Method for testing a primary circuit pressure

Abstract

The application relates to a method for one-loop pressure test applied to a pressurized water reactor nuclear power plant. The method comprises the following steps: setting the one-loop into an evaporator cooling normal shutdown mode (NS / SG); setting the average temperature of the reactor coolant system to be between 290 DEG C and 294.4 DEG C and the pressure to be 155 Bar; setting the pressure control of the reactor coolant system into an automatic mode; setting the boron and water supply system to operate in an automatic mode; obtaining the leakage rate of the reactor coolant system; checking the one-loop after the evaporator cooling normal shutdown mode is continuously operated for a first time length; and judging that the pressure test is qualified under the condition that the leakage rate is less than a preset value and no leakage or deformation occurs in the components of the one-loop. The application can replace the water pressure test with the leakage test as the one-loop system pressure test, and the operation is simple and the difficulty is small, so that the difficulty and the construction period of the pressure test are reduced, and the cost of the pressure test is saved.

Description

TECHNICAL FIELD

[0001] The present application relates to the technical field of nuclear power, in particular to a method for primary loop pressure test. BACKGROUND

[0002] The boundary of the primary loop system of a nuclear power plant is the second barrier for nuclear safety, and its sealing and integrity are directly related to public safety. The conventional method for verifying the safety and reliability of the boundary barrier is system pressure test.

[0003] According to the RSE-M specification, the system pressure test of domestic second-generation pressurized water reactor nuclear power plants is usually a primary loop hydrostatic test at 1.2 times the design pressure in a maintenance cold shutdown mode (without loading fuel). The pressure vessel test temperature is the plastic transition temperature + 30℃, and the test water temperature is 87±2℃.

[0004] The above use of hydrostatic test as system pressure test is relatively cumbersome and difficult, and the entire test duration is about 6 days, which is not conducive to reducing test cost and duration. SUMMARY

[0005] Therefore, it is necessary to provide a method for primary loop pressure test to reduce the difficulty of pressure test, reduce the test duration, and thus save the cost of pressure test.

[0006] The present application provides a method for primary loop pressure test, which is applied to a pressurized water reactor nuclear power plant. The primary loop includes a reactor coolant system (RCP), a chemical and volume control system (RCV), a boron and water supply system (REA), and a boron recovery system (TEP). The reactor coolant system includes a main pump, a pressure vessel, a pressurizer, a steam generator primary loop side, and a main pipeline sequentially connected with the main pump, the pressure vessel, the pressurizer, and the steam generator primary loop side. The method comprises the following steps:

[0007] Setting the primary loop to evaporator cooling normal shutdown mode (NS / SG);

[0008] Setting the average temperature of the coolant of the reactor coolant system to be between 290℃ and 294.4℃, and the pressure to be 155 Bar;

[0009] Setting the pressure control of the reactor coolant system to be in automatic form;

[0010] Setting the boron and water supply system to run in automatic form;

[0011] Obtaining the leakage rate of the reactor coolant system;

[0012] checking the primary loop after a first duration of continuous operation in the evaporator cooling normal shutdown mode;

[0013] In a case where the leakage rate is less than a preset value and no leakage or deformation occurs in the components of the primary loop, it is determined that the pressure test is qualified.

[0014] The method for primary loop pressure test provided in the application is based on the ASME specification, and the test conditions and acceptance standards of the leakage test are improved, so that the leakage test can replace the hydraulic test as the primary loop system pressure test, and can be successfully applied to the nuclear power plant which uses the RSE-M specification for the hydraulic test as the system pressure test. Since the primary loop is in the evaporator cooling normal shutdown mode (NS / SG) when the method is implemented, compared with the maintenance cold shutdown mode in the related art, the fuel, the pressure vessel and the like do not need to be disassembled, the operation is simple, the difficulty is small, and thus the difficulty and the period of the pressure test are reduced, and the cost of the pressure test is saved.

