Method for rapidly determining flow demand of nuclear reactor safety injection system

Abstract

The invention belongs to the technical field of reactor thermal hydraulic design and safety analysis, and particularly relates to a method for rapidly determining the flow demand of a nuclear reactor safety injection system. The method comprises three parts which are respectively aimed at three types of coolant loss accidents: the first part is a large-break coolant loss accident; the second part is medium and small break coolant loss accidents; and the third part is tiny crevasse coolant loss accidents. The method has the advantages that compared with multi-professional iterative design argumentation used in a traditional design method, efficient forward safety injection system capacity design and argumentation are achieved, the minimum flow requirement is proposed forward from the perspective of accident safety requirements, and subsequent equipment design can be optimized on the basis of the minimum flow requirement. On one hand, the design period of the nuclear reactor safety injection system is shortened, on the other hand, optimization of equipment design is facilitated, the occupied area of equipment is reduced, and the construction cost of safety injection system equipment and devices is reduced.

Description

Technical Field

[0001] The invention belongs to the technical field of reactor thermal hydraulic design and safety analysis, and in particular relates to a method for quickly determining flow demand of a nuclear reactor safety injection system. Background Art

[0002] The safe operation of nuclear reactors has always been a matter of great concern in the development and application of nuclear energy. Among them, the accident safety analysis of nuclear reactors involves the safety certification of reactors and is the focus of much attention. Among the many hypothetical initiating events of nuclear reactors, a loss of coolant accident refers to an accident in which a breach or rupture occurs in the pressure boundary of the reactor's main loop, and part or most of the coolant leaks. Due to the complexity of the phenomenon of loss of coolant accidents and serious consequences, loss of coolant accidents are one of the most important design basis accidents in reactor safety analysis. According to the size of the breach and the physical phenomenon, loss of coolant accidents are usually divided into several categories, such as large breach, medium breach, small breach, instrument pipe breach and micro breach.

[0003] In order to ensure the safety of nuclear reactors in the event of a loss of coolant accident, it is necessary to promptly add coolant to the reactor's main coolant system after the accident occurs to remove the residual heat in the reactor and prevent the core from being exposed and melted. The replenishment of coolant in the event of a loss of coolant accident is mainly performed by the injection system, which is usually divided into four parts according to its function: high-pressure injection system, injection tank, low-pressure injection system, and long-term core cooling system. In the actual design of nuclear reactors, in order to take into account both safety and economy, the injection system needs to be capacity-designed to ensure that the coolant flow it provides can ensure the safety of the reactor.

[0004] The current method for designing the capacity and flow rate of nuclear reactor injection systems is relatively passive. First, the structural geometric design of each device is given, and then a detailed model of the reactor is established using a thermal-hydraulic system program to demonstrate the safety and conservatism of the equipment design. If the equipment design cannot meet the safety requirements of the reactor, it is necessary to repeatedly modify the plan and perform modeling analysis. This iterative process is time-consuming and labor-intensive, and it is difficult to carry out economic optimization. Summary of the invention

[0005] The purpose of the present invention is to provide a method for quickly determining the flow demand of a nuclear reactor injection system. By analyzing the phenomena and hydraulic flow characteristics of the reactor in a loss of coolant accident, a calculation method for the minimum flow required to ensure reactor safety in a typical loss of coolant accident condition is developed to guide the structural design of various equipment and devices in the nuclear reactor injection system.

[0006] The technical solution of the present invention is as follows: A method for quickly determining the flow demand of a nuclear reactor injection system comprises three parts, respectively targeting three types of coolant loss accidents:

[0007] Part I: Large Break Loss of Coolant Accident;

[0008] Part II: Accidents of loss of coolant due to small and medium breaches;

[0009] Part 3: Coolant loss accident due to minor breach.

