Fuel rod burnout share analysis method and device under shaft jamming accident, and consequence evaluation method and system

Abstract

The invention discloses a fuel rod burnout share analysis method and device and a consequence evaluation method and system under a shaft jamming accident, and the method comprises the steps: simulating a transient process of the shaft jamming accident of a nuclear power plant, and determining a transient working condition parameter corresponding to a time point of a minimum transient DNBR under the shaft jamming accident; under the condition that the minimum transient DNBR is smaller than a safety limit value, determining an initial thermal hydraulic state of a reactor core of the nuclear power plant according to the transient working condition parameters; according to the initial thermal hydraulic state and a pre-established reactor core channel model, adjusting a radial power factor of a hottest sub-channel of the sub-channel model, and determining a critical power peak factor; and according to the critical power peak factor, determining the burnout share of the fuel rod with DNB in the nuclear power plant shaft jamming accident from a fuel statistical curve. Therefore, the accuracy of analyzing the burnout share of the fuel rod under the shaft clamping accident can be improved.

Description

Technical Field

[0001] The invention belongs to the technical field of nuclear industry, and in particular relates to a method and device for analyzing the burnout share of fuel rods in a shaft-stuck accident, and a method and system for evaluating the consequences. Background Art

[0002] The shaft jam accident in a nuclear power plant belongs to the Class IV operating condition (extreme accident). Assuming that the reactor coolant pump is jammed instantly at time zero, the flow rate of the affected reactor coolant loop will decrease rapidly, and the emergency shutdown will be triggered by the low flow signal. Since the reactor is in power operation when the accident occurs, the reduction of core flow will cause the coolant temperature to rise rapidly, which will cause the fuel rods to deviate from nucleate boiling (DNB).

[0003] In the accident analysis of nuclear reactors, if the deviation from nucleate boiling ratio (DNBR) is less than the limit and there is no reliable fuel damage model that can prove that some fuel rods are not damaged, the most conservative assumption is taken, that is, all potential fuel rods are considered damaged. Therefore, the existing analysis methods are difficult to accurately assess the actual extent of fuel rod burnout because they are too conservative (assuming that all fuel rods with DNBR below the limit are damaged) or fail to fully consider the impact of hydraulic instability.

[0004] Therefore, the existing nuclear reactor accident analysis methods have low accuracy in assessing the extent of fuel rod burning. Summary of the invention

[0005] The technical problem to be solved by the present invention is to provide a method, device, consequence assessment method and system for analyzing the burnt fraction of fuel rods in a shaft jam accident in view of the above-mentioned deficiencies in the prior art. The analysis method can be used to improve the accuracy of analyzing the burnt fraction of fuel rods in nuclear reactor accident analysis.

[0006] In a first aspect, an embodiment of the present invention provides a method for analyzing the burnout fraction of fuel rods in a shaft jam accident, comprising:

[0007] Simulate the transient process of shaft-stuck accidents in nuclear power plants and determine the transient operating parameters corresponding to the time point of the minimum transient DNBR under shaft-stuck accidents;

[0008] When the minimum transient DNBR is less than a safety limit, determining an initial thermal-hydraulic state of the core of the nuclear power plant according to the transient operating condition parameters;

[0009] According to the initial thermal-hydraulic state and the pre-established core sub-channel model, the radial power factor of the hottest sub-channel of the core sub-channel model is adjusted to determine the critical power peak factor, wherein the critical power peak factor is the power peak factor when the minimum transient DNBR of the hottest sub-channel is equal to the safety limit value;

[0010] According to the critical power peak factor, the fuel rod burnout proportion that will cause DNB in ​​the event of a shaft jam in the nuclear power plant is determined from a fuel statistical curve.

[0011] Optionally, simulating the transient process of a shaft-stuck accident in a nuclear power plant to determine the transient operating condition parameters corresponding to the time point of the minimum transient DNBR under the shaft-stuck accident specifically includes:

[0012] The nuclear power plant is modeled using a thermal hydraulic system analysis program to obtain a shaft jamming accident simulation model;

[0013] According to the initial operating parameters of the nuclear power plant and the shaft-stuck accident simulation model, the transient process of the reactor after the shaft-stuck accident occurs is simulated to obtain thermal parameters of the shaft-stuck accident;

[0014] The core subchannel model is established using the subchannel program;

[0015] According to the thermal parameters of the shaft-stuck accident and the core sub-channel model, a DNBR variation curve of the shaft-stuck accident is calculated;

[0016] Determine the time point of the minimum transient DNBR under the axle sticking accident according to the axle sticking accident DNBR variation curve;

[0017] Determine the transient operating condition parameters corresponding to the time point of the minimum transient DNBR in the thermal parameters of the stuck shaft accident.

[0018] Optionally, adjusting the radial power factor of the hottest sub-channel of the sub-channel model according to the initial thermal-hydraulic state and a pre-established core sub-channel model to determine the critical power peak factor specifically includes:

[0019] S1, inputting the initial thermal hydraulic state into the core sub-channel model;

[0020] S2, reducing the radial power factor of the hottest subchannel in the core subchannel model by a preset unit to obtain an updated radial power factor of the hottest subchannel;

[0021] S3, using a three-zone power adjustment method, according to the updated radial power factor of the hottest sub-channel, adjusting the radial power factors of other channels to obtain a radial power factor of each sub-channel;

[0022] S4. Calculate the predicted DNBR variation curve of the core according to the radial power factor of each sub-channel;

[0023] S5. Determine a minimum predicted DNBR based on the predicted DNBR change curve, where the minimum predicted DNBR is the DNBR with the smallest value among the DNBR data of all sub-channels;

[0024] S6. When the minimum predicted DNBR is greater than the safety limit, return to execute steps S2 to S5 until the minimum predicted DNBR is not less than the safety limit, and determine the updated hottest sub-channel radial power factor as the critical power peak factor.

[0025] Optionally, the three-zone power adjustment method is used to adjust the radial power factors of other channels according to the updated radial power factor of the hottest sub-channel to obtain the radial power factor of each sub-channel, specifically including:

[0026] Dividing the fuel rod assemblies in the core into a first partition, a second partition and a third partition, wherein the first partition includes the fuel rod assemblies corresponding to the hottest subchannel, the second partition includes the fuel rod assemblies adjacent to the fuel rod assemblies in the first partition, and the third partition includes other fuel rod assemblies except the fuel rod assemblies in the first partition and the second partition;

[0027] maintaining the radial power factor of the burner rod assembly in the second partition unchanged;

[0028] The radial power factor of the burner rod assembly in the third partition is increased so that the power factor obtained by weighted calculation of each burner rod assembly in the core is 1, thereby obtaining an updated radial power factor of each burner rod assembly in the third partition.

[0029] In a second aspect, an embodiment of the present invention further provides a method for evaluating the consequences of a shaft jamming accident in a nuclear power plant, comprising:

[0030] The method for analyzing the fuel rod burnout fraction under the shaft jamming accident described in the first aspect determines the fuel rod burnout fraction of the nuclear power plant;

[0031] The consequence of the shaft-stuck accident in the nuclear power plant is evaluated according to the fuel rod burnout fraction to obtain an evaluation result, and the evaluation result is used to indicate the accident safety level of the nuclear power plant.

