CN115410731A - Reactor entering feasibility analysis method, device and equipment for repairing assembly in reactor - Google Patents
Reactor entering feasibility analysis method, device and equipment for repairing assembly in reactor Download PDFInfo
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- CN115410731A CN115410731A CN202211011318.0A CN202211011318A CN115410731A CN 115410731 A CN115410731 A CN 115410731A CN 202211011318 A CN202211011318 A CN 202211011318A CN 115410731 A CN115410731 A CN 115410731A
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Abstract
The application relates to a method, a device, equipment and a medium for analyzing the reactor entering feasibility of a repair assembly in a reactor. The method comprises the following steps: when a fuel rod in a reactor core assembly of a reactor is damaged, replacing the damaged fuel rod with a repair rod to obtain a repair assembly; increasing the power of an undamaged fuel rod adjacent to the position of the repair rod in the repair assembly to the peak rod power of the undamaged fuel rod, placing the repair assembly with increased power at the hottest assembly of the reactor core of the reactor, and limiting the inlet flow at the hottest assembly of the reactor core to be the target flow so as to construct the target working condition; constructing different events of the reactor under a target working condition, and calculating the deviation nucleate boiling ratio of each event; and respectively comparing the deviation nucleate boiling ratio of each event with a target limit value, and judging the feasibility of normal operation of the repair assembly in the reactor according to the comparison result. By adopting the method, the feasibility of normal operation of the repair assembly in the reactor can be accurately judged.
Description
Technical Field
The present application relates to the field of nuclear power plant technologies, and in particular, to a method and an apparatus for analyzing reactor feasibility of repairing components in a reactor, a computer device, a storage medium, and a computer program product.
Background
In the operating process of a pressurized water reactor of a nuclear power plant, the fuel rods are damaged due to the existence of impurities or operation errors. Damage to the fuel rods may result in radioactive material leaking into the coolant or increase the risk of leakage.
At present, damaged fuel rods are mainly replaced by stainless steel rods directly, but whether the reactor core is safe after the repaired assembly is put into the reactor cannot be judged.
Disclosure of Invention
In view of the foregoing, there is a need for a reactor repair assembly entering feasibility analysis method, apparatus, computer device, computer readable storage medium and computer program product that can determine whether an adopted replacement damaged fuel rod strategy is feasible.
In a first aspect, a method for analyzing the reactor feasibility of repairing a component in a reactor comprises:
when fuel rods in a reactor core assembly of a reactor are damaged, replacing the damaged fuel rods with repair rods to obtain repair assemblies;
increasing the power of an undamaged fuel rod adjacent to the position of the repair rod in the repair assembly to the peak rod power of the undamaged fuel rod, placing the repair assembly with increased power at the hottest assembly of the reactor core, and limiting the inlet flow at the hottest assembly of the reactor core to be a target flow so as to construct a target working condition;
constructing different events of the reactor under the target working condition, and calculating the deviation nucleate boiling ratio of each event;
and respectively comparing the deviation nucleate boiling ratio of each event with a target limit value, and judging the feasibility of the normal operation of the repair assembly in the reactor according to the comparison result.
In one embodiment, the placing the repair assembly after the increasing of power at the hottest assembly of the reactor core comprises:
acquiring the power of each component in the reactor core, and determining the component with the highest power in the reactor core;
placing the repair assembly with increased power at the highest power assembly of the reactor core.
In one embodiment, the limiting the inlet flow at the hottest core assembly to a target flow comprises:
flow configuring an inlet flow of the core to limit the inlet flow at the hottest components of the core;
and determining the target flow according to the average flow under the thermal design flow of the reactor core.
In one embodiment, the configuring different events of the reactor occurring under the target operating condition, and calculating the deviation nucleate boiling ratio of each event comprises:
constructing a first event, a second event and a third event which occur under the target working condition; the accident severity levels of the first event, the second event and the third event are increased step by step;
and acquiring the deviation nucleate boiling ratio under the worst condition of each of the first event, the second event and the third event.
In one embodiment, the comparing the deviation nucleate boiling ratio of each event with a target limit value respectively, and determining the feasibility of the normal operation of the repair assembly in the reactor according to the comparison result comprises:
when the deviation nucleate boiling ratio of the event is larger than the target limit value, judging that the repair assembly can normally operate in the reactor;
when the off-nucleate boiling ratio of the event is less than or equal to the target limit, determining that the repair assembly is not operating properly in the reactor.
In one embodiment, the method further comprises:
when the repair assembly is judged not to normally operate in the reactor, sending a replacement instruction;
the replacement instruction is used for instructing to update the heap strategy matched with the repair component so that the updated repair component indicated by the updated heap strategy can normally operate in the reactor.
In a second aspect, an apparatus for analyzing the feasibility of repairing a component in a reactor, the apparatus comprising:
the repair module is used for replacing a damaged fuel rod with a repair rod to obtain a repair assembly when the fuel rod in the reactor core assembly of the reactor is damaged;
the operating condition construction module is used for increasing the power of an undamaged fuel rod adjacent to the position of the repair rod in the repair assembly to the peak rod power of the undamaged fuel rod, placing the repair assembly with increased power at the hottest assembly of the reactor core of the reactor, and limiting the inlet flow at the hottest assembly of the reactor core to be target flow so as to construct a target operating condition;
the event construction module is used for constructing different events of the reactor under the target working condition and calculating the deviation nucleate boiling ratio of each event;
and the analysis module is used for comparing the deviation nucleate boiling ratio of each event with a target limit value respectively, and judging the feasibility of the normal operation of the repair assembly in the reactor according to the result obtained by comparison.
In a third aspect, a computer device comprises a memory storing a computer program and a processor implementing the steps of the method described above when the processor executes the computer program.
In a fourth aspect, a computer-readable storage medium has stored thereon a computer program which, when executed by a processor, carries out the steps of the method described above.
