CN114420315A

CN114420315A - Method for obtaining safety margin

Info

Publication number: CN114420315A
Application number: CN202111369696.1A
Authority: CN
Inventors: 周金满; 崔怀明; 陈长; 陈亮; 郭锐; 王明利; 王亚曦
Original assignee: Nuclear Power Institute of China
Current assignee: Nuclear Power Institute of China
Priority date: 2021-11-18
Filing date: 2021-11-18
Publication date: 2022-04-29

Abstract

In order to solve the technical problem that the safety margin is obtained by sacrificing the economy of a nuclear power plant in the prior art, the embodiment of the invention provides a method for obtaining the safety margin. The method comprises the following steps: obtaining a safety margin by adjusting the configuration of a pressurized water reactor control rod group; by adjusting the configuration of the pressurized water reactor control rod set to obtain the safety margin, the defects that the safety margin is obtained by increasing the flow rate of the reactor coolant, the safety margin is obtained by reducing the average temperature of the reactor coolant, the capacity configuration of the safety system is increased and the like in the prior art, which sacrifice the economy of a nuclear power plant, are overcome.

Description

Method for obtaining safety margin

Technical Field

The invention relates to a method for acquiring a safety margin.

Background

Hualongyi is a three-generation pressurized water reactor nuclear power unit independently developed in China and bears the strategic importance of nuclear power development in China. In order to increase the competitiveness of Hualong I and further improve the economy of Hualong I, the power of a reactor core is increased from 3050MWth to 3180MWth, the cycle length of a balanced cycle is increased from 475EFPD to 495EFPD, and the cycle length of a first cycle is increased by 106 EFPD; in order to reduce the high-radioactivity solid waste and the purchasing cost of boron glass burnable poison, gadolinium-loaded rods are used for replacing boron glass burnable poison.

At present, the pressurized water reactor at home and abroad generally adopts the following methods to obtain the safety margin: the method comprises the following steps: the safety margin is obtained by increasing the flow of the reactor coolant, and the method puts higher requirements on a reactor body system and a reactor primary circuit system and increases the manufacturing cost of equipment; the second method comprises the following steps: the safety margin is obtained by reducing the average temperature of the reactor coolant, and the method causes the efficiency of the two-loop steam turbine to be reduced and greatly sacrifices the output efficiency of the steam turbine; the third method comprises the following steps: increasing the capacity configuration of the security system increases the cost of configuring the security system. In summary, the above methods all sacrifice the economics of a nuclear power plant in order to obtain a safety margin.

Disclosure of Invention

In order to solve the technical problem that the safety margin is obtained by sacrificing the economy of a nuclear power plant in the prior art, the embodiment of the invention provides a method for obtaining the safety margin.

The embodiment of the invention is realized by the following technical scheme:

in a first aspect, an embodiment of the present invention provides a method for obtaining a safety margin, including:

the safety margin is obtained by adjusting the configuration of the pressurized water reactor control rod set.

Further, the obtaining of the safety margin by adjusting the configuration of the pressurized water reactor control rod group comprises:

the configuration of the power compensation bar set is adjusted to obtain a safety margin.

the configuration of the set of temperature adjustment bars is adjusted to obtain a safety margin.

the trip bar set configuration is adjusted to obtain a safety margin.

Further, the adjusting the configuration of the power compensation rod set to obtain a safety margin includes:

and adjusting the configuration of the power compensation rod set to obtain a safety margin by taking the axial power deviation delta I of the reactor core and the axial power peak factor Fz of the reactor core as indexes.

Further, the adjusted power compensation bar set satisfied load following simulations from the beginning of life to 85% EOL.

Further, the number of absorber rods in the gray rod bundles of the adjusted power compensation rod group is 8-12; the overlapping steps of the power compensation rod groups are 100-90-90.

Further, the configuration of the temperature adjusting rod set is adjusted to obtain a safety margin; the method comprises the following steps:

the insertion limit of the adjusted temperature regulation bar set at full power level was 46 insertion steps, and 0% FP was 28 insertion steps.