[0032] In a third aspect, an embodiment of the present invention further provides a fuel rod burnout fraction analysis device, comprising:

[0033] The first determination module is used to simulate the transient process of the shaft sticking accident of the nuclear power plant and determine the transient operating condition parameters corresponding to the time point of the minimum transient DNBR under the shaft sticking accident;

[0034] A second determination module, connected to the first determination module, is used to determine the initial thermal-hydraulic state of the core of the nuclear power plant according to the transient operating condition parameters when the minimum transient DNBR is less than the safety limit;

[0035] a third determination module, connected to the second determination module, for adjusting the radial power factor of the hottest sub-channel of the sub-channel model according to the initial thermal-hydraulic state and the pre-established core sub-channel model, and determining a critical power peak factor, wherein the critical power peak factor is the power peak factor when the minimum transient DNBR of the hottest sub-channel is equal to the safety limit;

[0036] The analysis module is connected to the third determination module and is used to determine the fuel rod burnout proportion of DNB occurring in the nuclear power plant shaft jam accident from the fuel statistical curve according to the critical power peak factor.

[0037] In a fourth aspect, an embodiment of the present invention further provides a consequence assessment system for a shaft jam accident in a nuclear power plant, the system comprising:

[0038] The fuel rod burnout fraction analysis device in the case of a shaft jamming accident described in the third aspect is used to determine the fuel rod burnout fraction of the nuclear power plant;

[0039] The evaluation module is connected to the fuel rod burnout fraction analysis device under the shaft jamming accident, and is used to evaluate the consequences of the shaft jamming accident in the nuclear power plant according to the fuel rod burnout fraction to obtain an evaluation result, and the evaluation result is used to indicate the accident safety level of the nuclear power plant.

[0040] The method for analyzing the fuel rod burnout fraction under a shaft-stuck accident of the present invention captures the cooling condition corresponding to the minimum transient DNBR of the core during the shaft-stuck accident through transient simulation; then finely adjusts the power distribution in combination with the sub-channel model, accurately determines the critical power peak factor, and then uses the fuel statistical curve to accurately quantify the fuel rod burnout fraction. Through the detailed simulation of the transient process and the precise adjustment of the power distribution, the limitations of traditional empirical estimation are broken through. As a result, the accuracy of the analysis of the fuel rod burnout fraction under a shaft-stuck accident is improved. BRIEF DESCRIPTION OF THE DRAWINGS

[0041] Figure 1 : is a flow chart of a method for analyzing the fuel rod burnout fraction in a shaft jam accident according to Example 1 of the present invention;

[0042] Figure 2 : is a flow chart of another method for analyzing the fuel rod burnout fraction in the event of an axle jamming accident according to Example 2 of the present invention;

[0043] Figure 3 : is a curve showing changes in thermal parameters such as pressure, power, flow rate, etc. of the main system over time after a shaft-stuck accident in Example 2 of the present invention;

[0044] Figure 4 : is a DNBR transient curve after a shaft-stuck accident in Example 2 of the present invention;

[0045] Figure 5: A process for determining a critical power peak factor according to Embodiment 2 of the present invention;

[0046] Figure 6 : A nuclear power plant fuel statistical curve according to Embodiment 2 of the present invention;

[0047] Figure 7 : This is a structural diagram of a fuel rod burnout fraction analysis device under an axle sticking accident according to Example 4 of the present invention.

[0048] The meanings of the marks in the figure are:

[0049] 1. Pressurizer pressure curve after the shaft-stuck accident; 2. Pressurizer pressure value at the time point of minimum DNBR after the shaft-stuck accident; 3. Core flow curve after the shaft-stuck accident; 4. Core flow at the time point of minimum DNBR after the shaft-stuck accident; 5. Reactor heat flux curve after the shaft-stuck accident; 6. Reactor heat flux at the time point of minimum DNBR after the shaft-stuck accident; 7. DNBR change curve with time after the shaft-stuck accident; 8. Time point of minimum DNBR after the shaft-stuck accident; 9. Initial radial power factor of the hottest channel in the thermal component of the sub-channel model; 10. Power factor of the hottest channel after reduction in the thermal component of the sub-channel model; 11. Recalculated normalized radial power factor of the thermal component. DETAILED DESCRIPTION

[0050] In order to enable those skilled in the art to better understand the technical solution of the present invention, the present invention is further described in detail below with reference to the accompanying drawings and embodiments.

[0051] It should be noted that, in this article, relational terms such as first and second, etc. are only used to distinguish one entity or operation from another entity or operation, and do not necessarily require or imply any such actual relationship or order between these entities or operations. Moreover, the terms "include", "comprise" or any other variants thereof are intended to cover non-exclusive inclusion, so that a process, method, article or device including a series of elements includes not only those elements, but also other elements not explicitly listed, or also includes elements inherent to such process, method, article or device. In the absence of further restrictions, the elements defined by the statement "include..." do not exclude the presence of other identical elements in the process, method, article or device including the elements.

[0052] Departure from nucleate boiling (DNB) refers to the phenomenon that the coolant changes from the efficient nucleate boiling state to the film boiling state with extremely low heat transfer efficiency. This phenomenon usually occurs when the local heat flux of the fuel rods of a nuclear reactor is too high or the coolant flow is insufficient, resulting in the formation of a steam blanket on the surface of the fuel rods, which significantly reduces the efficiency of heat transfer from the fuel rods to the coolant.

[0053] Departure from Nucleate Boiling Ratio (DNBR). Used to evaluate the margin of fuel rod cooling. If DNBR is less than the specified "DNBR limit", it means that the coolant heat transfer capacity is insufficient and deviation from nucleate boiling (DNB) may occur, which may lead to overheating or even damage of the fuel rod cladding.

[0054] DNBR limit is an important safety standard used to evaluate the heat transfer capacity of coolants and the thermal safety of fuel rods during nuclear reactor operation. It represents the minimum permissible value of DNBR. DNBR limit is set based on experience, experiments and theoretical analysis, and it stipulates the critical conditions under which fuel rods can remain safe and undamaged under various assumed operating conditions.

[0055] Standard Review Plan (SRP), the Standard Review Plan of the U.S. Nuclear Regulatory Commission (NRC)

[0056] Hundreds of fuel assemblies are loaded in the core of a nuclear power plant reactor. Currently, most fuel assemblies are in the form of rod bundles. When the coolant flows vertically upward along the rod bundle, there will be lateral flow between the fuel rods. Under accident conditions, phase changes may occur in the core and cause coolant flow fluctuations. At this time, the heat transfer capacity of the core coolant will be affected, causing the fuel rod cladding to heat up, which may in turn lead to a boiling crisis in the reactor core, that is, deviation from nucleate boiling (DNB). In severe cases, the fuel rod cladding may be damaged and lose its integrity.

[0057] The main mechanisms of fuel rod damage include thermal effect, mechanical effect and chemical effect. The fuel damage mechanism considered in the accident analysis of nuclear power plants is mainly thermal effect, that is, overheating of the fuel cladding. For LOCA accidents, the limiting criterion adopted is that the maximum temperature of the outer surface of the fuel cladding should be lower than 1204℃, and for rod ejection accidents, the fuel cladding temperature should be lower than 1482℃. For other types of accidents, it is required not to exceed the DNBR limit.

[0058] In nuclear reactor accident analysis, if the critical thermal conditions of the fuel rods are met (for example, DNBR is always above its limit), it can be assumed that the fuel rods will not be damaged. However, if DNBR is less than its limit (i.e., it may reach the DNB state), since it is impossible to directly measure or verify the changes in heat transfer performance during the accident, a conservative assumption must be made in the analysis that the affected fuel rods may be damaged. Unless it can be proved that the number of actually damaged fuel rods is less than the damage of all fuel rods (the most conservative assumption).