In a fifth aspect, a computer program product comprises a computer program which, when executed by a processor, performs the steps of the method described above.
The method, the apparatus, the computer device, the storage medium, and the computer program product for analyzing the stacking feasibility of the repair assembly in the reactor may obtain the most conservative power distribution by replacing the damaged fuel rod with the repair rod and increasing the power of the undamaged fuel rod adjacent to the repair rod to its peak rod power. The repair assembly with increased power is placed at the position of the hottest assembly of the reactor core, and the inlet flow at the hottest assembly of the reactor core is limited, so that the flow passing through the hottest assembly of the reactor core can be the required target flow. Under the premise of obtaining the most conservative power distribution and limiting the inlet flow, different events occurring in the reactor are constructed, and the deviation nucleate boiling ratio of each event is calculated. And comparing each obtained deviation nucleate boiling ratio with a target limit value, and further realizing accurate judgment on the feasibility of normal operation of the repair assembly in the reactor.
Drawings
FIG. 1 is a diagram of an exemplary implementation of a method for in-reactor feasibility analysis of a repair assembly in a reactor;
FIG. 2 is a schematic flow chart diagram illustrating a method for in-reactor feasibility analysis of a repair component in a reactor in accordance with one embodiment;
FIG. 3 is a diagram of a repair stick position in one embodiment;
FIG. 4 is a view showing a position of a repair stick in another embodiment;
FIG. 5 is a diagram of a subchannel architecture in one embodiment;
FIG. 6 is a diagram of a subchannel architecture in another embodiment;
FIG. 7 is a block diagram of an in-reactor feasibility analysis apparatus for repairing components in a reactor according to an embodiment;
FIG. 8 is a diagram illustrating an internal structure of a computer device according to an embodiment.
Detailed Description
In order to make the objects, technical solutions and advantages of the present application more apparent, the present application is described in further detail below with reference to the accompanying drawings and embodiments. It should be understood that the specific embodiments described herein are merely illustrative of and not restrictive on the broad application.
The method for analyzing the reactor-entering feasibility of the repair assembly in the reactor provided by the embodiment of the application can be applied to the application environment shown in fig. 1. Wherein the terminal 102 communicates with the server 104 via a network. The data storage system may store data that the server 104 needs to process. The data storage system may be integrated on the server 104, or may be located on the cloud or other network server. The server 104 acquires the damage condition of the fuel rod in the reactor core assembly of the reactor, and the damaged fuel rod is replaced by the repair rod according to the damage condition of the fuel rod acquired by the server 104. The server 104 increases the power of the undamaged fuel rods adjacent to the repair rods to the peak rod power of the undamaged fuel rods, places the repair assembly, which has increased power to the peak rod power, at the location of the hottest core assembly of the reactor core, and limits the inlet flow at the hottest core assembly to the target flow to construct the target operating condition. Under the premise of the target working condition, the server 104 also constructs different events of the reactor under the target working condition, and calculates the deviation nucleate boiling ratio of each event. And the server 104 compares the calculated deviation nucleate boiling ratios of the events with target limit values respectively, and then judges the feasibility of the normal operation of the repair assembly in the reactor according to the comparison result.
The terminal 102 may be, but not limited to, various personal computers, notebook computers, smart phones, tablet computers, internet of things devices and portable wearable devices, and the internet of things devices may be smart speakers, smart televisions, smart air conditioners, smart car-mounted devices, and the like. The portable wearable device can be a smart watch, a smart bracelet, a head-mounted device, and the like. The server 104 may be implemented as a stand-alone server or a server cluster comprised of multiple servers.
In one embodiment, as shown in fig. 2, a method for analyzing the reactor availability of repairing a component in a reactor is provided, which is described by taking the method as an example applied to the server 104 in fig. 1, and includes the following steps:
step 202, when a fuel rod in a reactor core assembly of the reactor is damaged, replacing the damaged fuel rod with a repair rod to obtain a repair assembly.
The reactor is a reactor which takes pressurized light water which is not boiled as a moderator and a coolant, consists of a fuel assembly, the moderator, a control rod assembly, a burnable poison assembly, a neutron source assembly, a reactor core hanging basket, a pressure shell and the like, and belongs to a reactor type with a large application amount and a large capacity in a nuclear power station.
The reactor core assembly is composed of hundreds of boxless fuel assemblies with square cross sections, the fuel assemblies are vertically placed on a lower grid plate of the reactor core at certain intervals to enable the formed reactor core to be similar to a cylinder, the weight of the reactor core is transmitted to a pressure shell for supporting through the lower grid plate of the reactor core and a hanging basket, and the size of the reactor core is determined according to the power level of the reactor and the loading number of the fuel assemblies.
The fuel rod kernel fuel is in the form of uranium dioxide ceramic pellets sintered from uranium blend powder. The ceramic core block is a cylinder with the diameter of 1 cm and the height of 1 cm. Several hundred pellets are stacked together and packed into a tube of an elongated zirconium alloy material of 1 cm diameter, about 4 meters in length, and about 1 mm in thickness, and are called a fuel rod because the nuclear fission reaction is like a burning atom.
The repair rod refers to a member that can perform work in place of a fuel rod when the fuel rod is broken, such as a stainless steel rod. The repair assembly is an assembly in which a damaged fuel rod is replaced with a repair rod. For example, when a fuel rod in a certain position of a reactor is damaged, the damaged fuel rod needs to be replaced by a repair rod in time, and an assembly after the repair rod is replaced by the repair rod is called a repair assembly.
Specifically, when the server obtains that a damaged fuel rod occurs in the reactor core assembly, the damaged fuel rod needs to be replaced by a repair rod, and the repair rod replaces the damaged fuel rod to continue to perform work, so that the reactor can operate normally. Further, the position of the fuel rod where the breakage occurs can be classified into two cases: one is the breakage of the fuel rod at the position adjacent to the guide tube, and the other is the breakage of the fuel rod at the position adjacent to the non-guide tube.