[0010] The large break coolant loss event is divided into four stages, namely: the blowdown stage, i.e. the stage from the beginning of the rupture to the end of the bypass; the refilling stage, i.e. the process from the end of the bypass to the beginning of flooding of the bottom of the core; the reflooding stage, i.e. the beginning of flooding of the core until the core is completely cooled; and the long-term cooling stage, i.e. the establishment of long-term cooling.

[0011] The first part comprises the following steps:

[0012] Step 1: Calculation of flow demand during the spraying phase;

[0013] Step 2: Calculation of heat in the pile during the refilling phase;

[0014] Step 3: Calculation of flow demand during the refilling phase;

[0015] Step 4: Calculation of flow demand during the reflooding phase;

[0016] Step 5: Calculation of heat in the pile during the long-term cooling phase;

[0017] Step 6: Calculation of flow requirements for the long-term cooling phase.

[0018] The step 1 is to determine the end time of the spraying stage. The end time of the spraying and the core temperature at the end of the spraying are calculated and determined by the thermal hydraulic system program.

[0019] The step 2 is to determine the heat source in the reactor during the accident stage. The heat source for heating the cladding is decay heat. The decay heat power is calculated using the ANS1973 standard curve, with an initial 102% rated power and a 2-times standard deviation correction for decay heat uncertainty, a power peak factor of 2.45, and a conservative correction of 15%.

[0020] Step 3 is to use the time t1 at the end of the spraying and the cladding temperature T1 as the initial value, and use the decay heat curve divided by the average heat capacity of the fuel and the cladding to calculate the heating rate v of the fuel rod. T , determine the maximum acceptable temperature T2 at the end of the refilling stage, and determine the end time t2 of the refilling stage based on the temperature, as shown in formula (51),

[0021]

[0022] After calculating the duration of the refilling phase (t2-t1), the flow injection requirement of the refilling phase can be obtained by dividing the water volume corresponding to the total volume of the descending section and the lower chamber in the pressure vessel by the duration of the refilling phase. In this calculation, it is assumed that the coolant injected into this volume is saturated water.

[0023] In step 4, the required reflooding rate is taken as the advancement rate of the quenching front. After determining the reflooding rate, the duration of the reflooding stage can be calculated by dividing half of the core active area height by the rate. The required injection flow rate is calculated based on the reflooding rate multiplied by the net flow cross-sectional area of ​​the core.

[0024] Step 5 is a long-term cooling phase, in which the decay heat power P d The ANS1973 standard curve is used for calculation, with an initial 102% rated power and a 2-time standard deviation correction for decay heat uncertainty. For metal heat storage, the total heat storage of the metal below the core coolant inlet and outlet pipelines is calculated based on the flow path during the low-pressure injection phase. The total heat follows an exponential decay law over time and is calculated based on formula (52):

[0025]

[0026] Where: H0 represents the total heat storage of the metal in the pile when the accident just happened; t represents the time; H S represents the total heat storage of the metal at time t; P S represents the release power of metal heat storage at time t; A and B are constants;

[0027] The total heat storage H0 of the metal in the pile when the accident just occurs can be calculated and determined based on the material properties and state changes, as shown in formula (53):

[0028]

[0029] Where: i represents the type number of the heat storage metal; m i represents the total mass of the ith metal; C Pi represents the specific heat capacity of the ith metal; T i ini represents the initial accident temperature of the ith metal, calculated based on the steady-state program; T i fin It represents the final stable temperature of the i-th metal, and obtains the curves of decay heat and metal heat storage power changing with time. The total waste heat release power curve is obtained by adding the decay heat power P d and metal heat storage release power P S Sure.