[0059] SRP15.3.3 states as follows: "If the DNBR is less than the DNBR limit, fuel damage must be assumed for all rods that do not meet these criteria, unless it can be demonstrated that the number of damaged fuel rods is smaller based on an acceptable fuel damage model that takes into account the possible adverse effects of hydraulic instability." That is, if the DNBR of the fuel rods is lower than the limit, all of these fuel rods must be considered damaged for safety reasons; however, it is allowed to use a more accurate fuel damage model to demonstrate that the actual number of damaged fuel rods may be smaller. Therefore, if nuclear power plant operators want to avoid overly conservative assessment results, they need to prove that the number of damaged fuel rods is smaller through a more accurate fuel damage prediction model.

[0060] Fuel damage prediction models must consider a variety of key factors, including but not limited to:

[0061] Hydraulic Instabilities: Accurately assess the effects of fluid instabilities that impair the heat transfer capabilities of coolants.

[0062] DNB-affected area: Theoretical or experimental data prove that DNB does not occur in all areas, but only in local areas.

[0063] Thermodynamics and local conditions: Analysis of the local temperature and power distribution of specific fuel rods can be compared with the empirical values ​​of fuel rod damage.

[0064] The shaft jam accident in a nuclear power plant belongs to the Class IV operating condition (extreme accident). Assuming that the reactor coolant pump is jammed instantly at time zero, the flow rate of the affected reactor coolant loop will decrease rapidly, and the emergency shutdown will be triggered by the low flow signal. Since the reactor is in power operation when the accident occurs, the reduction in core flow will cause the coolant temperature to rise rapidly, which will cause DNB of the fuel rods. However, the existing analysis methods are often too conservative (assuming that all fuel rods with DNBR below the limit are damaged) or fail to fully consider the impact of hydraulic instability, making it difficult to accurately assess the actual degree of fuel rod burnout.

[0065] Embodiment 1:

[0066] Based on the above research, in order to solve the above technical problems, such as Figure 1As shown, this embodiment provides a fuel rod burnout fraction analysis method, which is applied to nuclear power plants.

[0067] Specifically, the following steps are included from step 101 to step 104 .

[0068] Step 101, simulating the transient process of a shaft-stuck accident in a nuclear power plant, and determining transient operating condition parameters corresponding to the time point of the minimum transient DNBR under the shaft-stuck accident.

[0069] Specifically, transient simulations of the stuck shaft accident scenario of a nuclear power plant can be performed through thermal hydraulic system analysis programs (such as RELAP5 or TRACE). The focus of the simulation is to capture the changes in core thermal hydraulic parameters (such as pressure, flow, heat flux, etc.) caused by the sudden drop in coolant flow, provide basic data for the subsequent adjustment of the power factor, and ensure that the calculation is based on the most unfavorable conditions of the accident.

[0070] DNBR (Deviation from Nucleate Boiling Ratio) is a key indicator for measuring the cooling margin of fuel rods. Its minimum value corresponds to the most unfavorable moment for fuel rod cooling. By analyzing the DNBR curve over time, the time point of the minimum DNBR is determined and the corresponding transient operating parameters (such as pressurizer pressure, core flow, heat flux, etc.) are extracted.

[0071] Step 102, when the minimum transient DNBR is less than the safety limit, determine the initial thermal-hydraulic state of the core of the nuclear power plant according to the transient operating condition parameters.

[0072] Specifically, if the minimum transient DNBR is less than the safety limit (such as 1.3 or 1.5), it means that there is a risk of deviation from nucleate boiling (DNB) in the core.

[0073] According to the extracted transient operating parameters, the initial thermal-hydraulic state of the core is determined, including coolant flow distribution, pressure distribution, power distribution of fuel assemblies, etc. The initial thermal-hydraulic state can be used to establish or correct the sub-channel model of the core.

[0074] The transient simulation results are converted into the static thermal-hydraulic state of the core as input for subsequent sub-channel analysis.

[0075] Step 103, according to the initial thermal hydraulic state and the pre-established core sub-channel model, the radial power factor of the hottest sub-channel of the sub-channel model is adjusted to determine the critical power peak factor.

[0076] Among them, the subchannel model is a tool for fine analysis of the coolant flow and heat transfer inside the core, and is used to calculate the local thermal parameters of each subchannel. The critical power peak factor is the power peak factor when the minimum transient DNBR of the hottest subchannel is equal to the safety limit. The critical power peak factor reflects the power distribution state of the core under the most unfavorable cooling conditions and is an important parameter for evaluating the thermal safety margin of the core.

[0077] Specifically, by adjusting the radial power factor (i.e., local power distribution) of the hottest channel in the sub-channel model, the minimum DNBR of the hottest channel is calculated step by step.

[0078] When the minimum DNBR is equal to the safety limit, the corresponding power factor is the critical power peak factor. During the adjustment process, the total core power must be kept unchanged (compensated by adjusting the power factors of other channels).

[0079] Step 104, according to the critical power peak factor, the fuel rod burnout ratio that will cause DNB in ​​the event of a shaft jam accident in the nuclear power plant is determined from the fuel statistical curve.

[0080] Among them, the fuel statistical curve reflects the number distribution of fuel rods under different power peak factors, which is usually provided by the physical design data of the nuclear power plant.

[0081] Optionally, the nuclear power plant fuel statistical curve can be obtained through experimental testing (such as pulse heat load test, static thermal failure test, dynamic transient simulation test, etc.), theoretical calculation and modeling (such as probability-physics coupling model, Monte Carlo simulation, etc.), actual operation data analysis, etc.

[0082] Specifically, according to the critical power peak factor, the proportion of fuel rods exceeding this factor is found from the statistical curve; this proportion is the burnt share of fuel rods that may cause DNB under the condition of a shaft jam.

[0083] The proportion of fuel rods burned is an important indicator for assessing the severity of an accident and the amount of radioactive release, and provides a basis for evaluating the consequences of accidents in nuclear power plants.

[0084] In this embodiment, transient simulation is used to capture the cooling condition corresponding to the minimum transient DNBR of the core during the shaft-stuck accident; then the power distribution is finely adjusted in combination with the sub-channel model to accurately determine the critical power peak factor; thus, the fuel statistical curve is used to accurately quantify the fuel rod burnout fraction. Thus, the accuracy of the fuel rod burnout fraction analysis under the shaft-stuck accident is improved.

[0085] Optionally, the above step 101 specifically includes:

[0086] The nuclear power plant was modeled using the thermal hydraulic system analysis program to obtain a simulation model for a stuck shaft accident;

[0087] According to the initial operating parameters of the nuclear power plant and the shaft-stuck accident simulation model, the transient process of the reactor after the shaft-stuck accident is simulated to obtain the thermal parameters of the shaft-stuck accident;

[0088] The core subchannel model is established using the subchannel program;

[0089] According to the thermal parameters of the shaft-stuck accident and the core sub-channel model, the DNBR variation curve of the shaft-stuck accident was calculated;

[0090] Determine the time point of minimum transient DNBR under axle jamming accident according to the DNBR variation curve of axle jamming accident;

[0091] Determine the transient operating parameters corresponding to the time point of the minimum transient DNBR in the thermal parameters of the stuck shaft accident.

[0092] Specifically, (1) the nuclear power plant is modeled using a thermal hydraulic system analysis program to obtain a shaft-stuck accident simulation model. The thermal hydraulic system analysis program dynamically simulates the changes in coolant flow, pressure, temperature and other parameters by solving the mass, momentum and energy conservation equations, especially in the case of a shaft-stuck accident (such as a sudden blockage of the circulating pump) that causes a rapid decrease in coolant flow, and captures the accident transient process.

[0093] Mature thermal hydraulic system analysis programs (such as RELAP5, TRACE, THEMIS, etc.) can be selected to establish a simulation model for a shaft jam accident based on the design parameters of the nuclear power plant (such as core structure, coolant system, pressurizer, etc.). The model can include key components such as circulating pumps, coolant flow paths, and reactor power distribution to truly simulate the impact of a shaft jam accident on the thermal hydraulic state of the core.