[0015] In some embodiments, the chemical and volume control system includes a volume control tank, the boron recovery system includes a front storage tank, and the step of obtaining the leakage rate of the reactor coolant system includes:

[0016] increasing the liquid level in the volume control tank to 85%;

[0017] obtaining the average temperature T1 and the density ρ1 of the coolant, the volume control tank liquid level N1, the pressurizer liquid level N3, and the water density ρL in the pressurizer at a first time;

[0018] obtaining the average temperature T2 and the density ρ2 of the coolant, the volume control tank liquid level N2, the pressurizer liquid level N4, and the steam density ρG in the pressurizer at a second time;

[0019] obtaining the water volume change amount Δq TEP between the first time and the second time of the front storage tank;

[0020] obtaining the water supply volume Δq T between the first time and the second time of the reactor coolant system;

[0021] obtaining the boron supply volume Δq p between the first time and the second time of the reactor coolant system;

[0022] obtaining the leakage rate Fp of the reactor coolant system according to the formula:

[0023]

[0024] wherein, Δt is the absolute value of the difference between the first time and the second time; k1 is a liquid level correction coefficient of the containment tank; k2 is a liquid level correction coefficient of the pressurizer; k3 is a coolant density correction coefficient.

[0025] In some embodiments, the step of increasing the liquid level in the containment tank to 85% comprises:

[0026] adjusting an inlet three-way valve of the containment tank to a manual mode and making the inlet three-way valve communicate with the containment tank;

[0027] adjusting an exhaust isolation valve of the containment tank to an automatic mode.

[0028] In some embodiments, the method further comprises, after the step of obtaining the leakage rate Fp of the reactor coolant system according to the formula:

[0029] lowering the liquid level of the containment tank and adjusting an inlet three-way valve of the containment tank to an automatic mode.

[0030] In some embodiments, the method further comprises:

[0031] supplying water to a primary shaft seal of the main pump through the chemical and volume control system, so that the leakage rate of the primary shaft seal is greater than or equal to 1.8 m3 / L.

[0032] In some embodiments, the reactor coolant system further comprises a controller electrically connected to the pressurizer, and a spray valve and a heater electrically connected to the controller, respectively.

[0033] The step of setting the pressure control of the reactor coolant system to an automatic mode comprises:

[0034] setting the controller, the spray valve and the heater to an automatic mode;

[0035] When the pressure of the coolant is greater than 155 bar, the controller controls the spray valve to spray steam in the pressurizer.

[0036] When the pressure of the coolant is less than 155 bar, the controller controls the heater to heat water in the pressurizer.

[0037] In some embodiments, the reactor coolant system further comprises a safety isolation valve connected to the pressurizer and a safety protection valve, the safety isolation valve is in an open state, and the safety protection valve is in a closed state.

[0038] In some embodiments, the method further comprises:

[0039] The steam generated in the steam generator secondary side is discharged to the atmosphere or a condenser.

[0040] Water is supplied to the steam generator secondary side to maintain its water level set value at 34% of the narrow range water level.

[0041] In some embodiments, the primary circuit further comprises a reactor residual heat removal system (RRA) which is isolated from the reactor coolant system in the evaporator cooling normal shutdown mode of the primary circuit.

[0042] In some embodiments, the primary circuit further comprises a safety injection system (RIS) which communicates with the reactor coolant system through an injection line provided with a check valve, and the method further comprises:

[0043] Setting the check valve in a sealed state. BRIEF DESCRIPTION OF DRAWINGS

[0044] Figure 1 A flowchart of a method for primary circuit pressure test of an embodiment of the present application. DETAILED DESCRIPTION

[0045] In order to make the above objectives, features and advantages of the present application more apparent, specific embodiments of the present application will be described in detail below with reference to the accompanying drawings. In the following description, numerous specific details are set forth in order to provide a thorough understanding of the present application. However, the present application can be practiced in a number of different ways beyond the specific embodiments described and it is therefore contemplated that the present application covers any and all modifications, equivalents and alternatives falling within the scope of the present application.

[0046] In the description of the present application, it should be understood that if the terms "center", "longitudinal", "transverse", "length", "width", "thickness", "upper", "lower", "front", "back", "left", "right", "vertical", "horizontal", "top", "bottom", "inner", "outer", "clockwise", "counterclockwise", "axial", "radial", "circumferential" and the like appear, these terms indicate the orientation or positional relationship based on the orientation or positional relationship shown in the drawings, and are only for the convenience of describing the present application and simplifying the description, and therefore cannot be understood as indicating or implying that the device or element referred to must have a particular orientation, be constructed and operated in a particular orientation, and therefore cannot be understood as limiting the present application.