And 204, increasing the power of the undamaged fuel rods adjacent to the position of the repair rod in the repair assembly to the peak rod power of the undamaged fuel rods, placing the repair assembly with increased power at the hottest assembly of the reactor core of the reactor, and limiting the inlet flow at the hottest assembly of the reactor core to be the target flow so as to construct the target working condition.
Where unbroken fuel rods adjacent to the repair rod location refer to all unbroken fuel rods immediately adjacent to the repair rod. For example, in a repair assembly, if the repair rod is positioned at the edge of the repair assembly and only three unbroken fuel rods are directly adjacent to the repair rod, then the 3 unbroken fuel rods are the unbroken fuel rods adjacent to the repair rod position. For another example, if the repair rod is located at the center of the repair assembly and 8 unbroken fuel rods are directly adjacent to the repair rod, the 8 unbroken fuel rods are the unbroken fuel rods adjacent to the repair rod location. For another example, if the repair rod is positioned adjacent to the guide tube in the repair assembly and only 7 unbroken fuel rods are adjacent to the repair rod, the 7 unbroken fuel rods are the unbroken fuel rods adjacent to the repair rod. The undamaged fuel rods with increased power will form the hottest sub-channels inside the repair assembly, and the form of the sub-channels is divided into two cases according to the position of the damaged fuel rod. The peak rod power is not inherently different from the peak power, but is numerically more accurate than the peak power.
The hottest component of the reactor core refers to a location in the reactor core where the highest powered component is located. Generally, the most central component of the reactor is set as the hottest component in the reactor, and when the power of the most central component is not the highest power component, the power of the most central component needs to be adjusted to the highest power so as to ensure that the most central component is the hottest component of the core.
Limiting the inlet flow at the hottest core components is achieved by a flow distribution device, and the distribution of the core inlet flow can directly determine the critical heat flux density and the heat pipe factor, so as to further determine the operation limit of the nuclear power plant. The critical heat flux density is the maximum value that can be reached, and when it is exceeded, it causes the device to burn out. The heat pipe factor is the ratio of the maximum value to the average value of the quantity in the reactor core. For example, the ratio of the maximum linear power density to the average linear power density is the total heat flux density heat pipe factor.
The target flow is the minimum value of the flow required by the hottest components of the reactor core, and when the reactor works, the minimum flow required by the work must be achieved, so that accidents of the reactor are avoided. For example, when the flow distribution is unbalanced, the power of the local assemblies is changed greatly, fuel damage accidents may occur in the region with small flow, and when the flow distribution is balanced, the heat generated by each assembly in the reactor core can be guaranteed to be taken out, so that the fuel damage is avoided.
The target operating condition refers to a condition where the power of an unbroken fuel rod adjacent to the repair rod position is increased to its peak rod power while placing the repaired assembly with increased power at the hottest assembly of the reactor core and limiting the inlet flow at the hottest assembly of the core to a target flow.
Specifically, the server adjusts to peak rod power when the unbroken fuel rod power adjacent the repair rod in the repair assembly is not at its peak rod power, and does not need to do so when the unbroken fuel rod power is at its peak rod power, and the unbroken fuel rod increased to peak rod power will constitute the hottest sub-channel inside the repair assembly. And then placing the repair assembly at the position of the hottest assembly of the reactor core, and limiting the inlet flow at the hottest assembly of the reactor core to construct a target working condition.
And step 206, constructing different events of the reactor under the target working condition, and calculating the deviation nucleate boiling ratio of each event.
Wherein the different events are configured to yield a deviating nucleate boiling ratio for the different types of events under the target operating conditions. The offset nucleate boiling ratio is the ratio of the given offset nucleate boiling heat flux density to the actual heat flux density at the fuel element cladding surface. The most significant limitation constraining the thermal power output of modern pressurized water reactor cores is the minimum local deviation nucleate boiling ratio.
Specifically, under the target working condition, the server constructs different types of events which can occur in the reactor, and calculates the deviation nucleate boiling ratio of each event to obtain the deviation nucleate boiling ratio of each event.
And step 208, comparing the deviation nucleate boiling ratios of the events with target limit values respectively, and judging the feasibility of the normal operation of the repair assembly in the reactor according to the result obtained by comparison.
The target limit value is a standard value for judging whether the scheme is feasible or not, and can be obtained by a total statistical method or a determinism method. For example, the target limit obtained using the total statistic method is 1.35, and the target limit obtained using the deterministic method is 1.18.
Specifically, the server compares the deviation nucleate boiling ratio obtained by each constructed event with a target limit value, and judges whether the repair assembly can normally operate in the reactor according to a comparison result.
In one particular application, when the calculated offset nucleate boiling ratio for a build event is 2.6 and the target limit is 1.35, the offset nucleate boiling ratio is greater than the target limit, and the repair assembly can operate properly in the reactor. And when the deviation nucleate boiling ratio obtained by calculating the construction event is 1.1 and the target limit value is 1.35, and the deviation nucleate boiling ratio is smaller than the target limit value, the repair assembly cannot normally operate in the reactor.
In the reactor repair assembly entering feasibility analysis method, the damaged fuel rods are replaced by the repair rods, and the power of the undamaged fuel rods adjacent to the repair rods is increased to the peak rod power, so that the most conservative power distribution can be obtained. The repair assembly with increased power is placed at the position of the hottest assembly of the reactor core, and the inlet flow at the hottest assembly of the reactor core is limited, so that the flow passing through the hottest assembly of the reactor core can be the required target flow. Different events occurring in the reactor are constructed and the respective deviation nucleate boiling ratios of all events are calculated, with the most conservative power distribution and restricted inlet flow being obtained. And comparing each obtained deviation nucleate boiling ratio with a target limit value, and further realizing accurate judgment on the feasibility of normal operation of the repair assembly in the reactor.