[0094] (2) A stuck shaft accident can cause a rapid decay of coolant flow, which in turn leads to the risk of local overheating of the core or insufficient cooling capacity. Through transient simulation, the dynamic changes of thermal parameters during the accident process can be captured, especially the changes in coolant flow and heat exchange capacity, as follows: input the initial operating parameters of the nuclear power plant (such as coolant inlet temperature, flow, pressure, core power distribution, etc.) as the initial conditions of the simulation; run the stuck shaft accident simulation model to simulate the dynamic processes such as the sudden drop in coolant flow, changes in core thermal power distribution and decrease in cooling capacity after the accident; output the curve of the change of core thermal parameters over time during the simulation process, including coolant flow, pressure, temperature, heat flux density, etc.

[0095] (3) Use the subchannel program to establish the core subchannel model:

[0096] Subchannel programs (such as COBRA, VIPRE, FLICA, etc.) are used to establish a refined model of the core, dividing the core coolant area into multiple subchannels; the core geometric parameters (such as fuel rod diameter, rod spacing, guide tube distribution, etc.) and the power distribution of the nuclear power plant are input to establish the subchannel model. The core subchannel model can simulate the local thermal hydraulic state of each channel (such as coolant flow rate, temperature, pressure, etc.) by finely dividing the core coolant flow area, providing a basis for the subsequent calculation of DNBR.

[0097] (4) Based on the thermal parameters of the shaft stuck accident and the core sub-channel model, the DNBR variation curve of the shaft stuck accident is calculated: the thermal parameters (such as pressure, flow rate, heat flux, etc.) obtained from the shaft stuck accident simulation are used as the input conditions of the core sub-channel model; the sub-channel program calculates the DNBR value of each sub-channel in the core by calculating the local thermal state of each sub-channel (such as coolant temperature, pressure, flow rate, etc.) and combining it with the critical heat flux (CHF) correlation formula that deviates from nucleate boiling; the output curve of DNBR variation with time shows the dynamic change of the core cooling margin during the accident.

[0098] The deviation from nucleate boiling ratio (DNBR) is defined as the ratio of the local critical heat flux (CHF) to the actual heat flux. The lower the DNBR value, the worse the cooling capacity. When the DNBR is less than 1, deviation from nucleate boiling (DNB) may occur on the core surface, causing overheating or damage to the fuel rods. By calculating the DNBR change curve, the time point when the cooling capacity is the worst during the accident can be identified.

[0099] (5) According to the DNBR variation curve of the axle jamming accident, determine the time point of the minimum transient DNBR under the axle jamming accident:

[0100] In the output DNBR change curve, find the lowest point of the DNBR value, that is, the moment when the core cooling condition is the most unfavorable. The minimum transient DNBR corresponds to the critical state of the core cooling capacity and is a key parameter in accident analysis. It is used to evaluate whether the core meets the safety limit under the worst operating conditions.

[0101] (6) Determine the transient operating parameters corresponding to the time point of the minimum transient DNBR in the thermal parameters of the stuck shaft accident: extract the thermal parameters at the time when the minimum transient DNBR occurs, including coolant pressure, flow, temperature, heat flux, etc. These thermal parameters can be used as input for subsequent power adjustment and sub-channel analysis.

[0102] The operating parameters at the minimum transient DNBR moment directly reflect the thermal state of the core under the most unfavorable cooling conditions in the accident, and are the key basis for further analysis of the core thermal safety and power distribution optimization.

[0103] In this embodiment, the entire process of a nuclear power plant shaft jam accident can be dynamically simulated to capture changes in coolant flow, pressure and temperature, and provide a time series of core thermal parameters. Through refined core modeling and calculation, combined with the deviation from nucleate boiling correlation, the cooling margin (DNBR) of each sub-channel of the core can be accurately calculated. By analyzing the DNBR variation curve, the limit state of the core cooling capacity can be identified, and the corresponding transient operating parameters can be extracted to provide a basis for subsequent analysis.

[0104] Optionally, the above step 103 specifically includes:

[0105] S1, inputting the initial thermal hydraulic state into the core sub-channel model;

[0106] S2. reducing the radial power factor of the hottest subchannel in the core subchannel model by a preset unit to obtain an updated radial power factor of the hottest subchannel;

[0107] S3, using the three-zone power adjustment method, according to the updated radial power factor of the hottest sub-channel, adjusting the radial power factors of other channels to obtain the radial power factor of each sub-channel;

[0108] S4. Calculate the predicted DNBR variation curve of the core according to the radial power factor of each sub-channel;

[0109] S5. Determine the minimum predicted DNBR based on the predicted DNBR change curve, where the minimum predicted DNBR is the DNBR with the smallest value among the DNBR data of all sub-channels;

[0110] S6. When the minimum predicted DNBR is greater than the safety limit, return to execute steps S2 to S5 until the minimum predicted DNBR is not less than the safety limit, and determine the updated radial power factor of the hottest sub-channel as the critical power peak factor.

[0111] Among them, the subchannel program calculates the thermal hydraulic conditions of the fluid at different subchannel locations through a detailed simulation of the coolant flow and heat transfer inside the core, and then determines the minimum DNBR. The subchannel is to divide the coolant flow area of ​​the nuclear fuel assembly into multiple separate channels, and there is heat and mass exchange between these channels. The subchannel program predicts the possibility of deviation from nucleate boiling through detailed calculations of the local coolant flow (flow, pressure, temperature, etc.) in the reactor core and the heat transfer of the fuel surface.

[0112] Specifically, the coolant flow area inside the core is divided into sub-channels, and the thermal parameters and cooling capacity (DNBR value) of each sub-channel are calculated; the power factor is adjusted in three zones to maintain the overall power of the core constant while focusing on optimizing the cooling state of the hottest sub-channel; the power factor of the hottest sub-channel is gradually reduced, and the DNBR value is calculated using the sub-channel program. Through multiple rounds of iterations, the critical state where the minimum DNBR is equal to the safety limit is found; thereby determining the critical power peak factor, which reflects the power distribution of the core that does not exceed the safety design limit under the worst cooling capacity condition.

[0113] In this embodiment, the cooling capacity limit of the hottest sub-channel of the core is captured through the sub-channel model and multiple rounds of iterative calculations to ensure that the calculation results are highly accurate. That is, by iteratively reducing the power factor of the hottest sub-channel and combining the three-zone power adjustment method, the core power distribution is dynamically optimized while maintaining the normalization of the total core power, thereby achieving a conservative analysis of the core cooling capacity limit state, and then accurately determining the critical power peak factor of the core under the most unfavorable cooling conditions, ensuring that the cooling capacity of the hottest sub-channel meets the safety limit requirements.

[0114] Optionally, the three-zone power adjustment method is used to adjust the radial power factors of other channels according to the updated radial power factor of the hottest sub-channel to obtain the radial power factor of each sub-channel, specifically including:

[0115] Dividing the fuel rod assemblies in the core into a first partition, a second partition and a third partition, wherein the first partition includes the fuel rod assemblies corresponding to the hottest subchannel, the second partition includes the fuel rod assemblies adjacent to the fuel rod assemblies in the first partition, and the third partition includes other fuel rod assemblies except the fuel rod assemblies in the first partition and the second partition;

[0116] Maintaining the radial power factor of the burner rod assembly in the second partition unchanged;

[0117] The radial power factor of the burner rod assembly in the third partition is increased so that the power factor obtained by weighted calculation of each burner rod assembly in the core is 1, thereby obtaining an updated radial power factor of each burner rod assembly in the third partition.