Furthermore, the power compensation bar group after the adjustment and the configuration and the temperature adjustment bar group after the adjustment and the configuration both meet the LOCA limit line requirement.

Further, the configuration of the shutdown rod group is adjusted to obtain a safety margin; the method comprises the following steps:

the trip margin limit and/or the ARI-SABCD parameter are used to adjust the trip bar set configuration to obtain a safety margin.

Compared with the prior art, the embodiment of the invention has the following advantages and beneficial effects:

according to the method for obtaining the safety margin, the safety margin is obtained by adjusting the configuration of the pressurized water reactor control rod set, and the defects that in the prior art, the safety margin is obtained by increasing the flow of the reactor coolant, the safety margin is obtained by reducing the average temperature of the reactor coolant, the safety margin is obtained by increasing the capacity configuration of the safety system and the like, and the safety margin is obtained by sacrificing the economy of a nuclear power plant are overcome.

Drawings

In order to more clearly illustrate the technical solutions of the exemplary embodiments of the present invention, the drawings that are required to be used in the embodiments will be briefly described below, it should be understood that the following drawings only illustrate some embodiments of the present invention and therefore should not be considered as limiting the scope, and that for those skilled in the art, other related drawings can be obtained from these drawings without inventive effort.

Fig. 1 is a graph of Δ I as a function of power level.

Fig. 2 is a graph of Fz as a function of power level.

FIG. 3 is a graph of F Δ H versus burnup for different rod inserts.

Fig. 4a is a graph of Δ I over time for an 85% EOL load following procedure.

Fig. 4b is a graph of 85% EOL load tracking process Fz versus time.

FIG. 5 is a graph of Δ I as a function of R bar set position.

FIG. 6 is a LOCA limit line verification diagram.

Reference numbers and corresponding part names in the drawings:

1-100-90-90 overlapping steps; 2-90-90-90 overlapping steps; 3-90-90-90 overlapping steps; 4-100-90-90 overlapping steps; 5-8 absorber rod schemes; 6-12 absorber rod versions; 7-LOCA limit.

Detailed Description

In order to make the objects, technical solutions and advantages of the present invention more apparent, the present invention is further described in detail below with reference to examples and accompanying drawings, and the exemplary embodiments and descriptions thereof are only used for explaining the present invention and are not meant to limit the present invention.

In the following description, numerous specific details are set forth in order to provide a thorough understanding of the present invention. However, it will be apparent to one of ordinary skill in the art that: it is not necessary to employ these specific details to practice the present invention. In other instances, well-known structures, circuits, materials, or methods have not been described in detail so as not to obscure the present invention.

Throughout the specification, reference to "one embodiment," "an embodiment," "one example," or "an example" means: the particular features, structures, or characteristics described in connection with the embodiment or example are included in at least one embodiment of the invention. Thus, the appearances of the phrases "one embodiment," "an embodiment," "one example" or "an example" in various places throughout this specification are not necessarily all referring to the same embodiment or example. Furthermore, the particular features, structures, or characteristics may be combined in any suitable combination and/or sub-combination in one or more embodiments or examples. Further, those of ordinary skill in the art will appreciate that the illustrations provided herein are for illustrative purposes and are not necessarily drawn to scale. As used herein, the term "and/or" includes any and all combinations of one or more of the associated listed items.

In the description of the present invention, the terms "front", "rear", "left", "right", "upper", "lower", "vertical", "horizontal", "upper", "lower", "inner", "outer", etc. indicate orientations or positional relationships based on those shown in the drawings, and are only for convenience of description and simplicity of description, but do not indicate or imply that the device or element being referred to must have a particular orientation, be constructed in a particular orientation, and be operated, and therefore, should not be construed as limiting the scope of the present invention.

Examples

In the prior art, in order to obtain the safety margin, the economy of the nuclear power plant needs to be sacrificed, and in order to solve the technical problem that the safety margin is obtained by sacrificing the economy of the nuclear power plant in the prior art, the embodiment of the invention provides a method for obtaining the safety margin. The method comprises the following steps: the safety margin is obtained by adjusting the configuration of the pressurized water reactor control rod set.