[0047] In addition, the terms "first", "second", and the like, if any, are used herein for descriptive purposes only and should not be construed as indicating or implying relative importance or implicating the number of indicated technical features. Thus, a feature defined with "first", "second" may explicitly or implicitly include at least one of the features. In the description of the present application, if the term "plurality" appears, the meaning of "plurality" is at least two, such as two, three, etc., unless otherwise explicitly specified and limited.

[0048] In the present application, unless otherwise explicitly specified and limited, if the terms "mounting", "connecting", "connecting", "fixing" and the like appear, these terms should be interpreted broadly. For example, it can be fixedly connected, or it can be detachably connected, or it can be integrated; it can be mechanically connected, or it can be electrically connected; it can be directly connected, or it can be indirectly connected through an intermediate medium; it can be the internal communication of two elements or the interaction relationship between two elements, unless otherwise explicitly limited. For those skilled in the art, the specific meaning of the above terms in the present application can be understood according to the specific circumstances.

[0049] In the present application, unless otherwise explicitly specified and limited, if the first feature is described as "on" or "under" the second feature and the like, it can mean that the first and second features are in direct contact, or the first and second features are indirectly in contact through an intermediate medium. Moreover, the first feature "above", "above" and "above" the second feature can be the first feature directly above or obliquely above the second feature, or it can only mean that the first feature is higher than the second feature in horizontal height. The first feature "below", "below" and "below" the second feature can be the first feature directly below or obliquely below the second feature, or it can only mean that the first feature is lower than the second feature in horizontal height.

[0050] It should be noted that if an element is referred to as "fixed to" or "disposed to" another element, it can be directly on another element or there can be a middle element. If an element is considered to be "connected" to another element, it can be directly connected to another element or there can be a middle element. If present, the terms "vertical", "horizontal", "up", "down", "left", "right" and similar expressions used in the present application are only for illustrative purposes and do not represent the only implementation.

[0051] The embodiment of the present application provides a method for one-loop pressure test, which is applied to a pressurized water reactor nuclear power plant. The one-loop comprises a reactor coolant system (RCP), a chemical and volume control system (RCV), a boron and water supply system (REA) and a boron recovery system (TEP). The reactor coolant system comprises a main pump, a pressure vessel, a pressurizer, a steam generator one-loop side and a main pipeline which is sequentially connected with the main pump, the pressure vessel, the pressurizer and the steam generator one-loop side. The method for one-loop pressure test comprises the following steps:

[0052] setting the one-loop to an evaporator cooling normal shutdown mode (NS / SG);

[0053] setting the average temperature of the coolant of the reactor coolant system to be between 290 DEG C and 294.4 DEG C and the pressure to be 155 Bar;

[0054] setting the pressure control of the reactor coolant system to be in an automatic form;

[0055] setting the boron and water supply system to be in an automatic form;

[0056] obtaining the leakage rate of the reactor coolant system;

[0057] checking the one-loop after the evaporator cooling normal shutdown mode is continuously operated for a first time length;

[0058] in the case that the leakage rate is less than a preset value and no leakage or deformation occurs in the components of the one-loop, judging that the pressure test is qualified.

[0059] In the present application, the one-loop comprises the reactor coolant system (RCP), the chemical and volume control system (RCV), the boron and water supply system (REA) and the boron recovery system (TEP). The reactor coolant system is the main system of the one-loop, and the working process of the reactor coolant system is as follows: the main pump drives the coolant (containing boron water) to transmit the heat generated by nuclear fission in the reactor core in the pressure vessel to the steam generator one-loop side, exchanges heat with the cooling water in the steam generator two-loop side, and then returns to the core, so as to cool the core and prevent the dye from burning. The water temperature in the steam generator two-loop side is raised by heat to generate steam, which is supplied to the external steam turbine generator set. The pressurizer is used for controlling the pressure of the coolant to prevent the core from generating the off-boiling phenomenon which is not conducive to heat transfer. The reactor coolant system needs to withstand high pressure, constitutes a "pressure boundary" and is one of the three safety barriers of the nuclear power plant, so as to avoid the release of radioactive substances to the external environment.