[0030] Step 6 is based on the assumption that the initial injection water temperature of the low-pressure injection system is T c, at this time the pressure inside the containment is P con , determine the enthalpy h of the initial injected supercooled water c The low-pressure injection system injects coolant into the core, which is heated to saturation and then vaporized. It is finally ejected from the rupture and pressure relief valve in the form of a gas-liquid two-phase mixture. The mass gas content in the two-phase mixture is x. According to the containment pressure, the enthalpy values ​​of saturated water and saturated steam at the current pressure are determined to be h. sl and h sg The minimum injection flow rate required by the low-pressure injection system is calculated by formula (54):

[0031]

[0032] The beneficial effect of the present invention is that compared with the multi-professional iterative design demonstration used in the traditional design method, the present invention realizes efficient forward injection system capacity design and demonstration, and proposes the minimum flow requirement from the perspective of accident safety requirements. The subsequent equipment design can be optimized based on the minimum flow requirement. On the one hand, it shortens the design cycle of the nuclear reactor injection system, and on the other hand, it is also conducive to the optimization of equipment design, reducing the equipment footprint, and reducing the construction cost of injection system equipment and devices, which has important economic significance and use value. BRIEF DESCRIPTION OF THE DRAWINGS

[0033] Figure 1 The present invention provides a flow chart of a method for quickly determining flow requirements of a nuclear reactor safety injection system. DETAILED DESCRIPTION

[0034] The present invention is further described in detail below with reference to the accompanying drawings and specific embodiments.

[0035] like Figure 1 As shown, the present invention provides a method for quickly determining the flow demand of a nuclear reactor injection system, which includes three parts, respectively targeting three types of coolant loss accidents, as follows:

[0036] Part I: Large Break Loss of Coolant Accident

[0037] A large-break loss of coolant accident can usually be divided into four stages, namely: the blowdown stage, which is the stage from the beginning of the rupture to the end of the bypass; the refilling stage, which is the process from the end of the bypass to the beginning of flooding of the bottom of the core; the reflooding stage, which is the beginning of flooding of the core until the core is completely quenched; and the long-term cooling stage, which is the establishment of long-term cooling. In a large-break loss of coolant accident, the main heat sources in the core are decay heat, metal heat storage, and heat generated by the zirconium-water reaction. The main safety acceptance criterion is that the cladding temperature does not exceed 1204°C. Therefore, the capacity demand calculation target of this stage is to determine the minimum injection flow required to keep the cladding temperature from exceeding the target acceptable temperature.

[0038] Step 1: Calculation of flow requirements during the spraying phase

[0039] During the blowdown phase, due to the core flow reversal and bypass phenomenon in the early stage of the coolant loss accident due to the large rupture of the cold pipe section, it is conservatively assumed that all the coolant flow of the injection system will be lost from the rupture during the blowdown phase. In this phase, the injection flow is mainly provided by the high-pressure injection system and the injection tank. Since it is assumed that all flows flow out from the rupture, the calculation in this phase is mainly to determine the end time of the blowdown phase, and there is no need to propose the injection flow demand. The end time of the blowdown and the core temperature at the end of the blowdown can be calculated and determined by the thermal hydraulic system program (such as ARSAC, CATHARE and other thermal hydraulic system programs).

[0040] Step 2: Calculation of the heat in the pile during the refilling phase

[0041] The refilling phase begins when the coolant starts to be injected into the pressure vessel and ends when the bottom of the core begins to be flooded. During this process, the core is in a state of approximately exposed adiabatic heating. Therefore, the injection flow requirement during the refilling phase can be calculated based on the conservative assumption of core insulation. First, determine the heat source in the reactor during this accident phase. Since the core is in an exposed state during this phase, the heat source for heating the cladding is mainly decay heat. The decay heat power can be calculated using the ANS1973 standard curve, taking into account the initial 102% rated power and the 2 standard deviation correction of decay heat uncertainty, and taking into account the conservative power peak factor of 2.45 and the conservative correction of 15%.

[0042] The specific amendments are as follows:

[0043] The heat at a certain moment = (ANS1973 standard curve*102%+2 times standard deviation)*2.45*(1+15%).

[0044] Step 3: Calculation of flow requirements during the refilling phase

[0045] Using the time t1 at the end of the spraying and the cladding temperature T1 as the initial values, the fuel rod heating rate v is calculated by dividing the decay heat curve after conservative processing by the average heat capacity of the fuel and cladding T Based on the safety acceptance criteria and taking into account the conservative margin and the cladding temperature rise during the reflooding stage, the maximum acceptable temperature T2 at the end of the reflooding stage is determined, and the end time t2 of the reflooding stage is determined based on this temperature, as shown in formula (51).