In one embodiment, placing the power augmented repair assembly at the hottest assembly of the reactor core includes:
and acquiring the power of each component in the reactor core, and determining the component with the highest power in the reactor core.
And placing the repair assembly after power increase at the highest power assembly of the reactor core.
Wherein the power of the most central reactor assembly is typically adjusted to the highest power so that the most central assembly is the hottest core assembly. Further, when the power of the most central component of the reactor is the component with the highest power in the reactor, the most central component of the reactor is directly determined as the hottest component of the core without adjustment, and when the power of the most central component of the reactor is not the component with the highest power in the reactor, the power of the component is required to be adjusted to the highest power so that the most central component is the hottest component of the core.
Specifically, the server may determine the location of the highest power component of the reactor core, i.e., the hottest component, by taking the power of each component in the reactor core. And then the repair assembly with the power increased by the undamaged fuel rod is placed at the position of the hottest assembly of the reactor core.
In this embodiment, the component with the highest power can be determined by obtaining the power of each component of the reactor core, and then the repair component is placed at the hottest component position of the reactor core, so that a conservative power distribution can be obtained, and the feasibility of normal operation of the repair component in the reactor can be more accurately determined.
In one embodiment, limiting the inlet flow at the hottest core assembly to a target flow comprises:
the inlet flow to the core is flow configured to limit the inlet flow at the hottest components of the core.
And determining the target flow according to the average flow under the thermal design flow of the reactor core.
Wherein the configuration of the inlet flow rate to the core is realized by a flow rate distribution device. For example, when the flow required at the core inlet is 95% of the average flow at the thermal design flow, it can be achieved using a flow distribution device.
The average flow under the thermal design flow is obtained by subtracting uncertain factors of power plant flow measurement from the minimum measured flow, and the average flow under the thermal design flow is the minimum flow. For example, the minimum flow rate is measured as m, but there is an uncertainty factor n that affects the measured minimum flow rate value, and the value obtained by subtracting n from m is the average flow rate at the thermal design flow rate.
The target flow rate represents the minimum flow rate required when the reactor is in operation and is mainly determined by the average flow rate under the thermal design flow rate. For example, in thermal analysis, 95% of the average flow at the thermal design flow is the desired target flow value.
Specifically, the servers limit the core inlet flow to a desired target flow. According to the measured minimum flow of the reactor core and uncertain factors of the flow measurement of the power plant, the average flow under the thermal engineering design flow can be obtained. And determining the value of the target flow according to the obtained average flow under the thermal design flow.
In a specific application, if the average flow rate under the thermal design flow rate is m, the required target flow rate is x% of the average flow rate under the thermal design flow rate, that is, the value of the target flow rate is the product of m and x%, and then the inlet flow rate of the hottest component of the reactor core is made to be the required target flow rate through the flow rate distribution device.
In this embodiment, the average flow rate under the thermal design flow rate is obtained through calculation, a target flow rate value can be determined, and then the core inlet flow rate is limited to the target flow rate, so that the core inlet flow rate can be accurately distributed.
In one embodiment, different events occurring in the reactor under the target operating condition are constructed, and the deviation nucleate boiling ratio of each event is calculated, and the method comprises the following steps:
a first event, a second event, and a third event occurring at the target operating condition are constructed. The severity of the accident increases from one event to another.
The deviating nucleate boiling ratios for the worst case of each of the first, second, and third events are obtained.
Wherein the first event represents that the reactor is normally operated without any accident. The second event represents an event in which some minor accident has occurred to the reactor, such as a loss of off-site power event. The third event represents an event in which a serious accident, such as a rod drop accident, has occurred to the reactor. The offset nucleate boiling ratio is the ratio of a given offset nucleate boiling heat flux density to the actual heat flux density at the fuel element cladding surface.
Specifically, the server is used for constructing and analyzing a first event, a second event and a third event which may occur to the reactor core under the target working condition, the severity of the three events is increased step by step, and the deviation nucleate boiling ratio under the worst condition in each event is obtained.
In the embodiment, different events occurring under the target working condition are constructed, so that the deviation nucleate boiling ratios of the events with different grades can be obtained, and the deviation nucleate boiling ratio under the worst condition in each event can be determined according to the obtained deviation nucleate boiling ratio, so that the feasibility of normal operation of the repair assembly in the reactor can be more accurately judged.
In one embodiment, the deviation from nucleate boiling ratio of each event is compared with a target limit, and the feasibility of the normal operation of the repair assembly in the reactor is determined according to the comparison result, wherein the method comprises the following steps:
and when the deviation nucleate boiling ratio of the event is larger than the target limit value, judging that the repair assembly can normally operate in the reactor.
When the deviation nucleate boiling ratio of the event is less than or equal to the target limit, the repair assembly is determined to be unable to operate properly in the reactor.
Specifically, the server calculates the off-nucleate boiling ratio for each event to determine the feasibility of properly operating the repair assembly in the reactor. When the worst case departure nucleate boiling ratio for each event is greater than the target limit employed, then the repair assembly may function properly in the reactor. When the worst case departure nucleate boiling ratio per event is less than the target limit employed, the repair assembly is unable to function properly in the reactor. For example, when the deviation nucleate boiling ratio calculated for a build event is 2.6 and the target limit is 1.35, the deviation nucleate boiling ratio is greater than the target limit, the repair assembly may operate properly in the reactor. For another example, when the deviation nucleate boiling ratio obtained by calculating the worst condition of each event is 1.1 and the target limit value is 1.35, and the deviation nucleate boiling ratio is smaller than the target limit value, the repair assembly cannot normally operate in the reactor.
