m 2 ; P——fluid pressure, Pa; t——time, s; Semi-implicit difference is used to discretize the conservation equations of mass, momentum and energy, and numerical calculations are carried out based on the discretized equations; four flow types are considered for water working fluid, including horizontal flow, vertical flow in tubular, annular and rod-shaped areas, mixed flow in the pump and mixed flow in the injection area of ​​the emergency core cooling system; at the same time, critical flow model, horizontal stratified flow model, flow channel cross-section mutation model, gas-liquid two-phase critical flow flow model, thermal stratification effect model, lateral flow model, non-condensable gas model and air gap heat conduction model are added to realize auxiliary calculation; a point reactor dynamics model is used for the core power of the nuclear reactor, and the reactor power is assumed to be an equation of time and space. In the point reactor dynamics model, it is assumed that the core power distribution is close to a constant; Step 6: Obtain key parameters of the coolant loss accident; For the coolant loss accident of a nuclear reactor, the description of the extreme accident state is: a coolant main pipe undergoes brittle fracture and the pipe is completely disconnected and dislocated; in this state, the coolant is simultaneously ejected from the two disconnected ports, that is, from openings equivalent to twice the cross-section of the main pipe; in the large-break coolant loss accident, the continuous loss of coolant exposes the core, reduces the heat transfer capacity, and damages the fuel elements. At the same time, the zirconium cladding of the fuel elements will undergo the following series of violent chemical reactions with water vapor at high temperatures: Zr+2H20=ZrO2+2H2+6.51[MJ / kg(Zr)] Formula (8) In addition, high-temperature and high-pressure coolant is sprayed into the containment space, which increases the gas pressure and temperature in the containment, endangering the integrity of the containment. Based on the key results of the above accident process evolution, three parameters, namely, reactor core liquid level, fuel cladding maximum temperature, and pressure in the containment, are selected as key parameters that directly reflect whether the nuclear power plant is safe. The relatively more frequent medium-break coolant loss and small-break coolant loss in nuclear power plants cause similar accident consequences to large-break coolant loss, but their severity is relatively lower. Therefore, the same key parameters as large-break coolant loss are used to evaluate the accident consequences. The sequence of events in a coolant loss accident is defined as four consecutive stages, namely, blowdown, refilling, reflooding and long-term cooling. When the reactor undergoes the reflooding stage, the temperature of the fuel core and the cladding can be kept within a controllable range. The reactor can then be kept in a safe state by deriving decay heat through long-term cooling. The accident calculation reaches the core reflooding as the calculation completion condition. For the extremely small break coolant loss condition where the core is not exposed, the moment when the pressure of the first and second circuits of the reactor are consistent is selected as the calculation termination time. At this time, the reactor decay heat is continuously removed through the steam generator to achieve long-term cooling. Step 7: Statistical analysis is used to obtain the best estimate of the key parameters of the core of the nuclear reactor in the event of a loss of coolant accident; the number of safe calculations is confirmed to be 93 according to the double-sided 95% confidence interval, and the corresponding core water level {h1,h2,……,h 92 ,h 93 }、Maximum fuel cladding temperature {t1, t2, ..., t 92 ,t 93 }、The maximum pressure in the containment {p1,p2,……,p 92 ,p 93 }, perform statistical analysis on these parameters, calculate the best estimated value of each parameter, and use formula (9) to calculate the best estimated value; In the formula - best estimates of key parameters of pressurized water reactor nuclear power plants; x i ——The key parameter values ​​of the pressurized water reactor nuclear power plant calculated in the i-th example.

Images