Therefore, the embodiment of the invention overcomes the defects that the safety margin is obtained by sacrificing the economy of the nuclear power plant by adjusting the configuration of the pressurized water reactor control rod set to obtain the safety margin, increasing the flow rate of the reactor coolant to obtain the safety margin, reducing the average temperature of the reactor coolant to obtain the safety margin, increasing the capacity configuration of the safety system and the like in the prior art.

Alternatively, this is achieved by a review of several of the following aspects: a power compensation rod set configuration, a temperature adjustment rod set configuration, and a shutdown rod set configuration. The power compensation bar set configuration study includes: the number and arrangement of absorber rods in the gray rod bundle; and step two, overlapping the power compensation rod sets. The temperature adjustment bar group (R bar group) configuration investigation includes: the R rod group has the capability of controlling the axial power deviation (delta I) of a reactor core; insertion limit of the R bar set as a function of power level. And verifying the LOCA limit line. The shutdown rod group configuration study comprises the following steps: firstly, a shutdown allowance is obtained; and secondly, pile stopping bar configuration under cold and hot pile stopping working conditions.

Optionally, the obtaining the safety margin by adjusting the configuration of the pressurized water reactor control rod group comprises:

Further, the adjusting the configuration of the power compensation rod set to obtain a safety margin includes: and adjusting the configuration of the power compensation rod set to obtain a safety margin by taking the axial power deviation delta I of the reactor core and the axial power peak factor Fz of the reactor core as indexes.

The Hualong I operation Mode G mainly compensates reactivity caused by power change through gray rod groups (G1, G2 rod groups) and partial black rod groups (N1, N2), and core power is maintained at a required power level through the power compensation rod groups, but the core disturbance can be caused by the insertion of the power compensation rod groups, and the core disturbance has certain influence on the core safety. The influence of the power compensation rod set on the core safety is mainly measured by indexes such as core axial power deviation (delta I) and core axial power peak factor (Fz).

According to the design experience and operation feedback of a pressurized water reactor nuclear power unit in China, the material of the absorber rod in the Hualongyi gray rod bundle is silver-indium-cadmium, the other material is stainless steel, and the black rod bundle adopts 24 silver-indium-cadmium rods. Since the 24 guide tube positions available for the insertion of control rods in the fuel assembly are fixed and the black bundle employs 24 ag-in-cd rods, only the number and arrangement of absorber rods in the gray bundle is varied. The number and arrangement of absorber rods in a gray rod cluster may have some effect on its controllability, and thus, the number and arrangement of absorber rods in a gray rod cluster need to be studied.

Since the contribution of the gray bundles to the trip margin is small relative to the set of trip bars, the contribution of the gray bundles to the trip margin is not considered here. The gray rod cluster is designed based primarily on its control capability and impact on core safety. First, the number arrangement of absorber rods (silver-indium-cadmium rods) in the gray rod group was investigated. And (3) performing operation simulation calculation at different power levels at the beginning and the end of the service life by adopting one-dimensional reactor core calculation software, thereby obtaining: the number of different absorber rods in a gray rod cluster, Δ I, as a function of power level at the beginning and end of life, is shown with reference to fig. 1; the number of different absorber rods in a gray rod cluster Fz as a function of power level at the beginning and end of life is shown with reference to FIG. 2.

The following results can be obtained from fig. 1 and 2: the increase of absorber rods in the ash rod bundle enables the reactor core delta I to be changed steeply along with the horizontal change, and the reactor core delta I can put higher requirements on delta I control capability in the load tracking process, and especially increase the risk of delta I out of control in the load tracking process at the end of the cycle life; secondly, under the low-power operation, the control rods are inserted deeply, and more absorber rods in the gray rod cluster (namely the gray rods have higher value) cause the delta I of the reactor core to shift to the negative direction to a greater extent, so that the axial power distribution factor Fz is increased. Thus excluding the 16 absorber rod solution and the 20 absorber rod solution. Under the condition that the axial power distribution peak factor Fz is equivalent in size (namely, the 4-absorber-rod scheme is equivalent to the 8-absorber-rod scheme and the 12-absorber-rod scheme in terms of Fz control), the scheme is selected from the aspect of control capacity, and the scheme of selecting more absorber rods in the gray rod bundle is favorable for improving the operation control capacity of the nuclear power plant. Thus, with this study, the 8 and 12 absorber rod versions in gray rod bundles were now selected as the subsequent study.