[0060] The chemical and volume control system (RCV) is an auxiliary system in the primary loop. The temperature of the reactor coolant is always changing, which causes the water volume of the reactor coolant system to change. In addition, the reactor coolant system is under high pressure and inevitably leaks. Therefore, the chemical and volume control system is used to adjust the water volume of the reactor coolant system. Specifically, the chemical and volume control system (RCV) comprises a volume control tank and an upper charging flow channel and a lower discharging flow channel in communication with the volume control tank, and the upper charging flow channel and the lower discharging flow channel are in communication with the reactor coolant system. In this way, when the water volume of the reactor coolant system changes, for example, leaks, the water level of the pressurizer also changes. When the water level of the pressurizer deviates from the set value, the water level of the pressurizer can be maintained by adjusting the flow of the upper charging flow channel, so that the water volume of the reactor coolant system is substantially in a stable state. That is, the water level in the pressurizer can be maintained by the volume control tank. It is easy to understand that the chemical and volume control system also has other functions, such as controlling the PH value and gas content of the reactor coolant, reducing the radioactivity level of the coolant, and adjusting the boron concentration in the coolant. Since it is prior art, it will not be described here.

[0061] The boron and water supply system (REA) is also an auxiliary system in the primary loop, which is a supporting system of the chemical and volume control system and assists in the realization of its functions. For example, when the chemical and volume control system adjusts the water volume of the reactor coolant system, the boron and water supply system can provide the required deaerated and desalted boron-containing water; when the chemical and volume control system adjusts the PH value and gas content of the reactor coolant, the boron and water supply system can provide chemicals such as hydrazine and lithium hydroxide.

[0062] The boron recovery system (TEP) is also an auxiliary system in the primary loop, which is used to receive radioactive waste water from the reactor coolant system to provide sufficient storage volume for reactor coolant drainage. After treatment and detection, the waste liquid is decomposed into water and boric acid for reuse in the primary loop reactor coolant system.

[0063] The method for primary loop pressure test provided in the present application does not use the primary loop water pressure test in the related art according to the RSE-M specification, but a new method for primary loop pressure test is developed based on the ASME specification. The ASME specification states that the system pressure test can be a system leakage test at a pressure not less than 100% of the rated power of the reactor (i.e. operating pressure) during the unit uplink stage (loading fuel). However, the specification only proposes that the system leakage test can be used as a primary loop system pressure test, but does not detail the test.

[0064] Thus, the application specifies the test conditions and acceptance criteria of the system leakage test based on the ASME specification, then performs the leakage test on the reactor coolant system according to the test conditions, and finally judges whether the test is passed according to the acceptance criteria, thereby serving as the pressure test of the primary loop. Specifically, the primary loop is set to the evaporator cooling normal shutdown mode (NS / SG), and then the average temperature of the reactor coolant system is set to 290-294.4°C and the pressure is set to 155 Bar. Then, the pressure control of the reactor coolant system is set to the automatic mode, so that the pressure stabilizer can maintain the internal pressure at any time. Further, the leakage rate of the reactor coolant system is obtained and compared with the preset value. Finally, the primary loop is continuously operated in the evaporator cooling normal shutdown mode for a first duration, and then the primary loop is checked: when the leakage rate is less than the preset value and the components of the primary loop do not have leakage or deformation, it is judged that the pressure test is qualified.

[0065] The preset value refers to the test standard value compared with the actual leakage rate, which is different in different types of nuclear power plants. The first duration refers to the time for which the pressurized water reactor is continuously operated in the evaporator cooling normal shutdown mode. According to different types of nuclear power plants, the first duration also has differences. Taking the Daya Bay nuclear power plant M310 type pressurized water reactor as an example, the applicant has obtained the preset value of 230 L / H and the first duration of 4 hours through in-depth research. When the reactor coolant system of the M310 type pressurized water reactor has a leakage rate less than 230 L / H according to the above process, and the primary loop components such as the primary pump, the pressure vessel, and the main pipeline do not have any leakage or severe deformation after being continuously operated in the evaporator cooling normal shutdown mode for 4 hours, it indicates that the system pressure test is qualified.

