A reactor shaft jam accident safety analysis method and device

Abstract

The invention discloses a reactor shaft-stuck accident safety analysis method and device. The method selects the minimum DNBR as the transient minimum DNBR. When the transient minimum DNBR is less than a preset DNBR threshold, the current thermal hydraulic parameters are obtained to adjust the initial enthalpy rise factor, and the DNBR calculation program is called to calculate the adjusted enthalpy rise factor; when the calculated DNBR is less than the preset DNBR threshold, the adjusted enthalpy rise factor is used as the new initial enthalpy rise factor, and the step of calling the DNBR calculation program is continuously executed until the calculated DNBR is not less than the preset DNBR threshold, and the corresponding enthalpy rise factor is used as the effective enthalpy rise factor; based on the effective enthalpy rise factor, the corresponding fuel rod share is selected from the fuel statistical curve as the fuel rod share that causes DNB in ​​the shaft-stuck accident, and the fuel rod share that causes DNB in ​​the transient state is automatically calculated without manual participation, thereby improving the efficiency of shaft-stuck accident analysis and reducing the error rate.

Description

Technical Field

[0001] The invention relates to the technical field of safety control of pressurized water reactors, and in particular to a safety analysis method and device for reactor shaft jamming accidents. Background Art

[0002] Reactor shaft jam accidents generally lead to DNB (Departure from Nucleate Boiling) of fuel rods. The proportion of fuel rods that experience DNB needs to be calculated during reactor safety analysis. The DNB fuel rod proportion is related to thermal-hydraulic parameters and the transient minimum DNBR (Deviation from Nucleate Boiling Ratio) value. In the prior art, the transient characteristics of the device under the accident are calculated according to a thermal-hydraulic transient analysis program (such as TRANTH), and the technicians input the calculated thermal-hydraulic data into the core thermal-hydraulic sub-channel analysis program (such as CORTH) to calculate all transient DNBR values. Then, the proportion of fuel rods that experience DNB is calculated under the fuel statistical curve line of different fuel cycles. This is time-consuming and labor-intensive, and is prone to human errors. Summary of the invention

[0003] The technical problem to be solved by the present invention is that in the process of calculating the proportion of fuel rods that have undergone DNB, data needs to be manually transmitted, which is time-consuming, labor-intensive and prone to human errors. Therefore, the present invention provides a method and device for safety analysis of reactor shaft jamming accidents. Since the parameters calculated by the thermal-hydraulic transient analysis program are relatively complicated, manual screening is prone to human errors and the workload is large. Therefore, the present application sets a core DNBR parameter extraction script to accurately extract the parameters required for calculating the DNBR at all transient moments, and sets a method for calculating the proportion of fuel rods that have undergone DNB to automatically calculate the proportion of fuel rods that have undergone DNB in ​​the transient state, thereby improving the efficiency of shaft jamming accident analysis and reducing the error rate.

[0004] The present invention is achieved through the following technical solutions:

[0005] A reactor shaft jam accident safety analysis method, comprising:

[0006] Obtain the DNBR at all transient calculation moments, and select the smallest DNBR as the transient minimum DNBR;

[0007] When the instantaneous minimum DNBR is less than the preset DNBR threshold, the thermal hydraulic parameters at the moment corresponding to the instantaneous minimum DNBR are read as the current thermal hydraulic parameters;

[0008] Obtaining an initial enthalpy rise factor and adjusting the initial enthalpy rise factor based on the current thermal hydraulic parameters, and calling a DNBR calculation program to calculate the adjusted enthalpy rise factor;

[0009] When the calculated DNBR is less than the preset DNBR threshold, the adjusted enthalpy rise factor is used as the new initial enthalpy rise factor, and the steps of calling the DNBR calculation program are continued until the calculated DNBR is not less than the preset DNBR threshold, and the corresponding enthalpy rise factor is used as the effective enthalpy rise factor;

[0010] Based on the effective enthalpy rise factor, a corresponding fuel rod share is selected from the fuel statistical curve as the fuel rod share that causes DNB in ​​the event of an axle sticking accident.