The geometrical arrangement of the gray rod cluster absorber rods was studied next. Since the geometrical arrangement of the absorber rods is referred to as the arrangement in the radial direction, it is only necessary to investigate its effect on the nuclear enthalpy rise factor (F Δ H). The effect of all geometrical arrangements of the gray rod bundle absorber rods (7 schemes in total) on F Δ H was now studied with 12 silver-indium-cadmium absorber rods in the gray rod bundle as a reference. The change of F delta H along with the fuel consumption under the condition that the G1 rod group and the G1G2 rod group are inserted into the bottom of the reactor core is calculated by adopting three-dimensional reactor core calculation software, and the change is shown by referring to FIG. 3. As can be seen from fig. 3: the geometrical arrangement of the gray rod cluster absorber rods has little, if any, effect on F Δ H.

Therefore, the study in this section reveals that: the subsequent study was continued with the selection of 8 and 12 absorber rod solutions in gray rod bundles, the geometrical arrangement of which could be selected as an option. The scheme which has strong core power control capability and better control axial power peak factor (Fz) is selected from the angles of the number and the geometric arrangement of the absorber rods in the gray rod bundle, and a good foundation is laid for obtaining a large safety margin for the core.

Further, the adjusted power compensation rod group meets load tracking simulation from the beginning of the service life to 85% of EOL, and the number of absorber rods in gray rod bundles of the adjusted power compensation rod group is 8-12; the overlapping steps of the power compensation rod groups are 100-90-90.

In order to effectively control the reactivity and the axial power distribution of the reactor core, the influence of the movement of the power compensation rod groups on the axial power distribution of the reactor core at various power levels is as small as possible, so that the overlapping steps among the control rod groups need to be optimally set.

Two schemes of 8 absorber rods and 12 absorber rods in the gray rod cluster are taken as research standards, and four overlapping step setting schemes of 70-70-70 (namely, the overlapping step of G1 and G2 rod groups is 70, the overlapping step of G2 and N1 rod groups is 70, the overlapping step of N1 and N2 rod groups is 70), 80-80-80, 90-90-90 and 100-one-other-step setting schemes are respectively researched. The SMART program is adopted to carry out load tracking simulation calculation from the beginning of the life to 85% EOL of each overlapping step scheme, and the calculation results show that: the overlap step settings are 70-70-70, 80-80-80, 100-. Therefore, the scheme in which only the power compensation rod group overlap step is set to 90-90-90 in the study reference of 12 absorber rods in the gray rod cluster is an effective scheme.

Due to the low neutron flux in the top and bottom cores and the low value of the gray rod groups G1 and G2, in order to avoid large disturbances in reactivity and axial power shift when control rods are inserted into the core continuously in a certain order, the overlapping step between the gray rod groups G1 and G2 control rod groups needs to be further studied on the scheme that the overlapping step of the power compensation rod groups is set to 90-90-90. Core disturbances near end-of-life load tracking are greater relative to other core burnouts because: when the power is increased or decreased, the temperature change at the inlet of the reactor core is small, and the temperature change at the outlet of the reactor core is large, so that the reactivity change at the upper part of the reactor core is larger than that at the lower part of the reactor core. The power reduction results in a reduction in the core moderator temperature, and the upper moderator temperature changes more than the lower core, so the upper core releases more reactivity, the power contribution in the upper core rises, and Δ I becomes positive, and when the core power rises, the Δ I changes in the opposite direction. ② along with the increase of the reactor core burnup, the decrease of the concentration of the soluble boron leads to the increase of the feedback of the moderator, and the change of the delta I caused by the temperature of the moderator is inevitably increased when the power is increased and decreased. Therefore, only the overlap step between gray rod groups G1 and G2 control rod groups at 85% EOL needs to be studied.