[0066] The method for primary loop pressure test provided in the application is based on the ASME specification, and the test conditions and acceptance standards of the leakage test are improved, so that the leakage test can replace the hydraulic test as the primary loop system pressure test, and can be successfully applied to the nuclear power plant which uses the RSE-M specification for the hydraulic test. Since the method is implemented when the primary loop is in the evaporator cooling normal shutdown mode (NS / SG), compared with the maintenance cold shutdown mode in the related art, the fuel, the pressure vessel and the like do not need to be disassembled, the operation is simple, the difficulty is small, and thus the difficulty and the period of the pressure test are reduced, and the cost of the pressure test is saved. In a specific example, the technical personnel uses the above method for system leakage test and boundary check test in the L118 overhaul of the Daya Bay nuclear power plant, so that the test period is greatly saved. Specifically, the maintenance switch cover and the online isolation operation are not needed, and the single application saves the overhaul period by 6 days and reduces the labor cost by 300 man-days, and indirectly brings the power generation economic benefits of 50 million yuan. According to the calculation of 4 times of application in the whole life cycle of a single unit, the labor cost is reduced by 1200 man-days, and the overhaul period is saved by more than 24 days, and the potential power generation economic benefits brought by the period saving are about 230.4 million yuan.

[0067] In some embodiments, the boron recovery system includes a pre-storage tank, and the step of obtaining a leakage rate of the reactor coolant system comprises:

[0068] Increasing the liquid level in the containment tank to 85%;

[0069] Obtaining the average temperature T1 and the density p1 of the coolant, the containment tank liquid level N1, the pressurizer liquid level N3, and the water density pL in the pressurizer at the first time;

[0070] Obtaining the average temperature T2 and the density p2 of the coolant, the containment tank liquid level N2, the pressurizer liquid level N4, and the steam density pG in the pressurizer at the second time;

[0071] Obtaining the water volume change value Dq of the pre-storage tank between the first time and the second time TEP ;

[0072] Obtaining the water supply volume Dq of the reactor coolant system between the first time and the second time T ;

[0073] Obtaining the boron supply volume Dq of the reactor coolant system between the first time and the second time p ;

[0074] Obtaining the leakage rate Fp of the reactor coolant system according to the formula:

[0075]

[0076] Wherein, Δt is the absolute value of the difference between the first time and the second time; k1 is the liquid level correction coefficient of the containment tank; k2 is the liquid level correction coefficient of the stabilizer; k3 is the coolant density correction coefficient.

[0077] The embodiment provides a method for obtaining a leakage rate Fp of a reactor coolant system. After the liquid level of the containment tank is increased to 85%, the containment tank, a front storage tank of a boron recovery system and the reactor coolant system form a stable operation state. The front storage tank of the boron recovery system is also called a TEP head tank, and is described as a TEP head tank hereinafter. That is, the containment tank does not need to be supplied with water from an external environment during the test and operation, and when the containment tank is at the liquid level, the containment tank, the TEP head tank and the reactor coolant system are in dynamic balance. In this way, between the first time and the second time, the coolant volume entering the reactor coolant system and the coolant volume flowing out of the reactor coolant system are equal, and the coolant volume flowing out of the reactor coolant system includes the leakage amount to the outside, so that the leakage rate Fp can be obtained. Specifically, between the first time and the second time, the coolant volume entering the reactor coolant system includes: a change amount k1 (N1-N2) of the water volume of the containment tank, a change amount k2 (N3-N4) (ρL-ρG) of the water volume and the steam volume in the stabilizer, a water volume amount Δq supplied to the reactor coolant system from the outside T , and a boron volume amount Δq supplied to the reactor coolant system from the outside p . The coolant volume flowing out of the reactor coolant system includes: a TEP head tank volume change amount Δq TEP , and the leakage amount FpΔt to the outside. The two are subtracted to 0, and the above formula can be obtained. It is easy to understand that the TEP head tank volume change amount Δq TEP in the above formula is divided into two cases. The first case is that, in the degassing stage, since the TEP head tank is a component in the boron recovery system, the TEP head tank volume change amount Δq TEP is not 0 because the reactor coolant is discharged to the outside of the reactor coolant system. The second case is that, in the non-degassing stage, the coolant in the TEP head tank returns to the reactor coolant system, and therefore Δq TEP is 0.