[0011] Further, the obtaining of the DNBR at all transient calculation moments includes:

[0012] Obtaining a transient thermal-hydraulic calculation file, and calling a core DNBR parameter extraction script to sequentially extract core DNBR calculation parameters at all transient calculation moments from the transient thermal-hydraulic calculation file;

[0013] If the reactor thermal power in the core DNBR calculation parameters is less than the preset rated power, the initial enthalpy rise factor is corrected, and the initial radial power distribution is updated based on the corrected enthalpy rise factor to obtain an updated radial power distribution;

[0014] Based on the core DNBR calculation parameters, the corrected enthalpy rise factor and the updated radial power distribution at each transient calculation moment, the core thermal-hydraulic subchannel analysis program is called to calculate the DNBR at all transient calculation moments.

[0015] Furthermore, the correction of the initial enthalpy rise factor includes:

[0016] The enthalpy rise factor correction formula is called to correct the initial enthalpy rise factor, and the enthalpy rise factor correction formula is specifically:

[0017] F_ΔH NEW =F_ΔH0×(1+C×(1-P))

[0018] Where F_ΔH NEW represents the enthalpy rise factor after correction of the initial enthalpy rise factor, F_ΔH0 represents the initial enthalpy rise factor, C represents the enthalpy rise factor correction coefficient, and P represents the reactor thermal power in the core DNBR calculation parameters.

[0019] Further, updating the initial radial power distribution based on the modified enthalpy rise factor to obtain an updated radial power distribution includes:

[0020] After obtaining the corrected enthalpy rise factor, the radial power distribution update calculation formula is called to calculate a new radial power distribution;

[0021] Based on the new radial power distribution, the initial radial power distribution in the core thermal hydraulic sub-channel analysis program is replaced and updated to obtain an updated radial power distribution;

[0022] The radial power distribution update calculation formula is specifically:

[0023]

[0024] In the formula, Q NEW represents the new radial power distribution after the initial radial power distribution is updated, Q0 represents the initial radial power distribution, F_ΔH NEW represents the enthalpy rise factor after the initial enthalpy rise factor is corrected, and F_ΔH0 represents the initial enthalpy rise factor.

[0025] Further, the adjusting the initial enthalpy rise factor based on the current thermal hydraulic parameters includes:

[0026] Based on the current thermal-hydraulic parameters, the modified enthalpy rise factor is adjusted by using a dichotomy method.

[0027] Further, the selecting a corresponding fuel rod share from the fuel statistical curve based on the effective enthalpy rise factor as the fuel rod share that causes DNB in ​​the event of an axle-stuck accident includes:

[0028] Normalizing the fuel statistical curves of different fuel cycles based on the initial enthalpy rise factor to obtain a normalized fuel statistical curve;

[0029] Based on the adjusted enthalpy rise factor, a corresponding fuel rod share is selected from the normalized fuel statistical curve as the fuel rod share that causes DNB in ​​the event of an axle sticking accident.

[0030] Furthermore, the fuel statistical curves of different fuel cycles are normalized based on the initial enthalpy rise factor to obtain a normalized fuel statistical curve, including:

[0031] Obtaining the enthalpy rise factor of each fuel rod share on the fuel statistical curve of all fuel cycles, and selecting the maximum enthalpy rise factor corresponding to each fuel rod share as the enthalpy rise factor of each fuel rod share on the enveloped fuel statistical curve;

[0032] constructing an enveloped fuel statistical curve based on the enthalpy rise factor of the enveloped fuel statistical curve and the corresponding fuel rod fraction;

[0033] Selecting the largest enthalpy rise factor from the fuel statistical curve of the envelope as the curve representative enthalpy rise factor, and calculating the ratio of the initial enthalpy rise factor and the curve representative enthalpy rise factor as the curve adjustment coefficient;

[0034] All enthalpy rise factors in the enveloped fuel statistical curve are normalized based on the curve adjustment coefficient to obtain a normalized fuel statistical curve.

[0035] Furthermore, the reactor shaft jam accident safety analysis method further includes:

[0036] Call the core thermal hydraulic sub-channel analysis program and set the saturated water meter parameter range according to the pressurizer pressure in the core DNBR calculation parameters;

[0037] The corresponding saturation temperature is determined according to the minimum saturated water meter parameter in the saturated water meter parameter range, and the subcooling water meter parameter range is set in combination with the core inlet temperature.

[0038] A reactor shaft jam accident safety analysis device, comprising:

[0039] The transient minimum DNBR acquisition module is used to obtain the DNBRs of all transient calculation moments and select the minimum DNBR as the transient minimum DNBR;

[0040] A current thermal hydraulic parameter determination module, configured to read the thermal hydraulic parameter at the moment corresponding to the transient minimum DNBR as the current thermal hydraulic parameter when the transient minimum DNBR is less than a preset DNBR threshold;

[0041] An initial enthalpy rise factor processing module is used to obtain an initial enthalpy rise factor and adjust the initial enthalpy rise factor based on the current thermal hydraulic parameters, and call a DNBR calculation program to calculate the adjusted enthalpy rise factor;

[0042] An effective enthalpy rise factor acquisition module is used to use the adjusted enthalpy rise factor as a new initial enthalpy rise factor when the calculated DNBR is less than a preset DNBR threshold, continue to execute the step of calling the DNBR calculation program until the calculated DNBR is not less than the preset DNBR threshold, and use the corresponding enthalpy rise factor as the effective enthalpy rise factor;

[0043] The fuel rod share acquisition module is used to select the corresponding fuel rod share from the fuel statistical curve based on the effective enthalpy rise factor as the fuel rod share that causes DNB in ​​the event of an axle sticking accident.

[0044] Furthermore, the fuel rod share acquisition module includes:

[0045] A fuel statistical curve processing unit, used for normalizing the fuel statistical curves of different fuel cycles based on the initial enthalpy rise factor to obtain a normalized fuel statistical curve;

[0046] The fuel rod share calculation unit is used to select the corresponding fuel rod share from the normalized fuel statistical curve based on the adjusted enthalpy rise factor as the fuel rod share that causes DNB in ​​the event of an axle sticking accident.

[0047] The invention provides a reactor shaft-stuck accident safety analysis method and device, which obtains DNBRs at all transient calculation moments and selects the minimum DNBR as the transient minimum DNBR; when the transient minimum DNBR is less than a preset DNBR threshold, the thermal-hydraulic parameters at the moment corresponding to the transient minimum DNBR are read as the current thermal-hydraulic parameters to adjust the initial enthalpy rise factor, and a DNBR calculation program is called to calculate the adjusted enthalpy rise factor; when the calculated DNBR is less than the preset DNBR threshold, the adjusted enthalpy rise factor is used as the new initial enthalpy rise factor, and the step of calling the DNBR calculation program is continued to be executed until the calculated DNBR is not less than the preset DNBR threshold, and the corresponding enthalpy rise factor is used as the effective enthalpy rise factor; based on the effective enthalpy rise factor, the corresponding fuel rod share is selected from the fuel statistical curve as the fuel rod share that causes DNB under the shaft-stuck accident, and the fuel rod share that causes DNB in ​​the transient state is automatically calculated without manual participation, thereby improving the efficiency of shaft-stuck accident analysis and reducing the error rate. BRIEF DESCRIPTION OF THE DRAWINGS

[0048] The drawings described herein are used to provide a further understanding of the embodiments of the present invention, constitute a part of this application, and do not constitute a limitation of the embodiments of the present invention. In the drawings:

[0049] Figure 1 The present invention is a flow chart of a reactor shaft jam accident safety analysis method.

[0050] Figure 2 Schematic diagram of the principle of calculating DNBR at all transient calculation moments in one embodiment of the present invention

[0051] Figure 3 For the present invention Figure 1 A specific flow chart of step S50 in FIG.

[0052] Figure 4 For the present invention Figure 3 A specific flow chart of step S51 in FIG.

[0053] Figure 5 The present invention is another flow chart of a reactor shaft jam accident safety analysis method.

[0054] Figure 6 The present invention is a schematic diagram of a reactor shaft jam accident safety analysis device. DETAILED DESCRIPTION

[0055] In order to make the objectives, technical solutions and advantages of the present invention more clearly understood, the present invention is further described in detail below in conjunction with embodiments and drawings. The exemplary embodiments of the present invention and their description are only used to explain the present invention and are not intended to limit the present invention.

[0056] Example 1

[0057] like Figure 1 As shown, the present invention provides a reactor shaft jam accident safety analysis method, comprising:

[0058] S10: Obtain the DNBRs at all transient calculation moments, and select the minimum DNBR as the transient minimum DNBR.

[0059] Specifically, Figure 2 As shown, obtaining the DNBR at all transient calculation moments includes the following steps:

[0060] (1) Obtain the transient thermal-hydraulic calculation file output by the thermal-hydraulic transient analysis program (such as TRANTH), and call the core DNBR parameter extraction script to extract the core DNBR calculation parameters of all transient calculation moments from the transient thermal-hydraulic calculation file in sequence; wherein, the core DNBR parameter extraction script refers to a script that can extract the core DNBR calculation parameters, and any script form that can extract DNBR parameters is acceptable, and this embodiment does not limit this. The core DNBR calculation parameters refer to the parameters needed to calculate the DNBR at the moment of transient calculation, including but not limited to the reactor thermal power P, the core inlet temperature t, the regulator pressure p and the loop flow. Among them, the loop flow refers to the sum of the volume flow shares of all loop coolants, that is, the total flow share obtained by adding the volume flow shares of the coolant of each loop.

[0061] Furthermore, in order to meet the needs of different users, when calculating the DNBR of all transient calculation moments, the present embodiment also sets a conservative processing process. The conservative processing process is specifically as follows: the user determines whether it is necessary to perform conservative processing on the pressurizer pressure p and the loop flow according to the actual situation. If conservative processing is required, the conservative pressurizer pressure value and the deviation of the loop flow volume share are input, and the DNBR of all transient calculation moments are calculated in combination with the reactor thermal power P, core inlet temperature t, pressurizer pressure p and loop flow in the core DNBR calculation parameters. If conservative processing is not required, the DNBR calculation of subsequent transient calculation moments is performed according to the reactor thermal power P, core inlet temperature t, pressurizer pressure p and loop flow in the core DNBR calculation parameters.

[0062] (II) If the reactor thermal power in the core DNBR calculation parameters is less than the preset rated power, the enthalpy rise factor correction formula is called to correct the initial enthalpy rise factor F_ΔH0, and based on the corrected enthalpy rise factor, the radial power distribution update calculation formula is called to calculate a new radial power distribution, and then the new radial power distribution is input into the core thermal hydraulic sub-channel analysis program (such as CORTH) to replace the original initial radial power distribution in the core thermal hydraulic sub-channel analysis program (such as CORTH) to obtain an updated radial power distribution. Among them, the initial enthalpy rise factor F_ΔH0 refers to the enthalpy rise factor set at the beginning of the system, and the initial radial power distribution refers to the radial power distribution set at the beginning of the system.

[0063] The enthalpy rise factor correction formula in this embodiment is specifically:

[0064] F_ΔH NEW =F_ΔH0×(1+C×(1-P))

[0065] Where F_ΔH NEW represents the enthalpy rise factor after correction of the initial enthalpy rise factor, F_ΔH0 represents the initial enthalpy rise factor, C represents the enthalpy rise factor correction coefficient, and P represents the reactor thermal power in the core DNBR calculation parameters.

[0066] The radial power distribution update calculation formula in this embodiment is specifically:

[0067]

[0068] In the formula, Q NEW represents the new radial power distribution after the initial radial power distribution is updated, Q0 represents the initial radial power distribution, F_ΔH NEW represents the enthalpy rise factor after the initial enthalpy rise factor is corrected, and F_ΔH0 represents the initial enthalpy rise factor

[0069] (III) Based on the core DNBR calculation parameters (reactor thermal power P, core inlet temperature t, pressurizer pressure p and loop flow) at each transient calculation moment, the corrected enthalpy rise factor and the updated radial power distribution, the core thermal-hydraulic sub-channel analysis program is called to calculate the DNBR at all transient calculation moments.

[0070] Furthermore, if the reactor thermal power in the core DNBR calculation parameters is not less than the preset rated power, the DNBR at all transient calculation moments is calculated based on the core DNBR calculation parameters, the initial enthalpy rise factor and the initial radial power distribution at each transient calculation moment.

[0071] S20: When the transient minimum DNBR is less than the preset DNBR threshold, the thermal-hydraulic parameters at the moment corresponding to the transient minimum DNBR are read as the current thermal-hydraulic parameters.

[0072] S30: obtaining an initial enthalpy rise factor and adjusting the initial enthalpy rise factor based on current thermal hydraulic parameters, and calling a DNBR calculation program to calculate the adjusted enthalpy rise factor.

[0073] Specifically, after obtaining the current thermal-hydraulic parameters, the initial enthalpy rise factor F_ΔH0 is adjusted using the dichotomy method according to the current thermal-hydraulic parameters.

[0074] S40: When the calculated DNBR is less than the preset DNBR threshold, the adjusted enthalpy rise factor is used as the new initial enthalpy rise factor, and the step of calling the DNBR calculation program is continued until the calculated DNBR is not less than the preset DNBR threshold, and the corresponding enthalpy rise factor is used as the effective enthalpy rise factor F_ΔH_f.

[0075] S50: Selecting a corresponding fuel rod share from the fuel statistical curve based on the effective enthalpy rise factor as the fuel rod share that causes DNB in ​​the event of an axle-stuck accident.

[0076] Furthermore, if Figure 3 As shown, step S50, based on the effective enthalpy rise factor, selects the corresponding fuel rod share from the fuel statistical curve as the fuel rod share that causes DNB in ​​the event of an axle jam, specifically including the following steps:

[0077] S51: normalizing the fuel statistical curves of different fuel cycles based on the initial enthalpy rise factor to obtain a normalized fuel statistical curve.

[0078] S52: Based on the adjusted enthalpy rise factor, a corresponding fuel rod share is selected from the normalized fuel statistical curve as the fuel rod share that causes DNB in ​​the event of an axle sticking accident.

[0079] Furthermore, if Figure 4 As shown, in step S51, the fuel statistical curves of different fuel cycles are normalized based on the initial enthalpy rise factor, which specifically includes the following steps:

[0080] S511: Based on the fuel statistical curves of different fuel cycles, an envelope fuel statistical curve is obtained.

[0081] Specifically, since the fuel statistical curves of different fuel cycles are not the same, in order to consider all situations, this embodiment integrates the uncertainties of all fuel cycles, first obtains the enthalpy rise factor of each fuel rod share on the fuel statistical curves of all fuel cycles, and selects the maximum enthalpy rise factor corresponding to each fuel rod share as the enthalpy rise factor of each fuel rod share on the enveloped fuel statistical curve, and then constructs the enveloped fuel statistical curve using the enthalpy rise factor of the enveloped fuel statistical curve as the ordinate and the corresponding fuel rod share as the abscissa, so as to obtain the material statistical curve that envelopes all fuel cycles, so that subsequent calculation results are more conservative.

[0082] S512: Select the largest enthalpy rise factor from the enveloped fuel statistical curve as the curve representative enthalpy rise factor, and calculate the ratio of the initial enthalpy rise factor and the curve representative enthalpy rise factor as the curve adjustment coefficient.

[0083] S513: Normalizing all enthalpy rise factors in the enveloped fuel statistical curve based on the curve adjustment coefficient to obtain a normalized fuel statistical curve.

[0084] Specifically, the enthalpy rise factor in the enveloped fuel statistical curve is generally lower than the initial enthalpy rise factor. Therefore, after obtaining the curve adjustment coefficient, all enthalpy rise factors in the enveloped fuel statistical curve are multiplied by the curve adjustment coefficient to obtain a more conservative fuel statistical curve and improve the accuracy of DNBR calculation.

[0085] Furthermore, in order to improve the accuracy of DNBR calculation at the transient calculation time, the saturated water meter and the subcooled water meter need to be adaptively adjusted according to the actual situation to suit the current reaction state of the reactor. Figure 5 As shown, the reactor shaft jam accident safety analysis method also includes:

[0086] S61: Call the core thermal hydraulic sub-channel analysis program and set the saturated water meter parameter range according to the stabilizer pressure in the core DNBR calculation parameters.

[0087] S62: Determine the corresponding saturation temperature according to the minimum saturated water meter parameter in the saturated water meter parameter range, and set the subcooling water meter parameter range in combination with the core inlet temperature.

[0088] Specifically, the saturated water meter parameter range is set according to the pressurizer pressure p in the core DNBR calculation parameters, and then the corresponding saturation temperature is determined according to the minimum saturated water meter parameter in the saturated water meter parameter range. Combined with the core inlet temperature, the core thermal-hydraulic sub-channel analysis program (such as CORTH) is called to generate the subcooling water meter parameter range.

[0089] For ease of understanding, this embodiment is illustrated by the following example: the saturated water meter parameter range of p-5bar to p+5bar is set according to the stabilizer pressure p in the core DNBR calculation parameters, and then the corresponding saturated temperature t2 is determined according to the smallest saturated water meter parameter p-5bar in the saturated water meter parameter range, and combined with the core inlet temperature t1, the core thermal hydraulic sub-channel analysis program (such as CORTH) is called to generate a subcooled water meter parameter range in the temperature range of t1 to t2-0.1℃.

[0090] The invention provides a reactor shaft jam accident safety analysis method. The method comprises the following steps: setting a core DNBR parameter extraction script to accurately extract parameters required for calculating DNBR at all transient moments, and setting a method for calculating the fuel rod share that has DNB: reading the thermal hydraulic parameters at the moment corresponding to the transient minimum DNBR as the current thermal hydraulic parameters to adjust the initial enthalpy rise factor, then calling a DNBR calculation program to calculate the adjusted enthalpy rise factor, and when the calculated DNBR is less than a preset DNBR threshold, using the adjusted enthalpy rise factor as a new initial enthalpy rise factor, continuing to execute the step of calling the DNBR calculation program until the calculated DNBR is not less than the preset DNBR threshold, and stopping, and using the corresponding enthalpy rise factor as an effective enthalpy rise factor; finally, selecting the corresponding fuel rod share from a fuel statistical curve based on the effective enthalpy rise factor as the fuel rod share that has DNB under a shaft jam accident, and automatically calculating the fuel rod share that has DNB in ​​the transient state, and automatically calculating the fuel rod share that has DNB in ​​the transient state without manual participation, thereby improving the efficiency of shaft jam accident analysis and reducing the error rate.

[0091] Example 2

[0092] like Figure 6 As shown, the present invention provides a reactor shaft sticking accident safety analysis device which corresponds one-to-one to a reactor shaft sticking accident safety analysis method in Example 1, comprising a transient minimum DNBR acquisition module 10, a current thermal-hydraulic parameter determination module 20, an initial enthalpy rise factor processing module 30, an effective enthalpy rise factor acquisition module 40 and a fuel rod share acquisition module 50.

[0093] The transient minimum DNBR acquisition module 10 is used to acquire the DNBRs at all transient calculation moments and select the minimum DNBR as the transient minimum DNBR.

[0094] The current thermal-hydraulic parameter determination module 20 is used to read the thermal-hydraulic parameter at the moment corresponding to the transient minimum DNBR as the current thermal-hydraulic parameter when the transient minimum DNBR is less than a preset DNBR threshold.

[0095] The initial enthalpy rise factor processing module 30 is used to obtain the initial enthalpy rise factor and adjust the initial enthalpy rise factor based on the current thermal hydraulic parameters, and call the DNBR calculation program to calculate the adjusted enthalpy rise factor.

[0096] The effective enthalpy rise factor acquisition module 40 is used to use the adjusted enthalpy rise factor as the new initial enthalpy rise factor when the calculated DNBR is less than the preset DNBR threshold, continue to execute the step of calling the DNBR calculation program, and stop when the calculated DNBR is not less than the preset DNBR threshold, and use the corresponding enthalpy rise factor as the effective enthalpy rise factor.

[0097] The fuel rod share acquisition module 50 is used to select the corresponding fuel rod share from the fuel statistical curve based on the effective enthalpy rise factor as the fuel rod share that causes DNB in ​​the event of a shaft jam.

[0098] Furthermore, the fuel rod share acquisition module 50 includes a fuel statistical curve processing unit and a fuel rod share calculation unit.

[0099] The fuel statistical curve processing unit is used to perform normalization processing on the fuel statistical curves of different fuel cycles based on the initial enthalpy rise factor to obtain a normalized fuel statistical curve.

[0100] The fuel rod share calculation unit is used to select the corresponding fuel rod share from the normalized fuel statistical curve based on the adjusted enthalpy rise factor as the fuel rod share that causes DNB in ​​the event of an axle jam.

[0101] The above specific implementation methods further illustrate the objectives, technical solutions and beneficial effects of the present invention in detail. It should be understood that the above are only specific implementation methods of the present invention and are not intended to limit the scope of protection of the present invention. Any modifications, equivalent substitutions, improvements, etc. made within the spirit and principles of the present invention should be included in the scope of protection of the present invention.

Claims

1. A reactor shaft jam accident safety analysis method, characterized in that: include: Obtain the DNBR at all transient calculation moments, and select the smallest DNBR as the transient minimum DNBR; When the instantaneous minimum DNBR is less than the preset DNBR threshold, the thermal hydraulic parameters at the moment corresponding to the instantaneous minimum DNBR are read as the current thermal hydraulic parameters; Obtaining an initial enthalpy rise factor and adjusting the initial enthalpy rise factor based on the current thermal hydraulic parameters, and calling a DNBR calculation program to calculate the adjusted enthalpy rise factor; When the calculated DNBR is less than the preset DNBR threshold, the adjusted enthalpy rise factor is used as the new initial enthalpy rise factor, and the steps of calling the DNBR calculation program are continued until the calculated DNBR is not less than the preset DNBR threshold, and the corresponding enthalpy rise factor is used as the effective enthalpy rise factor; Based on the effective enthalpy rise factor, a corresponding fuel rod share is selected from the fuel statistical curve as the fuel rod share that causes DNB in ​​the event of an axle sticking accident.

2. A reactor shaft jam accident safety analysis method according to claim 1, characterized in that: The step of obtaining the DNBR at all transient calculation moments includes: Obtaining a transient thermal-hydraulic calculation file, and calling a core DNBR parameter extraction script to sequentially extract core DNBR calculation parameters at all transient calculation moments from the transient thermal-hydraulic calculation file; If the reactor thermal power in the core DNBR calculation parameters is less than the preset rated power, the initial enthalpy rise factor is corrected, and the initial radial power distribution is updated based on the corrected enthalpy rise factor to obtain an updated radial power distribution; Based on the core DNBR calculation parameters, the corrected enthalpy rise factor and the updated radial power distribution at each transient calculation moment, the core thermal-hydraulic subchannel analysis program is called to calculate the DNBR at all transient calculation moments.

3. A reactor shaft jam accident safety analysis method according to claim 2, characterized in that: The correction of the initial enthalpy rise factor comprises: The enthalpy rise factor correction formula is called to correct the initial enthalpy rise factor, and the enthalpy rise factor correction formula is specifically: F_ΔH NEW =F_ΔH0×(1+C×(1-P)) Where F_ΔH NEW represents the enthalpy rise factor after correction of the initial enthalpy rise factor, F_ΔH0 represents the initial enthalpy rise factor, C represents the enthalpy rise factor correction coefficient, and P represents the reactor thermal power in the core DNBR calculation parameters.

4. A reactor shaft jam accident safety analysis method according to claim 3, characterized in that: The updating of the initial radial power distribution based on the modified enthalpy rise factor to obtain an updated radial power distribution includes: After obtaining the corrected enthalpy rise factor, the radial power distribution update calculation formula is called to calculate a new radial power distribution; Based on the new radial power distribution, the initial radial power distribution in the core thermal hydraulic sub-channel analysis program is replaced and updated to obtain an updated radial power distribution; The radial power distribution update calculation formula is specifically: In the formula, Q NEW represents the new radial power distribution after the initial radial power distribution is updated, Q0 represents the initial radial power distribution, F_ΔH NEW represents the enthalpy rise factor after the initial enthalpy rise factor is corrected, and F_ΔH0 represents the initial enthalpy rise factor.

5. A reactor shaft jam accident safety analysis method according to claim 1, characterized in that: The adjusting the initial enthalpy rise factor based on the current thermal hydraulic parameters includes: Based on the current thermal-hydraulic parameters, the initial enthalpy rise factor is adjusted using a dichotomy method.

6. A reactor shaft jam accident safety analysis method according to claim 1, characterized in that: The selecting a corresponding fuel rod share from the fuel statistical curve based on the effective enthalpy rise factor as the fuel rod share that causes DNB in ​​the event of an axle jamming accident includes: Normalizing the fuel statistical curves of different fuel cycles based on the initial enthalpy rise factor to obtain a normalized fuel statistical curve; Based on the adjusted enthalpy rise factor, a corresponding fuel rod share is selected from the normalized fuel statistical curve as the fuel rod share that causes DNB in ​​the event of an axle sticking accident.

7. A reactor shaft jam accident safety analysis method according to claim 1, characterized in that: The fuel statistical curves of different fuel cycles are normalized based on the initial enthalpy rise factor to obtain a normalized fuel statistical curve, including: Obtaining the enthalpy rise factor of each fuel rod share on the fuel statistical curve of all fuel cycles, and selecting the maximum enthalpy rise factor corresponding to each fuel rod share as the enthalpy rise factor of each fuel rod share on the enveloped fuel statistical curve; constructing an enveloped fuel statistical curve based on the enthalpy rise factor of the enveloped fuel statistical curve and the corresponding fuel rod fraction; Selecting the largest enthalpy rise factor from the fuel statistical curve of the envelope as the curve representative enthalpy rise factor, and calculating the ratio of the initial enthalpy rise factor and the curve representative enthalpy rise factor as the curve adjustment coefficient; All enthalpy rise factors in the enveloped fuel statistical curve are normalized based on the curve adjustment coefficient to obtain a normalized fuel statistical curve.

8. A reactor shaft jam accident safety analysis method according to claim 1, characterized in that: The reactor shaft jam accident safety analysis method further comprises: Call the core thermal hydraulic sub-channel analysis program and set the saturated water meter parameter range according to the pressurizer pressure in the core DNBR calculation parameters; The corresponding saturation temperature is determined according to the minimum saturated water meter parameter in the saturated water meter parameter range, and the subcooling water meter parameter range is set in combination with the core inlet temperature.

9. A reactor shaft jam accident safety analysis device, characterized in that: include: The transient minimum DNBR acquisition module is used to obtain the DNBRs of all transient calculation moments and select the minimum DNBR as the transient minimum DNBR; A current thermal hydraulic parameter determination module, configured to read the thermal hydraulic parameter at the moment corresponding to the transient minimum DNBR as the current thermal hydraulic parameter when the transient minimum DNBR is less than a preset DNBR threshold; An initial enthalpy rise factor processing module is used to obtain an initial enthalpy rise factor and adjust the initial enthalpy rise factor based on the current thermal hydraulic parameters, and call a DNBR calculation program to calculate the adjusted enthalpy rise factor; An effective enthalpy rise factor acquisition module is used to use the adjusted enthalpy rise factor as a new initial enthalpy rise factor when the calculated DNBR is less than a preset DNBR threshold, continue to execute the step of calling the DNBR calculation program until the calculated DNBR is not less than the preset DNBR threshold, and use the corresponding enthalpy rise factor as the effective enthalpy rise factor; The fuel rod share acquisition module is used to select the corresponding fuel rod share from the fuel statistical curve based on the effective enthalpy rise factor as the fuel rod share that causes DNB in ​​the event of an axle sticking accident.

10. A reactor shaft jam accident safety analysis device according to claim 9, characterized in that: The fuel rod share acquisition module includes: A fuel statistical curve processing unit, used for normalizing the fuel statistical curves of different fuel cycles based on the initial enthalpy rise factor to obtain a normalized fuel statistical curve; The fuel rod share calculation unit is used to select the corresponding fuel rod share from the normalized fuel statistical curve based on the adjusted enthalpy rise factor as the fuel rod share that causes DNB in ​​the event of an axle sticking accident.

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