In this embodiment, the magnitude of the two values can be determined by comparing the off-nucleate boiling ratio for the worst case event to the target limit. According to the comparison result, accurate judgment on the feasibility of normal operation of the repair assembly in the reactor can be realized.
In one embodiment, the method for analyzing the reactor inlet feasibility of the repair assembly in the reactor further comprises:
and sending a replacement instruction when the repair assembly is judged not to normally operate in the reactor.
The replacement instruction is used for indicating that the heap strategy matched by the repair component is updated so that the updated repair component indicated by the updated heap strategy can normally operate in the reactor.
Specifically, when the adopted strategy for repairing the component in the reactor cannot enable the reactor to normally operate, the server sends a replacement instruction to the terminal, and the terminal replaces the matched reactor entering strategy according to the received instruction information, so that the reactor normally operates.
In the embodiment, the replacement instruction is sent to indicate the replacement of the stacking strategy of the infeasible repair component, so that the stacking strategy of the new repair component can be adopted in time, and the normal operation of the reactor is realized.
The application also provides an application scenario, and the application scenario applies the reactor entering feasibility analysis method of the repair assembly in the reactor. Specifically, the application of the reactor entering feasibility analysis method for repairing the component in the reactor in the application scene is as follows: when a fuel rod is damaged in a core assembly of a reactor, the damaged fuel rod needs to be replaced, and an assembly in which the damaged fuel rod is replaced with a repair rod is called a repair assembly. The cases of replacing the repair rods are mainly classified into two cases according to the position of the fuel rod where the breakage occurs, as shown in fig. 3 and 4. Fig. 3 shows a case where the fuel rods adjacent to the guide tube are damaged and replaced with the repair rods, and fig. 4 shows a case where the fuel rods adjacent to the non-guide tube are damaged and replaced with the repair rods. The power of the unbroken fuel rods adjacent to the position of the repair rod in the repair assembly is increased to the peak rod power of the unbroken fuel rods, the fuel rods with the increased power form the hottest sub-channel in the repair assembly, and the structure of the hottest sub-channel in the repair assembly is mainly divided into two cases, as shown in fig. 5 and 6. FIG. 5 illustrates the sub-channel condition formed by heating the fuel rods adjacent the repair rod, with the repair rod positioned adjacent the guide tube. FIG. 6 shows the sub-channel condition formed after heating the fuel rods adjacent to the repair rods, but not adjacent to the guide tubes.
And adjusting the power of the most central component of the reactor to the component with the highest reactor core power of the reactor, and placing the repair component with increased power at the component position with the highest reactor core power of the reactor, wherein the component with the highest reactor core power is the hottest component of the reactor core. Meanwhile, the inlet flow at the hottest component of the core is limited to a target flow by the flow distribution device, and the target flow is determined by the average flow under the thermal design. And different events occurring under the target working condition are constructed, and the deviation nucleate boiling ratio of each event is calculated according to the thermal hydraulic parameters of each constructed event. According to different severity levels of the occurrence events, the occurrence events are divided into a first event, a second event and a third event, and the severity levels of the occurrence events are gradually increased. The thermodynamic and hydraulic parameters for the worst case of the first, second, and third events are shown in tables 1, 2, and 3.
The results of calculating the off-nucleate boiling ratio for each event and the target limit are shown in table 4. According to the deviation nucleate boiling ratios and the target limit values of the first event, the second event and the third event shown in table 4, it is known that the deviation nucleate boiling ratios of the first event, the second event and the third event are all larger than the target limit value of the deviation nucleate boiling ratio, that is, the deviation nucleate boiling phenomenon does not occur, that is, the repair assembly can normally operate in the reactor.
TABLE 1 thermal hydraulic parameters of first event
| Parameter(s) | |
| Thermal power (MWt) | 2895 |
| Thermodynamic flow (m 3/h) | 68520 |
| Primary side average temperature (. Degree. C.) | 310 |
| Inlet temperature (. Degree.C.) | 292.4 |
| System pressure (MPa) | 15.5 |
| Enthalpy rise heat pipe factor F ΔH | 1.59 |
TABLE 2 thermal hydraulic parameters for the second event
| Parameter(s) | |
| Thermal power (MWt) | 2781 |
| Thermodynamic flow (m 3/h) | 53850 |
| Primary side average temperature (. Degree. C.) | 310 |
| Inlet temperature (. Degree.C.) | 292.4 |
| System pressure (MPa) | 16.27 |
| Enthalpy rise heat pipe factor F ΔH | 1.61 |
TABLE 3 thermodynamic hydraulic parameters of the third event
| Parameter(s) | |
| Thermal power (MWt) | 2875 |
| Thermodynamic flow (m 3/h) | 68520 |
| Primary side average temperature (. Degree. C.) | 310 |
| Inlet temperature (. Degree.C.) | 291.52 |
| System pressure (MPa) | 15.67 |
| Enthalpy rise heat pipe factor F ΔH | 1.86 |
TABLE 4 deviation from nucleate boiling ratio calculation results
In one embodiment, as shown in fig. 7, there is provided a reactor feasibility analysis apparatus for repairing a component in a reactor, the apparatus including:
and the repair module 702 is used for replacing the damaged fuel rods with repair rods when the fuel rods in the reactor core assembly of the reactor are damaged to obtain a repair assembly.
And the working condition construction module 704 is used for increasing the power of the undamaged fuel rods adjacent to the position of the repair rods in the repair assembly to the peak rod power of the undamaged fuel rods, placing the repair assembly with increased power at the hottest assembly of the reactor core, and limiting the inlet flow at the hottest assembly of the reactor core to be a target flow so as to construct a target working condition.
And an event construction module 706 for constructing different events of the reactor under the target working condition and calculating the deviation nucleate boiling ratio of each event.