[0118] The core is divided into three zones, that is, all the combustion rod assemblies in the entire core are divided into three zones:

[0119] The first zone: includes the combustion rod assembly corresponding to the hottest sub-channel, that is, the area with the highest power distribution and the largest heat load. The radial power factor of this area is adjusted first (usually reducing the power factor to alleviate the heat load).

[0120] Second zone: The combustion rod assembly adjacent to the first zone, usually in the "second hottest" position. The radial power factor of this zone remains unchanged during adjustment to ensure a smooth transition of power distribution between zones and avoid non-uniform effects caused by local adjustment.

[0121] The third partition: the remaining fuel rod assemblies except the first and second partitions. This partition bears the increase in power factor to compensate for the reduction in total core power caused by the reduction in power factor of the first partition, thereby ensuring that the total power is normalized (equal to 1).

[0122] The zoning is based on the thermal performance of the sub-channels, while taking into account the spatial characteristics of the core power distribution. The first zone is the area with the worst cooling performance and needs to focus on reducing the power factor, while the second zone is a transition zone that keeps the power factor unchanged, and the third zone balances the overall power by increasing the power.

[0123] Specifically, the radial power factor of the first partition is first adjusted, usually by reducing its power factor to reduce the local heat load pressure. The reduction range is a preset fixed unit, which can be determined based on experience or experiments; the radial power factor of the second partition remains completely unchanged to ensure the continuity and stability of the power distribution in adjacent areas and avoid local power "jumps" caused by adjustments; for the third partition, by calculating the comprehensive compensation amount (which can also be determined based on experience or experiments), its power factor is appropriately increased to keep the total power of the core normalized. After the adjustment is completed, the power factor of the first partition decreases, the power factor of the third partition increases, and the second partition remains unchanged, realizing the redistribution of the core power factor. The updated power factors of each fuel rod assembly are mapped to the core subchannel model as input parameters for further calculations (such as deviation from nucleate boiling ratio DNBR prediction).

[0124] In this embodiment, by reducing the power factor of the first partition (the hottest sub-channel), the thermal power density of the area is effectively reduced, so that the local heat load pressure is relieved, thereby achieving the purpose of improving the cooling capacity. This adjustment method embodies the conservative principle, that is, in the safety analysis, the cooling performance of the key area of ​​the core is ensured to meet the safety standard by optimizing the power distribution. The power factor of the second partition remains unchanged, and the third partition is adjusted through compensation to ensure that the overall power distribution of the core meets the normalization constraint. Local performance improvement is achieved without affecting the overall power output of the core.

[0125] While keeping the total core power constant, the heat load distribution in the hottest area can be dynamically optimized to improve the local cooling performance of the core, while ensuring that the power adjustment has no adverse effect on the balance and stability of the entire core, effectively improving the accuracy and conservatism of safety analysis. The application of the three-zone power adjustment method can accurately simulate the power distribution characteristics of the core under the most unfavorable conditions and improve the accuracy of determining the burnout share of the fuel rods.

[0126] Embodiment 2:

[0127] This embodiment provides a method for assessing the consequences of a shaft jam accident in a nuclear power plant, the method comprising:

[0128] Determine the fuel rod burnout fraction of a nuclear power plant by using the fuel rod burnout fraction analysis method under a shaft jamming accident provided in any of the above embodiments;

[0129] The consequences of a shaft jam accident in a nuclear power plant are evaluated according to the proportion of fuel rod burnt, and an evaluation result is obtained. The evaluation result is used to indicate the accident safety level of the nuclear power plant.

[0130] Specifically (1) using the fuel rod burnout share analysis method provided above in this application under the shaft jam accident, determine the fuel rod burnout share of the nuclear power plant.

[0131] (2) Compare the fuel rod burnout fraction with the nuclear power plant safety analysis standards (such as the International Nuclear Event Scale (INES) or other regulatory standards) to determine the severity and safety level of the accident. The safety level assessment of a nuclear power plant is based on the direct impact of the accident on the physical damage to the core, among which the fuel rod burnout fraction is the most direct quantitative indicator used to objectively measure the severity of the consequences of the accident.

[0132] In one example, the severity and safety level of an incident may be classified into multiple levels (such as minor, acceptable, severe, or extremely adverse levels):

[0133] If the proportion of fuel rod burnt is lower than a certain threshold (such as <5%), it is a minor consequence; if the proportion of fuel rod damage reaches a medium range (such as 5% to 20%), it is a serious consequence; if the proportion is higher than a certain limit value (such as >20%), it may trigger a serious nuclear safety accident that needs to be reported.

[0134] In addition, on the basis of considering the proportion of fuel rod burnt, a comprehensive assessment can also be conducted in combination with other factors (such as the scale of coolant leakage, the amount of radiation released, the impact on the environment outside the plant, etc.) to form an overall safety level assessment report on the consequences of the shaft jam accident.

[0135] The safety level assessment results of the axle jamming accident can be used to guide decision-making and accident emergency response: providing institutional safety assessment

[0136] Based on the requirements of nuclear accident management, this assessment method provides a systematic classification of levels to facilitate regulatory agencies and nuclear power plant operators to formulate subsequent disposal plans.

[0137] Based on the requirements of nuclear accident management, this assessment method provides a systematic classification of levels, which is convenient for regulatory agencies and nuclear power plant operators to formulate subsequent disposal plans. For example: providing institutional safety assessment (based on the requirements of nuclear accident management, this assessment method provides a systematic classification of levels, which is convenient for regulatory agencies and nuclear power plant operators to formulate subsequent disposal plans); providing data for design and operation optimization (consequence assessment can expose the weak links of core design or safety system under specific accident scenarios, and provide a basis for nuclear power plant design improvements, such as optimizing power distribution, improving cooling system capacity or improving emergency response design); radiation consequence control support (through accurate quantitative assessment of consequences, it can guide radiation impact analysis after the accident, determine the scope of impact on the environment and personnel and response strategies), etc.

[0138] In this embodiment, the burnt fraction of fuel rods in a nuclear power plant shaft jam accident can be calculated to quantify the degree of physical damage to the core caused by the accident, and the consequences of the accident can be systematically evaluated in combination with nuclear safety level standards.

[0139] In order to facilitate the understanding of the fuel rod burnout fraction analysis method under the shaft jamming accident provided in this embodiment, a practical application description of the above method is provided here. Figure 2 For details, see the following example:

[0140] In response to the shaft jam accident in a pressurized water reactor nuclear power plant, based on the acceptance criteria for accident analysis, a clear calculation and analysis method for the fuel rod burnout fraction is proposed, which can provide important support for the accident consequence assessment. The specific steps are as follows:

[0141] Step S01, using a thermal hydraulic system analysis program to calculate a shaft-stuck accident transient process (ie, simulating a shaft-stuck accident transient process in a nuclear power plant).

[0142] Use a system analysis program (such as RELAP5 or TRACE) to model the nuclear power plant and construct a flow and heat exchange model of the reactor cooling system; input the initial conditions before the accident to simulate the transient response of the reactor after the shaft jam, including parameters such as pressure, flow and heat flux density; run the simulation to obtain the main thermal parameter curves that change with time.

[0143] The calculation gives the curves of the main thermal parameters such as reactor pressure, flow rate, heat flux density and so on changing with time after the shaft stuck accident, such as Figure 3 The curve parts shown in 1, 3, and 5.