[0046]

[0047] After calculating the duration of the refilling stage (t2-t1), the flow injection requirement of the refilling stage can be obtained by dividing the water volume corresponding to the total volume of the descending section and the lower chamber in the pressure vessel by the duration of the refilling stage. In this calculation, it is conservatively assumed that the coolant injected into this volume is saturated water.

[0048] Step 4: Calculation of flow requirements during the reflooding phase

[0049] During the reflooding stage, the injection coolant gradually floods the core from the bottom of the core. During this process, there is a phenomenon of sudden cooling of the cladding surface and the upward advancement of the sudden cooling front. During this process, the required reflooding rate can be taken as the advancement rate of the sudden cooling front, and an appropriate amount of conservative correction is taken into account. The temperature rise during the reflooding stage is also calculated according to the program and conservative correction is taken into account. After determining the reflooding rate, the duration of the reflooding stage can be calculated by dividing half of the core active area height by this rate. The injection flow rate required for the reflooding stage can be calculated based on the reflooding rate multiplied by the net flow cross-sectional area of ​​the core (conservatively ignoring the wall thickness of components such as guide tubes).

[0050] Step 5: Calculation of the heat in the pile during the long-term cooling phase

[0051] Long-term cooling stage, during which the residual heat in the reactor is mainly decay heat and metal heat storage, of which the decay heat power P d The ANS1973 standard curve can be used for calculation, taking into account the initial 102% rated power and 2 times standard deviation correction of decay heat uncertainty.

[0052] The correction process is as follows:

[0053] The heat at a certain moment = ANS1973 standard curve * 102% + 2 times standard deviation.

[0054] For metal heat storage, according to the flow path in the low-pressure injection stage, the total heat storage of the metal below the core coolant inlet and outlet pipelines is mainly considered. This total heat follows the law of exponential decay over time and can be calculated based on formula (52).

[0055]

[0056] Where: H0 represents the total heat storage of the metal in the pile when the accident just happened; t represents the time; H S represents the total heat storage of the metal at time t; P S Represents the release power of metal heat storage at time t; A and B are constants.

[0057] The total heat storage capacity H0 of the metal in the pile when the accident just occurs can be calculated and determined based on the material properties and state changes, as shown in formula (53).

[0058]

[0059] Where: i represents the type number of the heat storage metal; m i represents the total mass of the ith metal; C Pi represents the specific heat capacity of the ith metal; T i ini represents the initial accident temperature of the ith metal, which can be calculated based on the steady-state program; T i fin It represents the final stable temperature of the ith metal, which can be conservatively assumed to be the saturated water temperature under the containment back pressure during the long-term cooling stage. All heat storage metals below the core coolant inlet and outlet pipelines can be divided into four categories according to their types: pressure vessels, reactor internals, fuel pellets, and zirconium cladding.

[0060] Thus, the curves of decay heat and metal heat storage power changing with time can be obtained. The total waste heat release power curve can be obtained by adding the decay heat power P d and metal heat storage release power P S Sure.

[0061] Step 6: Calculation of flow requirements for the long-term cooling phase

[0062] The flow demand during the long-term cooling phase must meet the total heat release power (P d +P S ). In the early stage of the long-term cooling phase, the low-pressure injection system mainly provides the injection flow. In order to fully remove the residual heat from the core, it is necessary to determine the minimum injection flow required by the low-pressure injection system. During the flow calculation process, it is necessary to consider that the gas content at the core outlet is less than 100% to fully cool the upper area of ​​the fuel and meet the requirement that the core boron does not crystallize.