[0078] In the embodiment, k1 is a containment tank liquid level correction coefficient, which refers to the change in water volume in the containment tank when the liquid level changes by 1 meter, and the unit is L / m. k2 is a pressurizer liquid level correction coefficient, which refers to the change in water volume and steam volume in the containment tank when the liquid level changes by 1 meter, and the unit is L / m. k3 is a coolant density correction coefficient, which refers to that the average temperature of the coolant changes, resulting in a change in water density, but the total volume of the reactor coolant system is fixed, so the water volume inside changes, but this is not due to leakage, so the coolant density correction coefficient is introduced to correct the calculation of the leakage rate. It is easy to understand that, due to the limitations of the size and shape of the containment tank, the size and shape of the pressurizer, and the operating temperature and pressure, the values of k1, k2 and k3 will be different in different pressurized water reactors. In a specific embodiment, taking the Daya Bay M310 pressurized water reactor as an example, k1 = 3464; k2 = 3.594; and k3 = 247.2.

[0079] In the embodiment, the water supply volume Δq T may be calculated by reading the value of the water supply counter of the primary loop system between the first time and the second time; and the boron supply volume Δq p may be calculated by reading the value of the boric acid supply counter of the primary loop system between the first time and the second time; and the change in water volume of the TEP head tank Δq TEP may be calculated by reading the water volume of the TEP head tank at the first time and the second time, respectively. Other parameters can be directly obtained from the existing panel readings of the primary loop, thereby facilitating further reduction of the difficulty of the pressure test and reduction of the test cost.

[0080] In some embodiments, the step of increasing the liquid level in the containment tank to 85% includes:

[0081] adjusting the inlet three-way valve of the containment tank to manual mode and making the inlet three-way valve communicate with the containment tank;

[0082] adjusting the exhaust isolation valve of the containment tank to automatic mode. In this way, the water in the containment tank is manually supplemented to 85% of the water level, which facilitates the control of the change in water volume of the containment tank. By setting the exhaust isolation valve to automatic mode, the containment tank can automatically exhaust, thereby improving the safety of system operation.

[0083] In some embodiments, after the step of obtaining the leakage rate Fp of the reactor coolant system according to the formula, the method further includes:

[0084] The liquid level of the containment tank is reduced, and the inlet three-way valve of the containment tank is adjusted to the automatic mode. In this way, the liquid level of the containment tank after the test is restored to normal, and the primary loop system is kept in the automatic operation state, thereby facilitating the improvement of the stability of the primary loop system.

[0085] In some embodiments, the method further comprises:

[0086] The water is supplied to the primary seal of the main pump through the chemical and volume control system, so that the leakage rate of the primary seal is greater than or equal to 1.8m 3 / L. When the primary loop is set to the evaporator cooling normal shutdown mode, at least two main pumps are operated. At this time, water needs to be supplied to the primary seal of the main pump to ensure that the leakage rate of the primary seal is greater than or equal to 1.8m 3 / L, so that the shaft seal lubrication of the main pump is normal, and the main pump works in a normal state, thereby ensuring the normal operation of the evaporator cooling normal shutdown mode.

[0087] In some embodiments, the reactor coolant system further comprises a controller electrically connected with the pressurizer, and a spray valve and a heater electrically connected with the controller respectively. The spray valve and the heater are used to control the internal pressure of the pressurizer.

[0088] Further, the step of setting the pressure control of the reactor coolant system to the automatic form comprises:

[0089] The controller, the spray valve, and the heater are set to the automatic mode;

[0090] When the pressure of the coolant is greater than 155 bar, the controller controls the spray valve to spray steam in the pressurizer;

[0091] When the pressure of the coolant is less than 155 bar, the controller controls the heater to heat the water in the pressurizer.

[0092] In this embodiment, when the pressure of the coolant is greater than 155 bar, the steam is liquefied into water by spraying, so that the pressure of the coolant is reduced; when the pressure of the coolant is greater than 155 bar, the water in the pressurizer is heated to evaporate into steam, so that the amount of steam increases, thereby increasing the pressure of the coolant. In this way, the pressure of the coolant can be maintained at 155 bar for testing.

[0093] In some embodiments, the reactor coolant system further comprises a safety isolation valve connected with the pressurizer and a safety protection valve, the safety isolation valve is in an open state, and the safety protection valve is in a closed state.