[0144] Figure 3In the figure, the curve shown by number 1 is the pressurizer pressure curve after the shaft sticking accident, the curve shown by number 2 is the pressurizer pressure value corresponding to the minimum DNBR time point after the shaft sticking accident, the curve shown by number 3 is the core flow curve after the shaft sticking accident, the curve shown by number 4 is the core flow at the minimum DNBR time point after the shaft sticking accident, the curve shown by number 5 is the reactor heat flux density curve after the shaft sticking accident, and the curve shown by number 6 is the reactor heat flux density at the minimum DNBR time point after the shaft sticking accident.

[0145] This step simulates the rapid drop in coolant flow after a shaft jam, which will affect the heat transfer performance of the core, resulting in a pressure drop and a temperature rise. Through time segmentation, the changes in reactor parameters at each moment after the accident are obtained to ensure a dynamic understanding and evaluation of the reactor's thermal state.

[0146] Step S02, using the sub-channel program to complete the DNBR transient calculation under the shaft stuck accident (ie, according to the thermal parameters of the shaft stuck accident and the core sub-channel model, the DNBR variation curve of the shaft stuck accident is calculated).

[0147] The sub-channel program is used to model the nuclear power plant. The pressure, flow, and heat flux values ​​after the shaft-stuck accident shown in the above curves 1, 3, and 5 are written into the sub-channel program input model, and the DNBR transient change curve after the shaft-stuck accident is calculated. Figure 4 shown.

[0148] Figure 4 In the figure, the curve shown by number 7 is the curve of DNBR changing with time after the axle sticking accident, and the curve shown by number 8 is the time point when the minimum DNBR occurs after the axle sticking accident.

[0149] Specifically, the pressure, flow rate and heat flux density data obtained in step S01 are input into the sub-channel program model. The sub-channel model can analyze the fluid flow state and heat transfer state of each channel in detail and generate a DNBR change curve over time.

[0150] The subchannel program uses a multi-channel flow and heat transfer model to fully evaluate the interaction between the phase change of the fluid and the heat channel under transient conditions. By calculating the DNBR in different cooling channels, it is determined whether deviation from nucleate boiling (DNB) occurs.

[0151] Step S03, determining the minimum DNBR time point during the accident process and its corresponding transient parameters.

[0152] The transient change curve of DNBR after the axle jamming accident calculated by S02 ( Figure 4 ), determine the time point when the minimum DNBR appears, that is, Figure 4 The time point indicated by the middle label 8. The pressure, flow rate, and heat flux density corresponding to this moment are extracted and determined, namely Figure 3The values ​​corresponding to this moment in the curves shown by numbers 2, 4, and 6.

[0153] Step S04, determining a critical power peak factor.

[0154] The following operations are performed in the input model of the subchannel program to determine the critical power crest factor.

[0155] First, the pressure, flow rate, heat flux and other parameter values ​​at the time when the minimum DNBR occurs in the axle sticking accident calculated in step S03 are used as the initial working condition for the sub-channel program calculation.

[0156] Furthermore, if Figure 5 As shown in the figure, the radial power factor of the hottest channel in the thermal component of the reduced sub-channel model ( Figure 5 9, 10), calculate the adjusted thermal component power factor ( Figure 5 11). Again, to ensure the normalization of the power of the entire core, the radial power of other components other than the hot components is adjusted. The "three-zone power adjustment" method is used during the adjustment process. The hottest component is used as zone 1, and its radial power factor is kept reduced. The circle of components adjacent to the hottest component is used as zone 2, and its radial power factor remains unchanged. The other components of the core are used as zone 3, and their radial power factors are increased to ensure that the power factor obtained by weighted calculation of each component of the entire core is 1.0. Finally, after completing the first round of adjustment of the power factor in the sub-channel program, the DNBR calculation is re-performed, and the calculated minimum DNBR is compared with the safety analysis limit. If the DNBR value obtained is exactly equal to the limit, it can be determined that the radial power factor of the hottest channel of the entire core at this time is the critical power peak factor. If the calculated DNBR value is still higher than the limit, step S04 is repeated, and the power factor is adjusted again and the DNBR calculation is performed until the critical power peak factor is determined.

[0157] The critical power peak factor reflects the cooling capacity under the most unfavorable conditions and directly affects the risk of fuel rod burnout. By adjusting the power distribution to ensure the conservatism of the analysis results, the power distribution under the critical state is simulated under the condition of coolant flow attenuation.

[0158] Step S05, determining the proportion of fuel rod burnt after the shaft jam accident through the fuel statistical curve.

[0159] The nuclear power plant fuel statistical curve provided by the physical design data of the nuclear power plant, such as Figure 6 A nuclear power plant fuel statistical curve is shown. According to the critical power peak factor determined in step S04, the fuel rod share corresponding to this power factor is determined from the fuel statistical curve, and this share is the fuel rod share that burns after the shaft jam accident.

[0160] The fuel statistical curve shows the expected proportion of fuel burnout at a specific power factor, providing a cooling capacity indicator that can be used for accident consequence analysis and assessment. By comparing the critical power peak factor with the statistical curve, the proportion of fuel rods that have been burned can be accurately identified.

[0161] The method for calculating the fuel rod burnout fraction provided in this embodiment dynamically simulates the transient process of the shaft sticking accident through a thermal-hydraulic system analysis program and a sub-channel program, determines the minimum DNBR time point and its corresponding parameters such as pressure, flow, and heat flux density, and ensures that the analysis covers the most severe working conditions of the accident; determines the critical power peak factor of the core by adjusting the power factor and matching the minimum DNBR limit, and quantifies the influence of the core power distribution on the cooling capacity of the fuel rods; calculates the fuel rod burnout fraction in combination with the fuel statistical curve, and provides a quantitative basis for the evaluation of the consequences of the accident.

[0162] It can effectively analyze and evaluate the number of core fuel rods burned after a shaft jam accident in a nuclear power plant, and provide support and basis for whether the post-accident safety acceptance criteria are met and the evaluation of the radioactive consequences of the accident.

[0163] Embodiment 3:

[0164] like Figure 7 As shown, this embodiment provides a fuel rod burnout fraction analysis device 700 in the event of a shaft jam, comprising:

[0165] The first determination module 701 is used to simulate the transient process of the shaft sticking accident of the nuclear power plant and determine the transient operating condition parameters corresponding to the time point of the minimum transient DNBR under the shaft sticking accident;

[0166] The second determination module 702 is connected to the first determination module 701 and is used to determine the initial thermal hydraulic state of the core of the nuclear power plant according to the transient operating condition parameters when the minimum transient DNBR is less than the safety limit;

[0167] The third determination module 703 is connected to the second determination module 702 and is used to adjust the radial power factor of the hottest sub-channel of the sub-channel model according to the initial thermal hydraulic state and the pre-established core sub-channel model, and determine the critical power peak factor, where the critical power peak factor is the power peak factor when the minimum transient DNBR of the hottest sub-channel is equal to the safety limit;

[0168] The analysis module 704 is connected to the third determination module 703 and is used to determine the fuel rod burnout ratio that will cause DNB in ​​the event of a nuclear power plant shaft jam from the fuel statistical curve according to the critical power peak factor.

[0169] Optionally, the second determining module 702 includes:

[0170] The first creation unit is used to model the nuclear power plant using a thermal hydraulic system analysis program to obtain a simulation model for a stuck shaft accident;

[0171] The first determination unit is connected to the first creation unit and is used to simulate the reactor transient process after the shaft sticking accident occurs according to the initial operating parameters of the nuclear power plant and the shaft sticking accident simulation model to obtain the thermal parameters of the shaft sticking accident;

[0172] A second creation unit is used to establish a core sub-channel model using a sub-channel program;

[0173] The first calculation unit is connected to the first determination unit and the second creation unit respectively, and is used to calculate the DNBR variation curve of the stuck shaft accident according to the thermal parameters of the stuck shaft accident and the core sub-channel model;

[0174] A second determination unit, connected to the calculation unit, is used to determine the time point of the minimum transient DNBR under the axle jam accident according to the DNBR variation curve of the axle jam accident;

[0175] The third determining unit is connected to the second determining unit, and is used to determine the transient operating condition parameters corresponding to the time point of the minimum transient DNBR among the thermal parameters of the stuck shaft accident.