[0063] Assume that the initial injection water temperature of the low-pressure injection system is T c , conservatively consider that this temperature needs to be the maximum design temperature, at which time the pressure inside the containment is P con , so the enthalpy value h of the initial injection of supercooled water can be determined c The low-pressure injection system injects coolant into the core, where it is heated to saturation and then vaporized, and finally ejected from the rupture and pressure relief valve in the form of a gas-liquid two-phase mixture. The mass gas content of the two-phase mixture is x. According to the containment pressure, the enthalpy values ​​of saturated water and saturated steam at the current pressure can be determined as h sl and h sg Therefore, the minimum injection flow rate required for the low-pressure injection system can be calculated by formula (54).

[0064]

[0065] Part II: Small and medium-sized breaches and loss of coolant accidents

[0066] During a small- and medium-sized break coolant loss accident, when the primary circuit of the nuclear power plant is depressurized to the low pressure setting value of the pressurizer, the reactor shutdown signal will be triggered. When the low pressure injection setting value of the pressurizer is reached, the injection system will be activated. Compared with a large-break coolant loss accident, a small- and medium-sized break coolant loss accident is mainly used as a capacity design and demonstration condition for the low-pressure injection system. Therefore, the principle of the calculation method of the minimum flow requirement in this condition is consistent with the medium- and long-term cooling stage of a large-break coolant loss accident.

[0067] Part III: Minor breach coolant loss accidents

[0068] For small breach accident conditions such as instrument pipe rupture, due to the small breach size, the loss of primary circuit coolant and the low rate of pressure drop, the injection tank, low-pressure injection system and long-term core cooling system will not be triggered in such accident conditions, and only the high-pressure injection system needs to be relied on to meet the long-term water replenishment demand. Therefore, when determining the minimum flow demand in such conditions, it is assumed that the minimum flow demand is the breach outflow flow, and a certain conservative margin is considered.

[0069] In the actual calculation process, the best estimated critical flow model can be used to calculate the discharge flow of the rupture based on information such as the internal and external pressures of the rupture and the rupture diameter, and a conservative correction of 20% to 30% can be considered. Finally, the minimum flow rate required to be provided by the high-pressure injection system and its minimum design capacity can be determined through physical property conversion calculations.

[0070] Step 1: Information on the internal and external pressures and diameter of the breach;

[0071] Step 2: Based on the best estimate critical flow model, such as the Henry-Fauske model or the Ransom-Trapp model, the nominal breach flow rate is calculated using information such as the breach internal and external pressures and the breach diameter;

[0072] Step 3: Actual breach flow = nominal breach flow * (1 + 20% ~ 30%);

[0073] Step 4: Due to the difference in temperature and pressure, the coolant density at the breach is different from the coolant density injected into the high-pressure injection system. Based on the density difference, the "actual breach flow" is converted into the minimum flow required.

Claims

1. A method for quickly determining flow requirements of a nuclear reactor safety injection system, characterized in that: It contains three parts, respectively for three types of loss of coolant accidents: Part I: Large Break Loss of Coolant Accident; Part II: Accidents of loss of coolant due to small and medium breaches; Part 3: Coolant loss accident due to minor breach.

2. A method for quickly determining flow requirements of a nuclear reactor safety injection system according to claim 1, characterized in that: The large break coolant loss event is divided into four stages, namely: the blowdown stage, i.e. the stage from the beginning of the rupture to the end of the bypass; the refilling stage, i.e. the process from the end of the bypass to the beginning of flooding of the bottom of the core; the reflooding stage, i.e. the beginning of flooding of the core until the core is completely cooled; and the long-term cooling stage, i.e. the establishment of long-term cooling.

3. A method for quickly determining flow requirements of a nuclear reactor safety injection system according to claim 1 or 2, characterized in that: The first part comprises the following steps: Step 1: Calculation of flow demand during the spraying phase; Step 2: Calculation of heat in the pile during the refilling phase; Step 3: Calculation of flow demand during the refilling phase; Step 4: Calculation of flow demand during the reflooding phase; Step 5: Calculation of heat in the pile during the long-term cooling phase; Step 6: Calculation of flow requirements for the long-term cooling phase.

