Rod falling accident safety analysis method and calculation device

Abstract

A rod falling accident safety analysis method comprises the following steps: carrying out envelope analysis on a rod falling accident publicly, and determining the range of rod falling value; establishing a feature rod falling value sequence; calculating to obtain a limit state point; and according to the reactor core axial power distribution and the radial power peak factor of the limit state point, calculating to obtain the reactor core deviation nucleation boiling ratio. According to the method, reasonable and conservative safety analysis can be carried out on the rod falling accident, the rod falling accident safety analysis efficiency is improved, and meanwhile the operation economy of the nuclear power plant is optimized. The invention further provides a computing device.

Description

Technical Field

[0001] The invention belongs to the field of nuclear power, and in particular relates to a rod drop accident safety analysis method and a calculation device. Background Art

[0002] A control rod drop accident will cause the core power level to drop in the initial stage of the accident and disturb the radial power distribution of the reactor. After the control rod drop occurs, the reduction in the core power level and the continued demand for steam by the turbine will cause the reactor coolant temperature to drop. If the temperature of the reactor cooling system drops too much, the turbine power may not be maintained. In this case, the power plant control system in the automatic control mode detects the drop in the core power level and may increase the adjustment assistance, which will lead to the occurrence of power overshoot. Power overshoot and the increase in peak factor caused by the change in the position of the control rods may eventually cause the surface of the fuel cladding to burn. Therefore, in the accident safety analysis, it is necessary to focus on demonstrating whether the core consequences caused by power overshoot under the rod drop accident condition are beyond the acceptable level. The existing analysis analyzes the case of a bundle of control rods falling and the case of a control rod subgroup falling separately, and uses the boundary conditions of the limit envelope for evaluation, which makes the accident consequence analysis too conservative. Providing a more reasonable rod drop accident safety analysis method is of positive significance for improving the operating capacity and economy of power plants while ensuring the safety of power plants. Summary of the invention

[0003] The purpose of the present invention is to provide a safety analysis method for a rod drop accident, so as to improve the accuracy of the consequence analysis of the rod drop accident. The present invention also provides a calculation device.

[0004] According to an embodiment of one aspect of the present invention, a rod drop accident safety analysis method is provided, the method comprising the following steps:

[0005] Step a): Provide a nuclear power plant accident analysis model, conduct a comprehensive reactor full cycle and full life cycle physical analysis for a given rod drop accident condition, and determine the range of the rod drop value;

[0006] Step b): within the maximum drop value range, select drop value points at intervals of a given drop value step length to form a characteristic drop value sequence;

[0007] Step c): Calculate the core axial power distribution and radial power peak factor F at each rod drop value point in the characteristic rod drop value sequence ΔH N; Using a safety analysis program to conduct accident analysis on the rod drop value points in the characteristic rod drop value sequence, and obtain a rod drop condition analysis result sequence, wherein the rod drop condition analysis result sequence includes the thermal-hydraulic parameters corresponding to each rod drop value point in the characteristic rod drop value sequence, and the limit state point is calculated based on the thermal-hydraulic parameters, and the limit state point is the point where the ratio of the average heat flux density of the core to the core flow rate is the largest;

[0008] Step d): According to the core axial power distribution and radial power peak factor F at the limit state point ΔH N Calculate the core deviation from nucleate boiling ratio.

[0009] Through the above method, the evaluation of the core deviation from nucleate boiling ratio (DNBR) can be carried out reasonably, the safety analysis margin can be released, and the economic efficiency of the power plant can be improved; at the same time, it can guide the optimization of the power plant safety management plan and improve the safety of the power plant.

[0010] Furthermore, in some embodiments, when the analyzed reactor has a rod-lifting prohibition protection logic, the step of optimizing the rod-lifting prohibition protection constant is also included: performing repeated sensitivity analysis calculations through the steps c) and d) to obtain F ΔH N The prohibition of lifting the rod protection setting value.

[0011] Furthermore, in some embodiments, the step size of the rod drop value in step b) is 5 pcm-15 pcm.

[0012] Furthermore, in some embodiments, the safety analysis program in step c) is configured as RELAP5.

[0013] Furthermore, in some embodiments, in the step c), the accident analysis includes calculating rod drop accident conditions under different control rod differential values, moderator temperature coefficients and rod lifting values ​​to obtain thermal-hydraulic parameters.

[0014] Furthermore, in some embodiments, the thermal-hydraulic parameters include core average heat flux density, core flow rate, reactor power and coolant temperature.

[0015] Furthermore, in some embodiments, in the step c), the process of calculating the limit state point also includes a reliability verification step, and the reliability verification step selects a plurality of the rod drop value points close to the limit state point for verification calculation.

[0016] According to an embodiment of another aspect of the present invention, there is provided a computing device, which includes a memory and a processor, wherein the memory stores a computing program, and when the computing program is executed by the processor, the rod drop accident safety analysis method provided in any of the aforementioned embodiments can be implemented. BRIEF DESCRIPTION OF THE DRAWINGS

[0017] Figure 1 4 is a flow chart of a rod drop accident safety analysis method in one embodiment.

[0018] The purpose of the above drawings is to explain the present invention in detail so that those skilled in the art can understand the technical concept of the present invention, but it is not intended to limit the present invention. DETAILED DESCRIPTION

[0019] The present invention is further described in detail below through specific embodiments in conjunction with the accompanying drawings.

[0020] Reference to "embodiments" herein means that a particular feature, structure, or characteristic described in conjunction with the embodiments may be included in at least one embodiment of the present invention. The phrase appearing in various locations in the specification does not necessarily refer to the same embodiment, nor is it limited to mutually exclusive independent or alternative embodiments. Those skilled in the art should be able to understand that the embodiments herein may be combined with other embodiments without causing structural conflicts.

[0021] In the description herein, "plurality" means at least two.

[0022] In a reactor rod drop accident, one or more control rods from the same subgroup fall, causing the core power level to drop in the initial stage of the accident and disturbing the radial power distribution in the core. Depending on the reactivity value of the dropped control rods, the drop of control rods in some accident conditions may cause the allowable design power peak factor to be exceeded. After the control rod drop, the core power level decreases but the demand for steam from the steam turbine does not change immediately, which will cause the reactor coolant temperature to drop. If the reactor is in manual control mode, due to reactivity feedback, the equilibrium condition can be re-established, and the negative moderator temperature coefficient causes the reactor power to increase monotonically and return to the initial power level, which will only cause the inlet temperature to drop without power overshoot.

[0023] However, in the automatic operation mode, if the temperature of the reactor cooling system drops significantly, the turbine power may not be maintained due to the decrease in secondary steam pressure and the limitation of the volume flow of the turbine system. In this case, the equilibrium power level is less than the turbine power level. In the automatic control mode, the power plant control system detects the decrease in the core power level and will increase the power of the regulating rod group and then insert the regulating rod group to return the power of the power plant to the initial level. In this process, power overshoot may occur. When power overshoot occurs and the position of the control rod changes, resulting in an increase in the peak factor, it may cause the fuel cladding surface to burn, that is, the core may deviate from nucleate boiling (DNB) phenomenon, causing an accident.

[0024] Therefore, in the accident safety analysis, it is necessary to focus on proving whether the core consequences caused by power overshoot under the rod drop accident condition are beyond the acceptable level.

[0025] Whether power overshoot occurs and the extent of power overshoot depends on the characteristics of the power plant control system, the core reactivity coefficient, the value of dropped rods, the value of available regulating rod groups and the design of the power plant protection system.

[0026] Existing analyses of rod-drop accidents usually use the boundary conditions of the limit envelope (rod-drop value, radial power peak factor, and reactivity feedback coefficient). The conclusions obtained by this analysis method are usually too conservative. It can neither accurately reflect the actual operation of the power plant to more accurately assess the risk of rod-drop accidents, nor is it conducive to optimizing the economic efficiency of the power plant.

[0027] In order to overcome the above problems, the embodiments of the present invention provide a rod drop accident safety analysis method, which can quickly carry out the consequence evaluation of the rod drop accident through a more convenient, efficient and reasonably conservative analysis method, and improve the operation capacity of the power plant and optimize the economic efficiency under the premise of ensuring the safety of the power plant. The implementation process of the method is as follows: Figure 1 As shown, the following steps are included:

[0028] Step a): Determine the drop value range.

[0029] Firstly, for the accident conditions of dropping a bundle of control rods and dropping a subgroup of control rods, analytical models are used to carry out enveloped physical analysis of the reactor's full cycle and life cycle to confirm the range of the rod drop value.

[0030] Step b): Establish a feature drop value sequence.

[0031] Within the maximum drop value range, a series of drop value points are selected at intervals of a certain drop value step length to form a characteristic drop value sequence. In a preferred embodiment, the drop value step length can be set to 5pcm-15pcm, and in a further preferred embodiment, 10pcm can be selected.

[0032] Step c): Calculate and determine the limit state point.

[0033] Next, physical analysis is carried out for each rod drop value point in the characteristic rod drop value sequence, and mature analysis models and methods are used to determine the core axial power distribution and radial power peak factor F at each rod drop value point. ΔH N .

[0034] Use a suitable safety analysis program to conduct a rod drop accident analysis for each rod drop value point. In a preferred embodiment, the safety analysis program can use RELKAP5, etc. Specifically, for each rod drop value point, different control rod differential values, moderator temperature coefficients and rod lifting values ​​are arranged and combined to obtain different rod drop accident conditions. Each rod drop accident condition is calculated by the program to obtain analysis results, and the analysis results include changes in the main thermal-hydraulic parameters during the rod drop transient process. Among them, the thermal-hydraulic parameters include the average core heat flux density, core flow rate, reactor power, coolant temperature, etc. The calculated analysis results constitute a state point, and all state points constitute a rod drop condition analysis result sequence.

[0035] From the sequence of analysis results, find the limit state point. The limit state point is the point where the ratio of the average heat flux density of the core to the core flow rate is the largest. The larger the ratio of the average heat flux density of the core to the core flow rate, the higher the risk of core burnout. Therefore, the limit state point is the point with the greatest risk of a core burnout accident. In a preferred embodiment, the step of performing reliability verification on the limit state point is also included. In the reliability verification, multiple state points near the limit state point are selected (state points with slightly different parameters such as control rod differential value, moderator temperature coefficient, rod lifting value, etc., and the specific difference can be adjusted according to the actual state), and verification calculations are performed to determine that the limit state point has selected the most extreme state.

[0036] After a rod drop accident occurs, the control rods fall and introduce negative reactivity. The reactivity control system will automatically lift the rods to control the reactor power. Due to the automatic rod lifting and reactivity feedback characteristics of the reactivity control system, power overshoot may occur during the rod drop process due to lifting the rods too quickly or too much.

[0037] For reactors with protection logic prohibiting control rod lifting, set protection logic prohibiting control rod lifting, such as adding radial power peak factor F ΔH N The monitoring signal will be greater than a certain set value, and the control rod will be prohibited from automatically lifting. ΔH NIt is automatically calculated by the core detection system of the power plant. In the rod drop accident analysis, simulating this function can reduce the rod drop accident analysis conditions. When the rod drop value is large enough, it will cause the reactor power to drop rapidly, thereby triggering the control rod lifting prohibition logic. Under this condition, the power overshoot is small and not extreme, and no analysis is required. This function can also reduce the conservatism of the uncertainty effect of power measurement. The uncertainty effect factor of power measurement is a multiplication factor of the nuclear power signal. In one embodiment, 0.6 is used in the calculation. After the rod drop, the control system will consider the power to be reduced by 40%. If the uncertainty effect factor of the power measurement is small enough, it is possible that the rod lifting prohibition protection signal will always be triggered, unless the rod lifting prohibition protection signal setting value is increased. Therefore, for a given rod drop value, the rod lifting prohibition protection signal setting value limits the value of the uncertainty effect factor of the power measurement.

[0038] For a specific rod drop value condition, the uncertainty effect factor data of the smaller power measurement is increased until the specific rod drop value condition searched for does not trigger the rod lifting protection signal, and the rod lifting protection signal is triggered when the specific rod drop value condition searched for is greater than the specific rod drop value condition searched. In addition, other rod lifting values ​​are analyzed until the ratio of the peak heat flux density to the core flow rate remains stable and no longer increases. Although each repeated calculation will generate a set of state points of the rod drop value condition, only the state points of the specific rod drop value calculation are valid.

[0039] For other control rod differential values, rod drop values, and rod drop conditions of moderator temperature coefficient (MTC), the above process is repeated, that is, different control rod differential values, rod drop values, and moderator temperature coefficients are selected to form new rod drop accident conditions, and the state points corresponding to different rod drop accident conditions are obtained by searching the state points according to the above method of adjusting the uncertainty effect factor of power measurement.

[0040] In one embodiment, the rod drop condition is: the control rod differential value is 8pcm / step, the rod drop value is 100pcm, the MTC is 0pcm / ℃, and the rod lifting value is 100pcm. Assuming that when the uncertainty effect factor of the power measurement is increased to 0.9, only the condition with a rod drop value less than 100pcm will not trigger the rod lifting prohibition protection signal, while the condition with a rod drop value greater than 100pcm will trigger the rod lifting prohibition protection signal. Then, different rod lifting values ​​(200pcm, 210pcm, etc.) are analyzed until the peak heat flux density no longer increases with the increase of the rod lifting value. Through the above process, the effective state point corresponding to the rod drop condition with a control rod differential value of 8pcm / step, a rod drop value of 100pcm, and an MTC of 0pcm / ℃ is obtained. Afterwards, the above process is repeated for different combinations of control rod differential values, MTC and rod drop values, and finally a complete rod drop state point table is obtained.

[0041] For reactor types without rod lifting prohibition protection logic, the main focus is on operating conditions that do not trigger the reactor shutdown signal. The reactor shutdown signal may be an overpower ΔT reactor shutdown signal, an overtemperature ΔT reactor shutdown signal, or a pressurizer low pressure reactor shutdown signal.

[0042] In this accident, if the shutdown is triggered by a shutdown signal (which may be an overpower ΔT reactor shutdown signal, an overtemperature ΔT reactor shutdown signal, or a pressurizer low pressure reactor shutdown signal), the accident is not extreme. Conservatively considering that the reactor shutdown signal is not triggered, after the control rods fall, the power level of the core will be reestablished due to reactivity feedback or the lifting of the control rod group. In the manual control mode, without the participation of the reactor control system, the process of restoring balance is monotonous; in the automatic control mode, due to the participation of the reactor control system, overshoot may occur. Therefore, a bunch of control rods falling accident in the automatic control mode is the extreme case of a bunch of control rods falling accident. The power measurement uncertainty effect, that is, the uncertainty effect factor of power measurement, also needs to be considered in the analysis. For a given rod drop value, different shutdown signal setting values ​​limit the value of the uncertainty effect factor of power measurement.

[0043] In another embodiment, the rod drop condition is: the control rod differential value is 8pcm / step, the rod drop value is 100pcm, the MTC is 0pcm / ℃, and the rod lifting value is 100pcm. Assuming that when the uncertainty effect factor of the power measurement is increased to 0.9, only the condition with a rod drop value less than 100pcm will not trigger the reactor shutdown signal, while the condition with a rod drop value greater than 100pcm will trigger the reactor shutdown signal. Then, different rod lifting values ​​(200pcm, 210pcm, etc.) are analyzed until the peak heat flux no longer increases with the increase of the rod lifting value. Through the above process, the effective state point corresponding to the rod drop condition with a control rod differential value of 8pcm / step, a rod drop value of 100pcm, and an MTC of 0pcm / ℃ is obtained. Afterwards, the above process is repeated for different combinations of control rod differential values, MTC and rod drop values, and finally a complete rod drop state point table is obtained.

[0044] Based on the analysis steps and logic in the above two cases, calculations are carried out according to different control rod differential values, rod drop values, rod lifting values, different reactive feedbacks, and the presence or absence of rod lifting protection signals to obtain a large number of state points, and a state point screening model (script or mini program) is established. After obtaining the batch state points, the most extreme state point is quickly found (in order to improve the credibility of the analysis, multiple state points near the limit point are usually selected).

[0045] Step d): Calculate the core deviation from nucleate boiling ratio (DNBR).

[0046] Next, select an appropriate core subchannel analysis program and calculate the core axial power distribution and radial power peak factor F at the limit state point determined in the previous steps. ΔH N Calculate the core deviation from nucleate boiling ratio (DNBR).

[0047] In a preferred embodiment, for a reactor type with a rod lifting prohibition protection logic, a step of optimizing the rod lifting prohibition protection constant is also included. If the protection constant is set too low, the constant will be reached very quickly during the transient process, and the control rods will be prohibited from lifting, which results in an overly conservative safety strategy and is not conducive to the optimization of the power plant's operating economy; and if the protection constant is set too high, the constant will be reached later during the transient process, and the risk of core burnout increases; by optimizing the rod lifting prohibition protection constant, it is possible to meet both the safety of the power plant and the economic efficiency of the power plant's operation. This step is achieved by repeating the sensitivity analysis of the aforementioned steps c) and d), and calculating the radial power peak factor F ΔH N The specific optimization calculation process can be implemented by those skilled in the art based on common knowledge and design experience.

[0048] The rod drop accident safety analysis method provided in the above embodiments can be automatically implemented by a computing device. The computing device includes a memory and a processor, wherein a technical program is stored in the memory, and when the computing program is executed by the processor, the rod drop accident safety analysis method provided in any of the above embodiments can be implemented. Specifically, the computing device can be configured as a general-purpose computer, or a dedicated computing device, or a cloud computing device or a virtual machine. The computing program can be set as a script, macro or program module written in a general-purpose computing program, or it can be written as a dedicated computing program using a general-purpose computer language.

[0049] The purpose of the above embodiments is to further explain the present invention in detail in conjunction with the accompanying drawings so that those skilled in the art can understand the technical concept of the present invention. Within the scope of the present invention, the optimization or equivalent replacement of the method steps involved, and the combination of the implementation methods in different embodiments without conflict of principle, all fall within the protection scope of the present invention.

Claims

1. A rod drop accident safety analysis method, characterized in that: The following steps are involved: Step a): Provide a nuclear power plant accident analysis model, conduct a comprehensive reactor full cycle and full life cycle physical analysis for a given rod drop accident condition, and determine the range of the rod drop value; Step b): within the maximum drop value range, select drop value points at intervals of a given drop value step length to form a characteristic drop value sequence; Step c): Calculate the core axial power distribution and radial power peak factor F at each rod drop value point in the characteristic rod drop value sequence ΔH N ; Using a safety analysis program to conduct accident analysis on the rod drop value points in the characteristic rod drop value sequence, and obtain a rod drop condition analysis result sequence, wherein the rod drop condition analysis result sequence includes the thermal-hydraulic parameters corresponding to each rod drop value point in the characteristic rod drop value sequence, and the limit state point is calculated based on the thermal-hydraulic parameters, and the limit state point is the point where the ratio of the average heat flux density of the core to the core flow rate is the largest; Step d): According to the core axial power distribution and radial power peak factor F at the limit state point ΔH N Calculate the core deviation from nucleate boiling ratio.

2. The rod drop accident safety analysis method according to claim 1, characterized in that: When the analyzed reactor has a rod-lifting prohibition protection logic, the method further includes the step of optimizing the rod-lifting prohibition protection constant value: performing repeated sensitivity analysis calculations through the steps c) and d) to obtain F ΔH N The prohibition of lifting the rod protection setting value.

3. The rod drop accident safety analysis method according to claim 1 or 2, characterized in that: The step length of the rod drop value in step b) is 5 pcm-15 pcm.

4. The rod drop accident safety analysis method according to claim 1 or 2, characterized in that: The safety analysis program in step c) is configured as RELAP5.

5. The rod drop accident safety analysis method according to claim 1 or 2, characterized in that: In the step c), the accident analysis method includes calculating the rod drop accident conditions under different control rod differential values, moderator temperature coefficients and rod lifting values ​​to obtain thermal hydraulic parameters.

6. The rod drop accident safety analysis method according to claim 5, characterized in that: The thermal hydraulic parameters include the average core heat flux density, core flow rate, reactor power and coolant temperature.

7. The rod drop accident safety analysis method according to claim 1 or 2, characterized in that: In the step c), the process of calculating the limit state point also includes a reliability verification step, and the reliability verification step selects a plurality of the rod drop value points close to the limit state point for verification calculation.

8. A computing device, comprising a memory and a processor, characterized in that: The memory stores a calculation program, and when the calculation program is executed by the processor, the rod drop accident safety analysis method according to any one of claims 1 to 7 can be implemented.

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