[0176] Optionally, the third determining module 703 includes:

[0177] A second calculation unit is used to input the initial thermal hydraulic state into the core sub-channel model;

[0178] A third calculation unit is used to reduce the radial power factor of the hottest sub-channel in the core sub-channel model by a preset unit to obtain an updated radial power factor of the hottest sub-channel;

[0179] A fourth calculation unit is used to perform a three-zone power adjustment method to adjust the radial power factors of other channels according to the updated radial power factor of the hottest sub-channel to obtain a radial power factor of each sub-channel;

[0180] A fifth calculation unit is used to calculate a predicted DNBR variation curve of the core according to the radial power factor of each sub-channel;

[0181] A sixth calculation unit, configured to determine a minimum predicted DNBR based on the predicted DNBR change curve, wherein the minimum predicted DNBR is a DNBR with the smallest value among the DNBR data of all sub-channels;

[0182] The seventh calculation unit is used to control the repeated operation of the second calculation unit, the third calculation unit, the fourth calculation unit, the fifth calculation unit and the sixth calculation unit when the minimum predicted DNBR is greater than the safety limit, until the minimum predicted DNBR is not less than the safety limit, and determine the updated radial power factor of the hottest sub-channel as the critical power peak factor.

[0183] Optionally, the fourth calculation unit specifically includes:

[0184] a dividing subunit, used to divide the fuel rod assemblies in the core into a first partition, a second partition and a third partition, wherein the first partition includes the fuel rod assemblies corresponding to the hottest subchannel, the second partition includes the fuel rod assemblies adjacent to the fuel rod assemblies in the first partition, and the third partition includes other fuel rod assemblies except the fuel rod assemblies in the first partition and the second partition;

[0185] A first adjustment subunit, used to keep the radial power factor of the combustion rod assembly in the second partition unchanged;

[0186] The second adjustment subunit is used to increase the radial power factor of the burner rod assembly in the third partition so that the power factor obtained by weighted calculation of each burner rod assembly in the core is 1, thereby obtaining an updated radial power factor of each burner rod assembly in the third partition.

[0187] In the fuel rod burnout fraction analysis device under the shaft jam accident of this embodiment, the cooling condition corresponding to the minimum transient DNBR of the core during the shaft jam accident can be captured through transient simulation; then the power distribution can be finely adjusted in combination with the sub-channel model to accurately determine the critical power peak factor; thus, the fuel rod burnout fraction can be accurately quantified using the fuel statistical curve. Thus, the accuracy of the fuel rod burnout fraction analysis under the shaft jam accident is improved.

[0188] Embodiment 4:

[0189] This embodiment provides a consequence assessment system for a shaft jam accident in a nuclear power plant, the system comprising:

[0190] The device for analyzing the burnout fraction of fuel rods in a shaft-stuck accident provided in any of the above embodiments is used to determine the burnout fraction of fuel rods in a nuclear power plant;

[0191] The evaluation module is connected to the fuel rod burnout fraction analysis device under the shaft jam accident, and is used to evaluate the consequences of the shaft jam accident in the nuclear power plant according to the fuel rod burnout fraction, and obtain the evaluation result, which is used to indicate the accident safety level of the nuclear power plant.

[0192] It should be clear that the present application is not limited to the specific configuration and processing described above and shown in the figures. For the sake of simplicity, a detailed description of the known method is omitted here. In the above embodiments, several specific steps are described and shown as examples. However, the method process of the present application is not limited to the specific steps described and shown, and those skilled in the art can make various changes, modifications and additions, or change the order between the steps after understanding the spirit of the present application.

[0193] The functional blocks shown in the above block diagram can be implemented as hardware, software, firmware or a combination thereof. When implemented in hardware, it can be, for example, an electronic circuit, an application specific integrated circuit (ASIC), appropriate firmware, a plug-in, a function card, etc. When implemented in software, the elements of the present application are programs or code segments that are used to perform the required tasks. The program or code segment can be stored in a machine-readable medium, or transmitted on a transmission medium or a communication link by a data signal carried in a carrier wave. "Machine-readable medium" can include any medium capable of storing or transmitting information. Examples of machine-readable media include electronic circuits, semiconductor memory devices, ROM, flash memory, erasable ROM (EROM), floppy disks, CD-ROMs, optical disks, hard disks, optical fiber media, radio frequency (RF) links, etc. The code segment can be downloaded via a computer network such as the Internet, an intranet, etc.

[0194] It should also be noted that the exemplary embodiments mentioned in this application describe some methods or systems based on a series of steps or devices. However, this application is not limited to the order of the above steps, that is, the steps can be performed in the order mentioned in the embodiment, or in a different order from the embodiment, or several steps can be performed simultaneously.

[0195] Aspects of the present disclosure are described above with reference to the flowchart and / or block diagram of the method, device (system) and computer program product according to the embodiment of the present disclosure. It should be understood that each box in the flowchart and / or block diagram and the combination of each box in the flowchart and / or block diagram can be implemented by computer program instructions. These computer program instructions can be provided to a processor of a general-purpose computer, a special-purpose computer, or other programmable scheduling device to produce a machine so that these instructions executed by the processor of the computer or other programmable scheduling device enable the implementation of the function / action specified in one or more boxes of the flowchart and / or block diagram. Such a processor can be, but is not limited to, a general-purpose processor, a special-purpose processor, a special application processor, or a field programmable logic circuit. It can also be understood that each box in the block diagram and / or flowchart and the combination of boxes in the block diagram and / or flowchart can also be implemented by dedicated hardware that performs a specified function or action, or can be implemented by a combination of dedicated hardware and computer instructions.

[0196] The above are only specific implementation methods of the present application. Those skilled in the art can clearly understand that for the convenience and simplicity of description, the specific working processes of the systems, modules and units described above can refer to the corresponding processes in the aforementioned method embodiments, and will not be repeated here. It should be understood that the protection scope of the present application is not limited to this. Any technician familiar with the technical field can easily think of various equivalent modifications or replacements within the technical scope disclosed in this application, and these modifications or replacements should be included in the protection scope of this application.

Claims

1. A method for analyzing the fuel rod burnout fraction in a shaft jam accident, characterized in that: include: Simulate the transient process of shaft-stuck accidents in nuclear power plants and determine the transient operating parameters corresponding to the time point of the minimum transient DNBR under shaft-stuck accidents; When the minimum transient DNBR is less than a safety limit, determining an initial thermal-hydraulic state of the core of the nuclear power plant according to the transient operating condition parameters; According to the initial thermal-hydraulic state and the pre-established core sub-channel model, the radial power factor of the hottest sub-channel of the core sub-channel model is adjusted to determine the critical power peak factor, wherein the critical power peak factor is the power peak factor when the minimum transient DNBR of the hottest sub-channel is equal to the safety limit value; According to the critical power peak factor, the fuel rod burnout proportion that will cause DNB in ​​the event of a shaft jam in the nuclear power plant is determined from a fuel statistical curve.