4. A method for quickly determining flow requirements of a nuclear reactor safety injection system as claimed in claim 3, characterized in that: The step 1 is to determine the end time of the spraying stage. The end time of the spraying and the core temperature at the end of the spraying are calculated and determined by the thermal hydraulic system program.

5. A method for quickly determining flow requirements of a nuclear reactor safety injection system as claimed in claim 3, characterized in that: The step 2 is to determine the heat source in the reactor during the accident stage. The heat source for heating the cladding is decay heat. The decay heat power is calculated using the ANS1973 standard curve, with an initial 102% rated power and a 2-times standard deviation correction for decay heat uncertainty, a power peak factor of 2.45, and a conservative correction of 15%.

6. A method for quickly determining flow requirements of a nuclear reactor safety injection system as claimed in claim 3, characterized in that: Step 3 is to use the time t1 at the end of the spraying and the cladding temperature T1 as the initial value, and use the decay heat curve divided by the average heat capacity of the fuel and the cladding to calculate the heating rate v of the fuel rod. T , determine the maximum acceptable temperature T2 at the end of the refilling stage, and determine the end time t2 of the refilling stage based on the temperature, as shown in formula (51), After calculating the duration of the refilling phase (t2-t1), the flow injection requirement of the refilling phase can be obtained by dividing the water volume corresponding to the total volume of the descending section and the lower chamber in the pressure vessel by the duration of the refilling phase. In this calculation, it is assumed that the coolant injected into this volume is saturated water.

7. A method for quickly determining flow requirements of a nuclear reactor safety injection system as claimed in claim 3, characterized in that: In step 4, the required reflooding rate is taken as the advancement rate of the quenching front. After determining the reflooding rate, the duration of the reflooding stage can be calculated by dividing half of the core active area height by the rate. The required injection flow rate is calculated based on the reflooding rate multiplied by the net flow cross-sectional area of ​​the core.

8. A method for quickly determining flow requirements of a nuclear reactor safety injection system as claimed in claim 3, characterized in that: Step 5 is a long-term cooling phase, in which the decay heat power P d The ANS1973 standard curve is used for calculation, with an initial 102% rated power and a 2-time standard deviation correction for decay heat uncertainty. For metal heat storage, the total heat storage of the metal below the core coolant inlet and outlet pipelines is calculated based on the flow path during the low-pressure injection phase. The total heat follows an exponential decay law over time and is calculated based on formula (52): Where: H0 represents the total heat storage of the metal in the pile when the accident just happened; t represents the time; H S represents the total heat storage of the metal at time t; P S represents the release power of metal heat storage at time t; A and B are constants; The total heat storage H0 of the metal in the pile when the accident just occurs can be calculated and determined based on the material properties and state changes, as shown in formula (53): Where: i represents the type number of the heat storage metal; m i represents the total mass of the i-th metal; C Pi represents the specific heat capacity of the ith metal; T i ini represents the initial accident temperature of the ith metal, calculated based on the steady-state procedure; It represents the final stable temperature of the i-th metal, and obtains the curves of decay heat and metal heat storage power changing with time. The total waste heat release power curve is obtained by adding the decay heat power P d and metal heat storage release power P S Sure.

9. A method for quickly determining flow requirements of a nuclear reactor safety injection system as claimed in claim 3, characterized in that: Step 6 is based on the assumption that the initial injection water temperature of the low-pressure injection system is T c , at this time the pressure inside the containment is P con , determine the enthalpy h of the initial injected supercooled water c The low-pressure injection system injects coolant into the core, which is heated to saturation and then vaporized. It is finally ejected from the rupture and pressure relief valve in the form of a gas-liquid two-phase mixture. The mass gas content in the two-phase mixture is x. According to the containment pressure, the enthalpy values ​​of saturated water and saturated steam at the current pressure are determined to be h. sl and h sg The minimum injection flow rate required by the low-pressure injection system is calculated by formula (54):

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