Load following simulation calculations were performed at 85% EOL using one-dimensional core calculation software, the only difference between the two calculation inputs being the overlap between G1 and G2 at

steps

100 and 90, respectively, and the results of the calculations are shown with reference to fig. 4a and 4 b. Referring to FIG. 4a, steps 1 overlap 100-90-90 and steps 2 overlap 90-90-90; reference is made to fig. 4b for steps 3 overlapping 90-90-90 and steps 4 overlapping 100-90-90.

From fig. 4a and 4b it can be seen that: the overlap between the G1 and G2 bar groups was set to 100 steps which gave better results than 90 steps, with a smaller range of Δ I fluctuation and a smaller Fz. Therefore, the overlap step was set to 100-90-90 in the study standard for 12 absorber rods in the gray rod cluster. In the same way, the overlapping steps in the study benchmarks were also set to 100-90-90 for 8 absorber rods in the gray rod cluster.

Through the overlapped step setting research of the power compensation rod groups, the reactor core power distribution distortion in the load tracking process is effectively relieved, and the total requirement that the load is tracked to 85% EOL can be met. Because the core power distribution of load tracking is also the basis of safety analysis, the core power distribution distortion is effectively relieved, and the allowance of safety analysis is facilitated.

Hualong I adopts a MODE G operation MODE, and a temperature adjusting rod (R rod group) of the model G is used for adjusting the average temperature of a reactor core, compensating the slight change of the reactivity and controlling the axial power deviation (delta I). When R-bar sets are used to control axial offset (Δ I), it is necessary to ensure that the R-bar set movement direction is a monotonic function of Δ I, namely: the R-rod set lift (or plunge) biases the core power toward the top of the core (or the bottom of the core). The relation between the moving direction of the R rod group and the delta I under different power levels is calculated by adopting three-dimensional core calculation software, and the relation is shown in a monotone function relation by referring to the graph of FIG. 5. Thus, the R-rod set provides the ability to control the axial offset of the core, which ensures that the core is within the operating map during normal operating conditions.

the trip bar set configuration is adjusted to obtain a safety margin.

In Hualong I model G operation Mode, the R rod group is provided with a bite amount position and an insertion limit position. The R-rod set has a differential value of 2.5 pcm/step in order to ensure that the temperature conditioning rod set R has sufficient reactivity introduction capability to meet the requirement of compensating for core reactivity disturbances. The maximum insertion depth of the rod set R is also limited during core operation to meet the following requirements:

a shutdown allowance requirement;

rod ejection accident safety guidelines;

nuclear enthalpy increasing factor F delta H <1.63[1+0.3(1-Pr) ] (Pr ≦ 0 ≦ 1)

FΔH≤1.63 (Pr＞1)

And calculating the bite amount position of the R rod group by adopting three-dimensional core calculation software, thereby preliminarily determining that the insertion limit range of the R rod group is 46-51 insertion steps under the full power (HFP). The analysis shows that: the requirements for shutdown allowance and nuclear enthalpy rise factor can be met within the range of 46-51 insertion steps. According to the past design experience, the safety margin of the bullet rod accident analysis is quite short, and because the power of Hualong I is increased, the cycle length is increased, and gadolinium contained in the first cycle can sacrifice a certain safety margin, the subsequent bullet rod accident analysis can not meet the requirement of the criterion, so that the safety margin is left for the bullet rod accident analysis as far as possible when the insertion limit is set.

And (3) calculating key safety parameters required by the analysis of the rod ejection accident by adopting three-dimensional reactor core calculation software with the insertion limit of 46 steps and 51 steps respectively, and obtaining the following calculation results: compared with the insertion step 51, the insertion limit 46 is inserted into the step, the value of the elastic rod at the full power level at the end of the first cycle life is reduced by about 20%, and the FQ is reduced by about 10%, so that a larger safety margin is obtained for analyzing the accident of the elastic rod. Thus the insertion limit for the Hualong I R-bar group at full power level was set to 46 insertion steps and 0% FP was set to 28 insertion steps.