2. The method according to claim 1, characterized in that The transient process of the shaft-stuck accident in the nuclear power plant is simulated to determine the transient operating parameters corresponding to the time point of the minimum transient DNBR under the shaft-stuck accident, specifically including: The nuclear power plant is modeled using a thermal hydraulic system analysis program to obtain a shaft jamming accident simulation model; According to the initial operating parameters of the nuclear power plant and the shaft-stuck accident simulation model, the transient process of the reactor after the shaft-stuck accident occurs is simulated to obtain thermal parameters of the shaft-stuck accident; The core subchannel model is established using the subchannel program; According to the thermal parameters of the shaft-stuck accident and the core sub-channel model, a DNBR variation curve of the shaft-stuck accident is calculated; Determine the time point of the minimum transient DNBR under the axle sticking accident according to the axle sticking accident DNBR variation curve; Determine the transient operating condition parameters corresponding to the time point of the minimum transient DNBR in the thermal parameters of the stuck shaft accident.

3. The method according to claim 1, characterized in that: The step of adjusting the radial power factor of the hottest sub-channel of the sub-channel model according to the initial thermal hydraulic state and the pre-established core sub-channel model to determine the critical power peak factor specifically includes: S1, inputting the initial thermal hydraulic state into the core sub-channel model; S2, reducing the radial power factor of the hottest subchannel in the core subchannel model by a preset unit to obtain an updated radial power factor of the hottest subchannel; S3, using a three-zone power adjustment method, according to the updated radial power factor of the hottest sub-channel, adjusting the radial power factors of other channels to obtain a radial power factor of each sub-channel; S4. Calculate the predicted DNBR variation curve of the core according to the radial power factor of each sub-channel; S5. Determine a minimum predicted DNBR based on the predicted DNBR change curve, where the minimum predicted DNBR is the DNBR with the smallest value among the DNBR data of all sub-channels; S6. When the minimum predicted DNBR is greater than the safety limit, return to execute steps S2 to S5 until the minimum predicted DNBR is not less than the safety limit, and determine the updated hottest sub-channel radial power factor as the critical power peak factor.

4. The method according to claim 3, characterized in that The three-zone power adjustment method is used to adjust the radial power factors of other channels according to the updated radial power factor of the hottest sub-channel to obtain the radial power factor of each sub-channel, specifically including: Dividing the fuel rod assemblies in the core into a first partition, a second partition and a third partition, wherein the first partition includes the fuel rod assemblies corresponding to the hottest subchannel, the second partition includes the fuel rod assemblies adjacent to the fuel rod assemblies in the first partition, and the third partition includes other fuel rod assemblies except the fuel rod assemblies in the first partition and the second partition; maintaining the radial power factor of the burner rod assembly in the second partition unchanged; The radial power factor of the burner rod assembly in the third partition is increased so that the power factor obtained by weighted calculation of each burner rod assembly in the core is 1, thereby obtaining an updated radial power factor of each burner rod assembly in the third partition.

5. A method for assessing the consequences of a shaft jam accident in a nuclear power plant, characterized in that: include: Determine the fuel rod burnout fraction of the nuclear power plant by using the fuel rod burnout fraction analysis method under a shaft jam accident according to any one of claims 1 to 4; The consequence of the shaft-stuck accident in the nuclear power plant is evaluated according to the fuel rod burnout fraction to obtain an evaluation result, and the evaluation result is used to indicate the accident safety level of the nuclear power plant.

6. A fuel rod burnout fraction analysis device in the event of a shaft jam, characterized in that: The device comprises: The first determination module is used to simulate the transient process of the shaft sticking accident of the nuclear power plant and determine the transient operating condition parameters corresponding to the time point of the minimum transient DNBR under the shaft sticking accident; A second determination module, connected to the first determination module, is used to determine the initial thermal-hydraulic state of the core of the nuclear power plant according to the transient operating condition parameters when the minimum transient DNBR is less than the safety limit; a third determination module, connected to the second determination module, for adjusting the radial power factor of the hottest sub-channel of the sub-channel model according to the initial thermal-hydraulic state and the pre-established core sub-channel model, and determining a critical power peak factor, wherein the critical power peak factor is the power peak factor when the minimum transient DNBR of the hottest sub-channel is equal to the safety limit; The analysis module is connected to the third determination module and is used to determine the fuel rod burnout proportion of DNB occurring in the nuclear power plant shaft jam accident from the fuel statistical curve according to the critical power peak factor.

7. The device according to claim 6, characterized in that The second determining module includes: The first creation unit is used to model the nuclear power plant using a thermal hydraulic system analysis program to obtain a shaft jamming accident simulation model; a first determining unit connected to the first creating unit, configured to simulate a reactor transient process after a shaft-stuck accident occurs according to the initial operating parameters of the nuclear power plant and the shaft-stuck accident simulation model, and obtain thermal parameters of the shaft-stuck accident; A second creation unit is used to establish a core sub-channel model using a sub-channel program; A first calculation unit is connected to the first determination unit and the second creation unit respectively, and is used to calculate a DNBR variation curve of a stuck shaft accident according to the thermal parameters of the stuck shaft accident and the core sub-channel model; A second determination unit, connected to the calculation unit, is used to determine the time point of the minimum transient DNBR under the axle-stuck accident according to the axle-stuck accident DNBR change curve; The third determining unit is connected to the second determining unit, and is used to determine the transient operating condition parameters corresponding to the time point of the minimum transient DNBR in the thermal parameters of the stuck shaft accident.

8. The device according to claim 7, characterized in that The third determination module specifically includes: A second calculation unit, used for inputting the initial thermal-hydraulic state into the core sub-channel model; A third calculation unit is used to reduce the radial power factor of the hottest sub-channel in the core sub-channel model by a preset unit to obtain an updated radial power factor of the hottest sub-channel; A fourth calculation unit is used to perform a three-zone power adjustment method to adjust the radial power factors of other channels according to the updated radial power factor of the hottest sub-channel to obtain a radial power factor of each sub-channel; A fifth calculation unit is used to calculate a predicted DNBR variation curve of the core according to the radial power factor of each sub-channel; A sixth calculation unit, configured to determine a minimum predicted DNBR based on the predicted DNBR change curve, wherein the minimum predicted DNBR is a DNBR with the smallest value among the DNBR data of all sub-channels; The seventh calculation unit is used to control the repeated operation of the second calculation unit, the third calculation unit, the fourth calculation unit, the fifth calculation unit and the sixth calculation unit when the minimum predicted DNBR is greater than the safety limit, until the minimum predicted DNBR is not less than the safety limit, and determine the updated hottest sub-channel radial power factor as the critical power peak factor.

9. The device according to claim 8, characterized in that The fourth calculation unit specifically includes: a dividing subunit, used to divide the fuel rod assemblies in the core into a first partition, a second partition and a third partition, wherein the first partition includes the fuel rod assemblies corresponding to the hottest subchannel, the second partition includes the fuel rod assemblies adjacent to the fuel rod assemblies in the first partition, and the third partition includes other fuel rod assemblies except the fuel rod assemblies in the first partition and the second partition; A first adjustment subunit, used to keep the radial power factor of the combustion rod assembly in the second partition unchanged; The second adjustment subunit is used to increase the radial power factor of the burner rod assembly in the third partition so that the power factor obtained by weighted calculation of each burner rod assembly in the core is 1, and obtain the updated radial power factor of each burner rod assembly in the third partition.

10. A consequence assessment system for a shaft jam accident in a nuclear power plant, characterized in that: The system comprises: The device for analyzing the burnout fraction of fuel rods in a shaft-stuck accident according to any one of claims 6 to 9, used for determining the burnout fraction of fuel rods in the nuclear power plant; The evaluation module is connected to the fuel rod burnout fraction analysis device under the shaft jamming accident, and is used to evaluate the consequences of the shaft jamming accident in the nuclear power plant according to the fuel rod burnout fraction to obtain an evaluation result, and the evaluation result is used to indicate the accident safety level of the nuclear power plant.

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