[0094] The safety isolation valve and the safety protection valve in this embodiment play a role of overpressure protection for the reactor coolant system, and are the key to ensure normal operation of the reactor coolant system. In the evaporator cooling normal shutdown mode of the primary loop, the safety isolation valve is in an open state, and the safety protection valve is in a closed state. When the system pressure is greater than the opening pressure of the safety protection valve, the safety protection valve opens to release pressure of the system. When the pressure drops to the closing pressure threshold of the safety protection valve, the safety protection valve closes to prevent the pressure from continuing to drop. When the safety protection valve opens to release pressure and cannot be closed due to a fault, if the pressure continues to drop to the closing pressure threshold of the safety isolation valve, the safety isolation valve is in a closed state, thereby preventing the pressure from continuing to drop, and achieving overpressure and low-pressure protection for the system.

[0095] In some embodiments, the method further includes:

[0096] The steam generated by the secondary loop side of the steam generator is discharged to the atmosphere or a condenser.

[0097] Water is supplied to the secondary loop side of the steam generator to maintain the water level setting value at 34% of the narrow range water level.

[0098] The cooling water of the secondary loop side of the steam generator exchanges heat with the coolant of the primary loop to evaporate the cooling water into steam. In this embodiment, the steam generated by the secondary loop side of the steam generator is discharged to the atmosphere or a condenser, thereby facilitating the treatment of the steam. Since the steam is discharged to the atmosphere or a condenser, the secondary side water will be reduced, and therefore water can be supplied to the secondary loop side of the steam generator through the water flow regulation system (ARE) or the auxiliary water supply system (ASG) to maintain the water level at 34% of the narrow range water level. The narrow range water level refers to the range of the middle two pipes in the steam generator, which is specifically 3.6 m. When the water level is maintained at 34% (1.22 m) of the narrow range water level, it indicates that the internal load of the secondary side is 0. In this way, it is beneficial to reduce the influence of the primary loop test on the secondary side of the steam generator.

[0099] In some embodiments, the primary loop further includes a reactor residual heat removal system (RRA) which is isolated from the reactor coolant system when the primary loop is in the evaporator cooling normal shutdown mode. The reactor residual heat removal system is used for cooling the reactor core when the secondary loop is disabled. Since the pressure test of the present application is performed when the primary loop is in the evaporator cooling normal shutdown mode, the secondary side of the steam generator can normally cool the reactor, and the fuel does not need to be removed, so the reactor residual heat removal system can not be enabled. In the related art, the fuel needs to be removed after shutdown, so the reactor residual heat removal system needs to be opened to remove the reactor residual heat. Compared with the related art, the method of the pressure test of the present application is beneficial to further reduce the test difficulty, improve the test convenience, and thus further reduce the test cost and period.

[0100] In some embodiments, the primary loop further comprises a safety injection system (RIS) in communication with the reactor coolant system via an injection line having a check valve disposed thereon, and the method further comprises:

[0101] The check valve is set to be in a sealed state. In this embodiment, the sealing property of the check valve is detected before the pressure test is performed, so that the check valve is in a sealed state. If the sealing property is poor, the check valve needs to be repaired and replaced. In this way, it is beneficial to avoid the influence of the leakage rate calculation of the reactor coolant system due to the failure of the check valve, and to improve the accuracy of the test results. In addition, it is also beneficial to avoid the outflow of radioactive substances and improve the safety of the test.

[0102] The technical features of the above embodiments can be combined in any manner. To make the description concise, all possible combinations of the technical features in the above embodiments are not described, but as long as the combinations of the technical features do not exist contradictory, they should be considered as the scope of the present disclosure.

[0103] The above embodiments only express several implementation manners of the present application, and the description is more specific and detailed, but it should not be understood as a limitation on the scope of the patent. It should be pointed out that for ordinary skilled in the art, without departing from the concept of the present application, a number of modifications and improvements can be made, which are all within the scope of the present application. Therefore, the scope of the patent protection of the present application should be subject to the appended claims.

Claims

1. A method for primary loop pressure testing, applied to a pressurized water reactor nuclear power plant, wherein the primary loop includes a reactor coolant system (RCP), a chemical and volume control system (RCV), a boron and water replenishment system (REA), and a boron recovery system (TEP), wherein the reactor coolant system includes a main pump, a pressure vessel, a pressurizer, a primary loop side of a steam generator, and a main pipeline sequentially connecting the main pump, the pressure vessel, the pressurizer, and the primary loop side of the steam generator, characterized in that, The method includes: Set the primary loop to evaporator cooling normal shutdown mode (NS / SG); The average temperature of the coolant in the reactor coolant system is set between 290°C and 294.4°C, and the pressure is 155 Bar. The pressure control of the reactor coolant system is set to automatic. The boron and water supply system is configured to operate automatically. Obtain the leakage rate of the reactor coolant system; After running continuously for the first duration in the normal shutdown mode with evaporator cooling, check the primary loop. If the leakage rate is less than a preset value and no leakage or deformation occurs in the components of the primary circuit, the pressure test is deemed to be qualified. Water is supplied to the No. 1 shaft seal of the main pump through the chemical and volumetric control system to ensure that the leakage rate of the No. 1 shaft seal is greater than or equal to 1.8 m. 3 / L.

2. The method for primary circuit pressure testing according to claim 1, characterized in that, The chemical and volumetric control system includes a volumetric control tank, the boron recovery system includes a pre-storage tank, and the step of obtaining the leakage rate of the reactor coolant system includes: Increase the liquid level in the control tank to 85%; The average temperature T1 and density ρ1 of the coolant, the liquid level N1 of the capacity control tank, the liquid level N3 of the pressure regulator, and the water density ρL in the pressure regulator are obtained at the first moment. At the second moment, the average temperature T2 and density ρ2 of the coolant, the liquid level N2 of the control tank, the liquid level N4 of the pressure regulator, and the vapor density ρG in the pressure regulator are obtained. Obtain the change in water volume Δq of the pre-storage tank between the first time point and the second time point. TEP ; Obtain the water makeup volume Δq of the reactor coolant system between the first time point and the second time point. T ; Obtain the boron replenishment volume Δq of the reactor coolant system between the first time point and the second time point. p ; The leakage rate Fp of the reactor coolant system is obtained according to the formula: Where Δt is the absolute value of the difference between the first and second moments; k1 is the liquid level correction coefficient of the control tank; k2 is the liquid level correction coefficient of the regulator; and k3 is the coolant density correction coefficient.

3. The method for primary circuit pressure testing according to claim 2, characterized in that, The step of increasing the liquid level in the control tank to 85% includes: Adjust the inlet three-way valve of the capacity control box to manual mode and connect the inlet three-way valve to the capacity control box; Adjust the exhaust isolation valve of the control box to automatic mode.

4. The method for primary circuit pressure testing according to claim 2, characterized in that, After the step of obtaining the leakage rate Fp of the reactor coolant system according to the formula, the method further includes: Lower the liquid level in the control tank and adjust the inlet three-way valve of the control tank to automatic mode.

5. The method for primary circuit pressure testing according to any one of claims 1 to 4, characterized in that, The reactor coolant system also includes a controller electrically connected to the pressurizer, and a spray valve and a heater electrically connected to the controller respectively; The step of setting the pressure control of the reactor coolant system to automatic mode includes: Set the controller, the spray valve, and the heater to automatic mode; When the pressure of the coolant is greater than 155 bar, the controller controls the spray valve to spray steam into the pressure regulator; When the pressure of the coolant is less than 155 bar, the controller controls the heater to heat the water in the pressure regulator.

6. The method for primary loop pressure testing according to claim 5, characterized in that, The reactor coolant system also includes a safety isolation valve and a safety protection valve connected to the pressurizer. The safety isolation valve is in the open state, and the safety protection valve is in the closed state.

7. The method for primary circuit pressure testing according to any one of claims 1 to 4, characterized in that, The method for primary circuit pressure testing also includes: The steam generated on the secondary circuit side of the steam generator is discharged to the atmosphere or the condenser. Water is added to the secondary circuit side of the steam generator to maintain its water level setpoint at 34% of the narrow range water level.

8. The method for primary circuit pressure testing according to any one of claims 1 to 4, characterized in that, The primary loop also includes a reactor residual heat removal system (RRA), which is isolated from the reactor coolant system when the primary loop is in the normal shutdown mode of evaporator cooling.

9. The method for primary circuit pressure testing according to any one of claims 1 to 4, characterized in that, The primary loop also includes a safety injection system (RIS), which is connected to the reactor coolant system via an injection line. A check valve is installed on the injection line. The method for pressure testing the primary loop further includes: The check valve is set to a sealed state.

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