And the analysis module 708 is used for comparing the deviation nucleate boiling ratios of the events with target limit values respectively, and judging the feasibility of the normal operation of the repair assembly in the reactor according to the result obtained by comparison.
In one embodiment, the operating condition configuration module comprises:
and the power acquisition unit is used for acquiring the power of each component in the reactor core and determining the highest power component in the reactor core.
And the component arranging unit is used for arranging the repair component with increased power at the position of the highest power component of the reactor core.
In one embodiment, the operating condition configuration module further comprises:
and the flow limiting unit is used for carrying out flow configuration on the inlet flow of the core so as to limit the inlet flow at the hottest component of the core.
And the target flow determining unit is used for determining the target flow according to the average flow under the thermal design flow of the reactor core.
In one embodiment, the event construction module comprises:
and the event construction unit is used for constructing a first event, a second event and a third event which occur under the target working condition. The severity of the accident increases step by step for the first event, the second event, and the third event.
And an off-nucleate boiling ratio acquisition unit for acquiring off-nucleate boiling ratios in respective worst cases of the first event, the second event, and the third event.
In one embodiment, the analysis module comprises:
and the first judging unit is used for judging that the repair assembly can normally operate in the reactor when the deviation nucleate boiling ratio of the event is larger than the target limit value.
And the second judging unit is used for judging that the repair assembly cannot normally operate in the reactor when the deviation nucleate boiling ratio of the event is less than or equal to the target limit value.
In one embodiment, the analysis module further comprises:
and the instruction sending unit is used for sending a replacement instruction when the repair assembly is judged not to normally operate in the reactor.
And the updating unit is used for indicating the updating of the reactor entering strategy matched with the repair component by the replacing instruction so as to enable the updated repair component indicated by the updated reactor entering strategy to normally operate in the reactor.
The modules in the reactor inlet feasibility analysis device for repairing the components in the reactor can be wholly or partially realized by software, hardware and a combination thereof. The modules can be embedded in a hardware form or independent from a processor in the computer device, and can also be stored in a memory in the computer device in a software form, so that the processor can call and execute operations corresponding to the modules.
In one embodiment, a computer device is provided, which may be a server, and the internal structure thereof may be as shown in fig. 8. The computer device includes a processor, a memory, and a network interface connected by a system bus. Wherein the processor of the computer device is configured to provide computing and control capabilities. The memory of the computer device includes a non-volatile storage medium and an internal memory. The non-volatile storage medium stores an operating system, a computer program, and a database. The internal memory provides an environment for the operating system and the computer program to run on the non-volatile storage medium. The database of the computer device is used for storing the data of the fuel rod breakage condition of the reactor, the power of each fuel rod, each event and the deviation nucleate boiling ratio, the target flow and the target limit value of each event. The network interface of the computer device is used for communicating with an external terminal through a network connection. The computer program is executed by a processor to implement a method for in-reactor feasibility analysis of repaired components in a reactor.
Those skilled in the art will appreciate that the architecture shown in fig. 8 is merely a block diagram of some of the structures associated with the disclosed aspects and is not intended to limit the computing devices to which the disclosed aspects apply, as particular computing devices may include more or less components than those shown, or may combine certain components, or have a different arrangement of components.
In one embodiment, a computer device is provided, comprising a memory and a processor, the memory having a computer program stored therein, the processor implementing the following steps when executing the computer program:
when a fuel rod in a reactor core assembly of a reactor is damaged, replacing the damaged fuel rod with a repair rod to obtain a repair assembly; increasing the power of an undamaged fuel rod adjacent to the position of the repair rod in the repair assembly to the peak rod power of the undamaged fuel rod, placing the repair assembly with increased power at the hottest assembly of the reactor core of the reactor, and limiting the inlet flow at the hottest assembly of the reactor core to be the target flow so as to construct the target working condition; constructing different events of the reactor under a target working condition, and calculating the deviation nucleate boiling ratio of each event; and comparing the deviation nucleate boiling ratio of each event with a target limit value respectively, and judging the feasibility of normal operation of the repair assembly in the reactor according to the result obtained by comparison.
In one embodiment, the processor when executing the computer program further performs the steps of:
acquiring the power of each component in the reactor core, and determining the component with the highest power in the reactor core; and placing the repair assembly with increased power at the highest power assembly of the reactor core.
In one embodiment, the processor, when executing the computer program, further performs the steps of:
configuring the inlet flow of the reactor core to limit the inlet flow at the hottest component of the reactor core; and determining the target flow according to the average flow under the thermal design flow of the reactor core.
In one embodiment, the processor, when executing the computer program, further performs the steps of:
constructing a first event, a second event and a third event which occur under a target working condition; the accident severity levels of the first event, the second event and the third event are gradually increased; the deviating nucleate boiling ratios for the worst case of each of the first, second, and third events are obtained.
In one embodiment, the processor, when executing the computer program, further performs the steps of:
when the deviation nucleate boiling ratio of the event is larger than the target limit value, judging that the repair assembly can normally operate in the reactor; when the deviation nucleate boiling ratio of the event is less than or equal to the target limit, the repair assembly is determined to be unable to operate properly in the reactor.
In one embodiment, the processor, when executing the computer program, further performs the steps of:
when the repair assembly is judged not to normally operate in the reactor, sending a replacement instruction; the replacement instruction is used for instructing to update the heap strategy matched by the repair component so that the updated repair component indicated by the updated heap strategy can normally operate in the reactor.
In one embodiment, a computer-readable storage medium is provided, having a computer program stored thereon, which when executed by a processor, performs the steps of:
when a fuel rod in a reactor core assembly of a reactor is damaged, replacing the damaged fuel rod with a repair rod to obtain a repair assembly; increasing the power of an undamaged fuel rod adjacent to the position of the repair rod in the repair assembly to the peak rod power of the undamaged fuel rod, placing the repair assembly with increased power at the hottest assembly of the reactor core of the reactor, and limiting the inlet flow at the hottest assembly of the reactor core to be the target flow so as to construct the target working condition; constructing different events of the reactor under a target working condition, and calculating the deviation nucleate boiling ratio of each event; and comparing the deviation nucleate boiling ratio of each event with a target limit value respectively, and judging the feasibility of normal operation of the repair assembly in the reactor according to the result obtained by comparison.
In one embodiment, the computer program when executed by the processor further performs the steps of:
acquiring the power of each component in the reactor core, and determining the component with the highest power in the reactor core; and placing the repair assembly with increased power at the highest power assembly of the reactor core.
In one embodiment, the computer program when executed by the processor further performs the steps of:
configuring the inlet flow of the reactor core to limit the inlet flow at the hottest component of the reactor core; and determining the target flow according to the average flow under the thermal design flow of the reactor core.
In one embodiment, the computer program when executed by the processor further performs the steps of:
constructing a first event, a second event and a third event which occur under a target working condition; the accident severity levels of the first event, the second event and the third event are gradually increased; the off-nucleate boiling ratio for the worst case of each of the first, second, and third events is obtained.
In one embodiment, the computer program when executed by the processor further performs the steps of:
when the deviation nucleate boiling ratio of the event is larger than the target limit value, judging that the repair assembly can normally operate in the reactor; when the deviation from nucleate boiling ratio of the event is less than or equal to the target limit, it is determined that the repair assembly is not operating properly in the reactor.
In one embodiment, the computer program when executed by the processor further performs the steps of:
when the repair assembly is judged not to normally run in the reactor, sending a replacement instruction; the replacement instruction is used for indicating that the heap strategy matched by the repair component is updated so that the updated repair component indicated by the updated heap strategy can normally operate in the reactor.
In one embodiment, a computer program product is provided, comprising a computer program which, when executed by a processor, performs the steps of:
acquiring the power of each component in the reactor core, and determining the component with the highest power in the reactor core; and placing the repair assembly with increased power at the highest power assembly of the reactor core.
In one embodiment, the computer program when executed by the processor further performs the steps of:
configuring the inlet flow of the reactor core to limit the inlet flow at the hottest component of the reactor core; and determining the target flow according to the average flow under the thermal design flow of the reactor core.
In one embodiment, the computer program when executed by the processor further performs the steps of:
constructing a first event, a second event and a third event which occur under a target working condition; the accident severity levels of the first event, the second event and the third event are gradually increased; the deviating nucleate boiling ratios of the worst case of each of the first, second, and third events are obtained.
In one embodiment, the computer program when executed by the processor further performs the steps of:
when the deviation nucleate boiling ratio of the event is larger than the target limit value, judging that the repair assembly can normally operate in the reactor; when the deviation nucleate boiling ratio of the event is less than or equal to the target limit, the repair assembly is determined to be unable to operate properly in the reactor.
In one embodiment, the computer program when executed by the processor further performs the steps of:
when the repair assembly is judged not to normally run in the reactor, sending a replacement instruction; the replacement instruction is used for instructing to update the heap strategy matched by the repair component so that the updated repair component indicated by the updated heap strategy can normally operate in the reactor.
It should be noted that the user information (including but not limited to user device information, user personal information, etc.) and data (including but not limited to data for analysis, stored data, displayed data, etc.) referred to in the present application are information and data authorized by the user or sufficiently authorized by each party.
It should be understood that, although the steps in the flowcharts related to the embodiments as described above are sequentially displayed as indicated by arrows, the steps are not necessarily performed sequentially as indicated by the arrows. The steps are not limited to being performed in the exact order illustrated and, unless explicitly stated herein, may be performed in other orders. Moreover, at least a part of the steps in the flowcharts related to the embodiments described above may include multiple steps or multiple stages, which are not necessarily performed at the same time, but may be performed at different times, and the execution order of the steps or stages is not necessarily sequential, but may be rotated or alternated with other steps or at least a part of the steps or stages in other steps.
It will be understood by those skilled in the art that all or part of the processes of the methods of the embodiments described above may be implemented by hardware instructions of a computer program, which may be stored in a non-volatile computer-readable storage medium, and when executed, may include the processes of the embodiments of the methods described above. Any reference to memory, database, or other medium used in the embodiments provided herein may include at least one of non-volatile and volatile memory. The nonvolatile Memory may include Read-Only Memory (ROM), magnetic tape, floppy disk, flash Memory, optical Memory, high-density embedded nonvolatile Memory, resistive Random Access Memory (ReRAM), magnetic Random Access Memory (MRAM), ferroelectric Random Access Memory (FRAM), phase Change Memory (PCM), graphene Memory, and the like. Volatile Memory can include Random Access Memory (RAM), external cache Memory, and the like. By way of illustration and not limitation, RAM can take many forms, such as Static Random Access Memory (SRAM) or Dynamic Random Access Memory (DRAM), among others. The databases referred to in various embodiments provided herein may include at least one of relational and non-relational databases. The non-relational database may include, but is not limited to, a block chain based distributed database, and the like. The processors referred to in the various embodiments provided herein may be, without limitation, general purpose processors, central processing units, graphics processors, digital signal processors, programmable logic devices, quantum computing-based data processing logic devices, or the like.
The technical features of the above embodiments can be arbitrarily combined, and for the sake of brevity, all possible combinations of the technical features in the above embodiments are not described, but should be considered as the scope of the present specification as long as there is no contradiction between the combinations of the technical features.
The above-mentioned embodiments only express several embodiments of the present application, and the description thereof is more specific and detailed, but not construed as limiting the scope of the present application. It should be noted that, for a person skilled in the art, several variations and modifications can be made without departing from the concept of the present application, which falls within the scope of protection of the present application. Therefore, the protection scope of the present application shall be subject to the appended claims.
Claims (10)
1. A method for analyzing the availability of a repair assembly in a reactor on-site, the method comprising:
when fuel rods in a reactor core assembly of a reactor are damaged, replacing the damaged fuel rods with repair rods to obtain repair assemblies;
increasing the power of the undamaged fuel rods in the repair assembly adjacent to the repair rod position to the peak rod power of the undamaged fuel rods, placing the repair assembly with increased power at the hottest assembly of the reactor core, and limiting the inlet flow at the hottest assembly of the reactor core to be the target flow so as to construct the target working condition;
constructing different events of the reactor under the target working condition, and calculating the deviation nucleate boiling ratio of each event;
and respectively comparing the deviation nucleate boiling ratio of each event with a target limit value, and judging the feasibility of the normal operation of the repair assembly in the reactor according to the result obtained by comparison.
2. The method of claim 1, wherein the placing the repair assembly after the power increase at the reactor core hottest assembly comprises:
acquiring the power of each component in the reactor core, and determining the component with the highest power in the reactor core;
placing the repair assembly after power increase at the highest power assembly of the reactor core.
3. The method of claim 1, wherein said limiting the inlet flow at the hottest core assembly to a target flow comprises:
flow configuring an inlet flow of the core to limit the inlet flow at the hottest components of the core;
and determining the target flow according to the average flow under the thermal design flow of the reactor core.
4. The method of claim 1, wherein said constructing different events of said reactor occurring at said target operating condition, calculating a deviating nucleate boiling ratio for each of said events, comprises:
constructing a first event, a second event and a third event which occur under the target working condition; the accident severity levels of the first event, the second event and the third event are increased step by step;
and acquiring the deviation nucleate boiling ratio under the worst condition of each of the first event, the second event and the third event.
5. The method of claim 1, wherein comparing the off-nucleate boiling ratio of each of the events to a target limit and determining the feasibility of the repair assembly operating properly in the reactor based on the comparison comprises:
when the deviation nucleate boiling ratio of the event is larger than the target limit value, judging that the repair assembly can normally operate in the reactor;
when the deviation from nucleate boiling ratio of the event is less than or equal to the target limit, determining that the repair assembly is not operating properly in the reactor.
6. The method of claim 1, further comprising:
when the repair assembly is judged not to normally operate in the reactor, sending a replacement instruction;
the replacement instruction is used for indicating that the heap strategy matched by the repair component is updated so that the updated repair component indicated by the updated heap strategy can normally operate in the reactor.
7. An in-reactor feasibility analysis apparatus for repairing a component in a reactor, the apparatus comprising:
the repair module is used for replacing a damaged fuel rod with a repair rod to obtain a repair assembly when the fuel rod in the reactor core assembly of the reactor is damaged;
the operating condition construction module is used for increasing the power of an undamaged fuel rod adjacent to the position of the repair rod in the repair assembly to the peak rod power of the undamaged fuel rod, placing the repair assembly with increased power at the hottest assembly of the reactor core of the reactor, and limiting the inlet flow at the hottest assembly of the reactor core to be target flow so as to construct a target operating condition;
the event construction module is used for constructing different events of the reactor under the target working condition and calculating the deviation nucleate boiling ratio of each event;
and the analysis module is used for comparing the deviation nucleate boiling ratio of each event with a target limit value respectively, and judging the feasibility of the normal operation of the repair assembly in the reactor according to the result obtained by comparison.
8. A computer device comprising a memory and a processor, the memory storing a computer program, characterized in that the processor realizes the steps of the method of any one of claims 1 to 6 when executing the computer program.
9. A computer-readable storage medium, on which a computer program is stored, which, when being executed by a processor, carries out the steps of the method of any one of claims 1 to 6.
10. A computer program product comprising a computer program, characterized in that the computer program, when being executed by a processor, carries out the steps of the method of any one of claims 1 to 6.
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| CN202211011318.0A CN115410731A (en) | 2022-08-23 | 2022-08-23 | Reactor entering feasibility analysis method, device and equipment for repairing assembly in reactor |
| PCT/CN2023/074173 WO2024040872A1 (en) | 2022-08-23 | 2023-02-02 | Feasibility method and apparatus for adding repair assembly to reactor, and device |
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| CN202211011318.0A CN115410731A (en) | 2022-08-23 | 2022-08-23 | Reactor entering feasibility analysis method, device and equipment for repairing assembly in reactor |
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| WO2024040872A1 (en) * | 2022-08-23 | 2024-02-29 | 中广核研究院有限公司 | Feasibility method and apparatus for adding repair assembly to reactor, and device |
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| US5631937A (en) * | 1995-11-15 | 1997-05-20 | Westinghouse Electric Corporation | Method and apparatus for protecting a PWR from departure from nucleate boiling and hot leg boiling |
| US8811563B2 (en) * | 2004-12-30 | 2014-08-19 | General Electric Company | Method and system for assessing failures of fuel rods |
| JP4703766B2 (en) * | 2007-11-01 | 2011-06-15 | 四国電力株式会社 | Loose parts monitoring method and apparatus |
| CN107221367B (en) * | 2017-07-04 | 2019-07-02 | 苏州热工研究院有限公司 | A kind of safe evaluation method for changing rod assembly and reentering heap |
| CN109686467B (en) * | 2018-12-27 | 2023-03-21 | 中广核研究院有限公司 | Pressurized water reactor fuel assembly damage repair method |
| US11170903B2 (en) * | 2019-06-12 | 2021-11-09 | Westinghouse Electric Company Llc | Method and system to detect and locate the in-core position of fuel bundles with cladding perforations in candu-style nuclear reactors |
| CN115410731A (en) * | 2022-08-23 | 2022-11-29 | 中广核研究院有限公司 | Reactor entering feasibility analysis method, device and equipment for repairing assembly in reactor |
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