The core local power density is related to the total power peak factor FQT at maximum operating power. During normal operation, FQT must be maintained within the LOCA envelope limits.

The envelope of the linear power density as a function of core height for verifying the LOCA limit and the fuel melting limit is determined by:

Q^T(z)＝max[P(x,y,z)]×F_I

in the formula:

p (x, y, z) is the three-dimensional linear power density of the reactor core under the I-type working condition obtained by program calculation;

F_Iuncertainty considered in calculating the final envelope;

the analysis of normal operating conditions takes into account load tracking and fast frequency control. The analysis of the normal operating conditions includes the whole cycle range, namely the beginning, middle and end of the cycle, and also requires the operation of the base load and the tracking of various transient power distributions by the load.

Two schemes of 8 absorber rods and 12 absorber rods in gray rod bundles are taken as research bases, the overlapping steps of the power compensation rod groups of the two schemes are set to be 100-90-90, and the insertion limits of the R rod groups are 46 insertion steps under HFP and 28 extraction steps under 0% FP. The SMART program is adopted to carry out normal working condition simulation, the maximum FQT envelope value of different reactor core axial heights under all the working conditions is obtained, and finally the maximum FQT envelope value is compared with the LOCA limit value line, which is shown in a reference figure 6.

Comparing 8

absorber rod solutions

5, 12 absorber rod solutions 6 and LOCA limit 7, it can be seen from fig. 6: the 12 absorber rod schemes in the gray rod cluster and the 8 absorber rod schemes in the gray rod cluster meet the LOCA limit line requirement, and meanwhile, the arrangement of the power compensation rod group and the R rod group of the two schemes is proved to be reasonable.

According to past design experience, the margin of LOCA limit line verification is quite tight, the limit requirement is often exceeded, and the method for processing the overrun is usually to reduce an operation diagram and re-optimize the core loading, wherein the reduction of the operation diagram causes limited operation flexibility, and the re-optimization of the core loading requires a long-time iteration, and in any case, certain cost is paid. To leave more margin for subsequent design, Hualong No. one chooses the 12 absorber rods solution in gray rod bundle as the design solution, since this solution can get 1.6% more margin.

The function of the shutdown rod set is to ensure the negative reactivity necessary for reactor shutdown. When a primary steam line break event (SLB) or boron dilution event occurs, positive reactivity is introduced into the core. To prevent the reactor from re-criticizing after a shutdown, it is desirable that the reactor have sufficient shutdown margin. According to design experience, the safety margin of the main steam pipeline breakage accident is quite tight, and the margin as much as possible is necessary at this stage. The subsequent main steam pipeline breakage accident is analyzed by adopting a shutdown margin limit value 2300pcm (namely, the shutdown margin limit value of the reactor core power of 3050MWth Hualong I is 2300pcm), the criterion requirement cannot be met, and then the SLB analysis is carried out by adopting a shutdown margin limit value 2800pcm so that the criterion requirement can be met. This also proves to be necessary to leave sufficient trip margin here.

The arrangement of the shutdown rod group also considers the following aspects:

the reactor control mode is further optimized to form an expansion space;

support for further optimization of fuel management (24 month refuelling);

support is provided for backup fuel assembly utilization.

And calculating the shutdown allowance of each cycle of Hualong I by adopting three-dimensional reactor core calculation software, and obtaining that the configuration of the control rod of Hualong I meets the requirement of the shutdown allowance through calculation. This proves that not only the power compensation rod set and the temperature adjustment rod set are reasonable from the viewpoint of the shutdown margin, but also the configuration of the shutdown rod set is reasonable.

The shutdown rod configuration under the cold and hot shutdown working conditions is specified in the operation technical specification: some or all of the shutdown rod sets are extracted from the core, and the concentration of boron in the coolant corresponds to the minimum subcritical level required for the extraction of the core by these shutdown rod sets. During some accidents, the shutdown signal is triggered, and the shutdown rod group kept outside the reactor core immediately falls into the reactor core to prevent the reactor core from being re-critical. The accidents that the shutdown rod state configuration under the cold and hot shutdown working conditions can influence the influence comprise: boron dilution accidents, subcritical rod lifting group accidents.

The analysis shows that the boron dilution accident is analyzed under the arrangement of the shutdown rods under five cold and hot shutdown working conditions, and the result can meet the accident analysis criterion. According to the past design experience, the safety margin of the accident of the subcritical lifting bar group is quite tight. Therefore, an optimal scheme is selected by researching the shutdown rod state under the cold and hot shutdown working conditions to be configured into the safety margin of the sub-critical rod lifting group accident.

From the analysis results, it was found that: the other key parameters, except the maximum reactivity introduction rate of the BLX lifting rods, are ARI-SABCD (all the other control rod groups are inserted into the bottom of the reactor core except the shutdown rod groups SA, SB, SC and SD). Although the optimal BLX rod-lifting maximum reactivity introduction rate of 67pcm/s (from ARI-SACD) was slightly less than 76pcm/s (from ARI-SABCD), ARI-SABCD was still the optimal solution from the point of view of subcritical accident analysis. Therefore, the ARI-SABCD scheme is selected for the shutdown rod state configuration under the working condition of Hualong I cold and hot shutdown, and sufficient margin is left for the accident analysis of the subcritical rod lifting group.

The embodiment of the invention aims at the phenomenon that the economy of a nuclear power plant is sacrificed at home and abroad in order to meet the requirement of the safety allowance of the nuclear power plant, and provides a novel design method which can not only not sacrifice the economy of the nuclear power plant but also obtain a certain safety allowance. The method has the advantages that a certain safety margin is obtained from the configuration angle of the pressurized water reactor control rod group for the first time, multiple iterations caused by the fact that the power capacity analysis and the reactive accident analysis of the reactor core of Hualong I are out of limit are greatly reduced, and meanwhile, extra cost caused by the fact that the capacity of system equipment is increased due to the fact that the safety margin is insufficient is avoided. The embodiment of the invention is already applied to Zhangzhou nuclear power engineering, and ensures that Zhangzhou Hualong I can meet corresponding safety margin while improving power.

The above-mentioned embodiments are intended to illustrate the objects, technical solutions and advantages of the present invention in further detail, and it should be understood that the above-mentioned embodiments are merely exemplary embodiments of the present invention, and are not intended to limit the scope of the present invention, and any modifications, equivalent substitutions, improvements and the like made within the spirit and principle of the present invention should be included in the scope of the present invention.

Claims

1. A method of obtaining a safety margin, comprising:

2. The method for obtaining a safety margin as claimed in claim 1 wherein the obtaining a safety margin by adjusting a pressurized water reactor control rod set configuration comprises:

3. The method for obtaining a safety margin as claimed in claim 2 wherein the obtaining a safety margin by adjusting a pressurized water reactor control rod set configuration comprises:

4. The method for obtaining a safety margin according to claim 3, wherein the step of obtaining the safety margin by adjusting the configuration of the pressurized water reactor control rod set comprises the following steps:

the trip bar set configuration is adjusted to obtain a safety margin.

5. The method for obtaining a safety margin as claimed in any one of claims 2-4 wherein the adjusting the configuration of the power compensation rod set to obtain the safety margin comprises:

6. A method of deriving a safety margin as claimed in claim 3 wherein the adjusted set of power compensation bars meets load following simulations in the early life to 85% EOL.

7. The method for obtaining a safety margin according to claim 6 wherein the number of absorber rods in the gray rod cluster of the adjusted power compensation rod group is 8-12; the overlapping steps of the power compensation rod groups are 100-90-90.

8. The method of obtaining a safety margin of claim 7 wherein the configuration of the set of temperature adjustment bars is adjusted to obtain the safety margin; the method comprises the following steps:

9. The method of deriving a safety margin of claim 8 wherein the adjusted configuration set of power compensation rods and the adjusted configuration set of temperature adjustment rods both meet the LOCA limit line requirement.

10. The method of claim 4, wherein the adjusting a trip bar set configuration to obtain a safety margin; the method comprises the following steps: