JP2002357691A - Maintenance support system and method for core vessel welding part, program for functioning computer as the system, and computer-readable storage medium with the program stored therein - Google Patents

Abstract

PROBLEM TO BE SOLVED: To support the formation of a rational maintenance plan for core vessel welding part while ensuring the functional soundness of a reactor. SOLUTION: An influence parameter showing the degree of influence of a stress for the like acting on each baffle former bolt on a baffle former bolt damaging timing is shared by all bolts, and the influence parameter is set and updated so that the damage prediction of baffle former bolts using this influence parameter includes the actual inspection result. The damaging timing of a core vessel welding part is predicted by use of the influence parameter, and the necessity of inspection is determined from whether a damage is generable or not within the total operating scheduled time of the reactor.

Description

DETAILED DESCRIPTION OF THE INVENTION

[0001]

BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to a maintenance system, a method, and a maintenance system for a welded core of a pressurized water reactor, which assists in drafting a maintenance plan for inspection and repair. The present invention relates to a program for causing a computer to function, and a computer-readable recording medium on which the program is recorded.

[0002]

2. Description of the Related Art In a pressurized water reactor, a plurality of baffle plates defining a fuel region in a cylindrical core are generally mounted and supported on baffle mounting plates by baffle former bolts.

A baffle former bolt forming a part of such a furnace internal structure is generally formed of stainless steel or the like, but is exposed to a neutron beam in the furnace. (Irradiation Assisted Str
ess Corrosion Cracking (IASCC).

[0004] A cylindrical core is generally constructed by welding a plurality of thick steel plates. However, a weld between the plurality of thick steel plates is also exposed to a neutron beam. There is a concern that irradiation-induced stress corrosion cracking may occur in the same manner as described above.

[0005] In general, ultrasonic inspection can be employed for damage inspection of such baffle former bolts and core welds. For inspection and inspection, it is usually necessary to take out the furnace internal structure outside the furnace. Since it is necessary, a great deal of cost and time are required for incidental work such as assembly and disassembly of the inspection device.

Although there has been no damage to baffle former bolts or core welds in Japan, there have been few damages to baffle former bolts in overseas reactors with long operating times. Baffle former bolts are regularly inspected at regular intervals.

[0007]

On the other hand, since the core welding portion is located farther away from the core than the baffle former bolt, the environmental conditions such as the neutron irradiation amount and the temperature are slow, and the core is welded to the baffle former bolt. Hateful. However, on the contrary, there was no case of damage because it was hard to be damaged, and it was unclear whether or not inspection was necessary. For this reason, it was not possible to make a plan for a reasonable inspection time while ensuring the soundness of the reactor, and there was a risk that unnecessary inspection costs would be incurred.

SUMMARY OF THE INVENTION The present invention has been made in view of the above problems, and has been made in consideration of the above problem, and has been made in consideration of the above problem. An object of the present invention is to provide a maintenance plan support system, the method, a program for causing a computer to function as the system, and a computer-readable recording medium on which the program is recorded.

[0009]

SUMMARY OF THE INVENTION Based on the above-mentioned object, the inventors of the present invention have found that both the baffle former bolt and the core welded joint cause irradiation-induced stress corrosion cracking and damage, so that relatively early A system and method that can predict the timing of damage to core welds and support rational maintenance planning by appropriately utilizing the inspection results of baffle former bolts for which damage cases are considered available. Invented.

That is, the present invention relates to a maintenance plan support system for supporting the development of a maintenance plan of a core welded portion of a pressurized water reactor, wherein each of the baffle former bolts constituting a reactor internal structure is provided. The baffle former bolt damage time is determined for each baffle former bolt based on the environmental conditions assumed from the structural conditions of the reactor at the position and the influence parameter indicating the degree of influence of the environmental conditions on the damage time of the baffle former bolts. Baffle former bolt damage time prediction means for predicting, and damage baffle former bolt prediction number calculation means for calculating the damage baffle former bolt prediction number with respect to the operation time of the reactor from the baffle former bolt damage time predicted for each baffle former bolt And baffle former bolt damage Inspection result data acquisition means for acquiring inspection result data including the operation time of the reactor when the inspection is performed and the actual number of damaged baffle former bolts indicating the number of damaged baffle former bolts damaged during the operation time, Influence parameter setting means for updating the setting of the influence parameter based on the measured number of damaged baffle former bolts and the predicted number of damaged baffle former bolts during the operation time of the inspection result data each time the inspection result data is acquired. And, for the welded portion of the reactor core, based on the environmental conditions and the influence parameters assumed from the structural conditions of the nuclear reactor, based on the influence parameter, welded portion damage time prediction means for predicting the welded portion damage time, Inspection necessity judgment means for judging the necessity of inspection based on the total scheduled operation time of the reactor. It is an.

According to such a maintenance plan support system for a core welded portion, first, a baffle former bolt which is assumed to be damaged relatively early due to severer environmental conditions to be exposed than a core welded portion. The damage time is predicted using the provisional influence parameter, and the influence parameter used for predicting the damage time of the baffle former bolt is updated and set based on the prediction and the actual inspection results.

At this time, in predicting the damage time of each baffle former bolt, an environmental parameter which varies depending on the position of each baffle former bolt and an influence parameter indicating the degree of the influence of this environmental condition on the damage time of the baffle former bolt. This effect parameter can be shared by all baffle former bolts.

The damage time of each baffle former bolt predicted by using such an influence parameter and the actual inspection result include the inspection result which includes a variation in the tightening force of each baffle former bolt and a manufacturing error. It is inevitable that a shift will occur due to this. However, since the actual number of damaged bolts in which the number of damaged baffle former bolts is bundled cancels out manufacturing errors for each baffle former bolt, the estimated damage time of each baffle former bolt is calculated for all baffle former bolts. According to the estimated number of damaged baffle former bolts calculated by bundling,
Compared to the actual number of the damaged baffle former bolts, the difference between the predicted number and the actual number due to the influence of the manufacturing error can be reduced.

Each time the inspection result data is acquired, the setting of the influence parameter is updated based on the estimated number of damaged baffle former bolts, which is a combination of damage predictions of all baffle former bolts, and the actually measured number of damaged baffle former bolts. By reducing the deviation between the prediction and the actual that cannot be avoided for each bolt, it is possible to effectively use this even with a small inspection result, and the inspection result is used for the impact parameter commonly used for all baffle former bolts. And can be set appropriately.

[0015] When the influence parameter used for predicting the damage time of the baffle former bolt with high reliability is set, the damage time of the core weld is predicted using the influence parameter. It is possible to obtain a prediction of the damage time of a core weld having a high temperature.

[0016] Then, from the predicted damage time of the core weld and the total scheduled operation time of the reactor, the possibility of damage occurring within the scheduled operation period of the reactor is determined, and based on this, the core weld is determined. Since the necessity of the damage inspection is determined, it is possible to suppress the unnecessary damage inspection, reduce the cost, and support a rational maintenance plan.

It is preferable that the baffle former bolt damage time predicting means and the welded part damage time predicting means use a plurality of factors as the environmental conditions and use an influence parameter for each of the plurality of factors.

In this way, a plurality of factors which can independently affect the damage time of the baffle former bolt and the core weld are reflected in the prediction of the damage time of the baffle former bolt and the core weld. By doing so, a more reliable prediction can be made. Further, since the influence parameter for each factor is used, the degree of influence of each factor on the damage time can be separately adjusted, so that the damage time prediction can be flexibly made compatible with the inspection result data.

Specifically, the plurality of factors of the environmental conditions include stress acting on each baffle former bolt and weld, the temperature to which each baffle former bolt and weld are exposed, and each baffle former bolt and weld. It is desirable to include the amount of neutron irradiation applied to the neutron.

In this manner, the effects of stress, temperature, and neutron irradiation amount, which greatly affect the damage time of the baffle former bolt and the core weld, are taken into consideration.
Damage prediction that accurately reflects inspection results can be performed.

In addition, it is preferable that the influence parameter setting means is configured to update the setting of an influence parameter which has a large influence on a baffle former bolt damage time due to a change among the plurality of influence parameters. .

In this way, even when the appropriate values of a plurality of influence parameters are unknown, the setting is updated temporarily by giving priority to the influence parameters having a large influence, thereby quickly adapting to the inspection results. Influence parameters for realizing the predicted damage can be set.

Further, in such a maintenance plan support system for a welded core of a core, the baffle former bolt damage time predicting means may include a baffle former for each baffle in each reactor based on the influence parameter common to a plurality of reactors. The inspection result data acquiring means is configured to predict a baffle former bolt damage time for each bolt, the inspection result data acquisition means is configured to sequentially acquire inspection result data in a plurality of reactors, and the influence parameter setting means is configured to perform the inspection Each time performance data is obtained, the damage baffle former bolt actual measurement number during the operation time of the inspection performance data and the damage baffle former bolt estimated number of the reactor subjected to the damage inspection of the inspection performance data are based on the actual number. Update the setting of the above-mentioned influence parameters common to multiple reactors. It is desirable that configuration to.

As described above, in the prediction of the damage time of the baffle former bolt, the environmental condition where each bolt is placed and the influence parameter are separated, and the update of the influence parameter is performed by the damage inspection of the inspection result data. Based on the estimated number of damaged baffle formabolts obtained in consideration of the structural conditions of the reactor, the common influence parameter can be applied to the baffle formabolts of reactors having different structures.

Further, since the influence parameters common to a plurality of reactors are used and the influence parameters common to a plurality of reactors are updated each time inspection result data is obtained, the inspection parameters for a plurality of reactors are checked. The performance data can be applied to the impact parameter settings for all reactors that are subject to maintenance planning, thereby making more efficient use of less inspection performance and providing more appropriate and rational core welds. Can contribute to drafting a maintenance plan.

Further, a simulation in the operating state of the reactor is performed on the assumption that a crack is generated in the welded portion of the reactor core. It is provided with soundness evaluation means for evaluating whether or not the functional soundness of the reactor is ensured, wherein the inspection necessity / unnecessity judging means is such that the weld damage time is longer than the end of the total scheduled operation time of the reactor. Even if it is early, it is desirable to be configured to make a determination that there is no need for inspection if the soundness is confirmed by the soundness evaluation means.

In this way, even if the weld is damaged during the total scheduled operation time of the reactor as expected in the damage time prediction, the crack that has occurred in the core weld is stopped from growing, In addition, if it can be confirmed that the integrity of the reactor is ensured, it is determined that there is no need for inspection, so it is possible to ensure the integrity of the reactor and contribute to rational maintenance planning. can do.

The present invention also relates to a maintenance plan supporting method for supporting a maintenance plan for a core welded portion of a pressurized water reactor, wherein each of the baffle former bolts constituting a reactor internal structure is provided. The baffle former bolt damage time is determined for each baffle former bolt based on the environmental conditions assumed from the structural conditions of the reactor at the position and the influence parameter indicating the degree of influence of the environmental conditions on the damage time of the baffle former bolts. A step of calculating a predicted number of damaged baffle forma bolts with respect to the operation time of the reactor from the prediction step and the baffle former bolt damage time predicted for each baffle former bolt; and Operating hours of the reactor and damage during the operating hours Acquiring inspection result data including the actual number of damaged baffle former bolts indicating the number of baffle former bolts that have been obtained, and each time the inspection result data is acquired, the damaged baffle former bolts during the operation time of the inspection result data are obtained. Updating the setting of the influence parameter based on the actual measurement number and the damage baffle formabolt prediction number, and for the welded portion of the core, based on the environmental conditions assumed from the structural conditions of the reactor and the influence parameter. A step of estimating a weld damage time, and a step of determining whether or not an inspection is necessary based on the weld damage time and the total scheduled operation time of the nuclear reactor.

According to such a maintenance plan support method for a core welded portion, similarly to the above-mentioned maintenance plan support system, highly reliable using the influence parameter obtained based on the inspection results of the baffle former bolts. It can predict the timing of damage to core welds and support the formulation of rational maintenance plans.

Further, the present invention is a program for causing a computer to function as a maintenance plan support system for supporting the development of a maintenance plan for a core welding portion of a pressurized water reactor, wherein the baffle comprises a reactor internal structure. Regarding the former bolt, the baffle former bolt damage is determined based on the environmental conditions assumed from the structural conditions of the reactor at the position of each baffle former bolt and the influence parameter indicating the degree of the environmental condition affecting the damage time of the baffle former bolt. From the baffle former bolt damage time prediction means for predicting the time for each baffle former bolt and the baffle former bolt damage time predicted for each baffle former bolt, the number of damaged baffle former bolts with respect to the operation time of the reactor is calculated. Damage baffle formabolt prediction Inspection result data including calculation means, and the operation time of the reactor when the damage inspection of the baffle former bolts is performed and the actual number of damaged baffle former bolts indicating the number of damaged baffle former bolts during the operation time. Inspection result data acquisition means to be acquired, and each time the inspection result data is acquired, the influence parameter is determined based on the measured number of damaged baffle former bolts and the predicted number of damaged baffle former bolts during the operation time of the inspection result data. Influence parameter setting means for updating the setting of, and, for the welded portion of the reactor core, the weld damage time prediction for predicting the weld damage time based on the environmental conditions assumed from the structural conditions of the reactor and the influence parameter Means, the weld damage time, and the total scheduled operation time of the reactor, it is determined whether the inspection is necessary. And check necessity judgment means for a program for causing a computer to function with.

According to such a program, the computer can function as the above-described maintenance plan support system for the core welded portion, and the above-described effects can be obtained.

Further, the present invention is a computer-readable recording medium on which a program for causing a computer to function as a maintenance plan support system for supporting the development of a maintenance plan for a core weld of a pressurized water reactor is recorded. The baffle former bolts constituting the in-reactor structure have environmental conditions assumed at the position of each baffle former bolt based on the structural conditions of the reactor and the degree of influence of the environmental conditions on the damage time of the baffle former bolts. Baffle former bolt damage time prediction means for predicting the baffle former bolt damage time for each baffle former bolt based on the parameters, and the reactor operating time from the baffle former bolt damage time predicted for each baffle former bolt Baffle Formabolt Prediction for Steel Baffle former bolt calculation means for calculating the number of damaged baffle former bolts indicating the operating time of the reactor and the number of damaged baffle former bolts during the operation time when damage inspection of the baffle former bolts is performed Inspection result data acquisition means for acquiring inspection result data including an actual measurement number, and each time the inspection result data is acquired, the actual number of damaged baffle former bolts and the damaged baffle former bolt in the operation time of the inspection result data Influence parameter setting means for updating the setting of the influence parameter based on the predicted number, and for the welded portion of the core, based on the environmental conditions assumed from the structural conditions of the reactor and the effected parameter, the weld damage. Weld damage time prediction means for predicting the time, and the weld damage time And check necessity determining means for determining the necessity of inspection and a total scheduled driving time of the child furnace, a computer-readable recording medium on which a program is recorded for causing a computer to function with.

According to the computer-readable recording medium on which such a program is recorded, the above-described operation and effect can be obtained by causing the computer to function as the above-described maintenance plan support system for the core welded portion.

[0034]

BRIEF DESCRIPTION OF THE DRAWINGS FIG. 1 is a block diagram showing a maintenance support system for a core welding part according to the present invention.

First, prior to the description of the maintenance plan support system, the configuration of the internal structure of the reactor will be outlined.

FIG. 1 is an explanatory perspective view showing the internal structure of a nuclear reactor. As shown in this figure, the reactor internal structure of the nuclear reactor provided in the reactor vessel 80 is divided into an upper internal structure 81 and a lower internal structure 90. Upper furnace internal structure 8
Reference numeral 1 mainly controls and guides the control rod cluster, and includes an upper core plate 82, an upper core support column 83, an upper core support plate 84, a control rod cluster guide tube 85, and the like. The lower reactor internal structure 90 mainly supports a fuel assembly (core) 91, and includes a reactor core 92, a baffle plate (core baffle) 93, a baffle mounting plate (core baffle mounting plate) 94, and a lower core plate. 95, a lower core support column 96, a lower core support plate 97, and the like.

FIG. 2 is a horizontal sectional view of the lower furnace internal structure 90.
FIG. 3 is a perspective explanatory view showing an attached state of the baffle plate 93, and FIG.

As shown in FIG. 2, the core core 92 of the lower core structure 90 has a circular horizontal cross section.
As shown in the figure, a plurality of baffle plates 93 forming a boundary of a core region (fuel region) are attached to the inside of the core 92 via a baffle attachment plate 94. The core baffle mounting plates 94 are arranged in a plurality of stages (for example, seven stages) at predetermined intervals in the height direction, and a plurality of baffle plates 93 (for example, two Baffle former bolt 9)
8 is concluded.

Generally, about 600 to 1,000 baffle former bolts 98 are used in each reactor, and a plurality of baffle plates 93 are used to fix each baffle plate 93 to the baffle mounting plate 94.

As shown in FIG. 4, the core 92 includes a welded portion 92 in which a plurality of thick steel plates run in the circumferential direction and the vertical direction.
1 is constructed by welding. This core 9
A core region is formed in substantially the lower half of the core 2 by the above-described baffle plate 93 and the like. A heat shield plate 99 is provided around this portion in close proximity to the outer wall of the reactor core 92.

Next, a maintenance plan support system for supporting the maintenance plan formulation for inspection and repair of the core 92, particularly the welded portion 921 thereof, will be described in detail.

FIG. 5 is an overall configuration diagram of a maintenance plan support system for a core welding section according to an embodiment of the present invention.

This maintenance plan support system is composed of a computer provided with a CPU, a memory, an external storage device, and input / output means such as a keyboard, a mouse, and a monitor. Functionally, as shown in FIG. Baffle former bolt damage time prediction means 10, damaged baffle former bolt number calculation means 20, inspection result data acquisition means 30
, An influence parameter setting means 40, a weld damage time predicting means 50, a soundness evaluating means 55, an inspection necessity determining means 60, and a structural data storage section 70.

Each of these functions is realized by a program, and the program is provided by a download from a network such as the Internet or a computer-readable recording medium such as a CD-ROM. Hereinafter, each functional element will be described.

The baffle former bolt damage time predicting means 10 predicts the baffle former bolt damage time for each baffle former bolt.

The damage of the baffle former bolt is caused by irradiation-induced stress corrosion cracking (Irradiation Assisted Stress Corrosion Cracking).
ion Cracking (IASCC).
The crack initiation time due to this irradiation-induced stress corrosion cracking is considered to be the sum of the time until the susceptibility to stress corrosion cracking occurs and the time from the occurrence of the susceptibility to damage. The former is expected to be short,
In addition, since the safety side (earlier damage time) is predicted, here, the bolt damage time is predicted using only the latter time.

The prediction of the bolt damage time is based on environmental data (environmental conditions) assumed at the position of each baffle former bolt and an influence parameter indicating the degree of influence of this environmental condition on the damage time of the baffle former bolt. Done.

The environmental data is obtained for each baffle former bolt by the baffle former bolt damage timing predicting means 10 from the reactor structural data stored in the later described structural data storage unit 70. The environmental data uses a plurality of factors that can influence the bolt damage time independently of each other. That is, a plurality of factors are calculated as environmental data for each volt. By using multiple factors as environmental data (environmental conditions),
A more reliable prediction can be realized by reflecting the influence of each factor on the damage time of the bolt. In the present embodiment, three such factors are assumed: a stress acting on each baffle former bolt, a temperature to which each baffle former bolt is exposed, and a neutron irradiation amount irradiated to each baffle former bolt.

The influence parameters are stored in an influence parameter setting means 40, which will be described later, and are separated from different environmental conditions depending on the position of each bolt, so that they can be used in common by all bolts. . Further, bolts of a plurality of reactors can be used in common. This influence parameter is set at least for each factor of the environmental data, and may include a proportionality constant or the like that bundles the degree of influence of all the factors on the bolt damage timing, as described later. As described above, since the influence parameter for each factor is used, the degree of influence of each factor on the damage time can be separately adjusted for each factor.

The following equation (1) is an example of a prediction equation for predicting a bolt damage time from such environmental data and influence parameters.

[0051]

(Equation 1)

Where, t _N : bolt damage time (time until damage) k: proportionality constant (influence parameter of the whole) σ: stress acting on bolt σ _Y : yield stress of bolt material n: stress dependence Influence parameter T: Temperature to which volt is exposed R: Gas constant Q: Influence parameter (temperature of material activation) Φ: Neutron irradiation amount irradiated to volt m: Neutron irradiation amount dependence Influence parameter indicating

The stress σ acting on each baffle former bolt is, as shown in FIG.
(A)), baffle thermal deformation (Fig. 6 (b)), and swelling that fluctuates with the operation time of the reactor (Fig. 6
(C)), neutron irradiation hardening (FIG. 6 (d)), and creep under irradiation (FIG. 6 (e)). This is calculated as a value that fluctuates over time as shown in FIG. As shown in FIG. 6A, the bolt tightening force has a variation in the tightening torque. Also, since the size of other elements differs depending on the position of the baffle former bolt,
The sum of these stresses also differs depending on the position of the baffle former bolt.

The temperature T to which each baffle former bolt is exposed is obtained from the structure data in the structure data storage unit 70 according to the position of each baffle former bolt. In addition, since this temperature T affects each element of baffle thermal deformation, swelling, and creep of the stress σ,
These elements are calculated according to the temperature T.

The neutron irradiation amount Φ applied to each baffle former bolt is obtained from the structure data in the structure data storage unit 70 according to the position of each baffle former bolt. The neutron irradiation amount Φ affects each of the swelling, neutron irradiation hardening, and creep of the stress σ, and these elements are calculated according to the neutron irradiation amount Φ.

The influence parameters include an influence parameter n indicating stress dependency, an influence parameter Q indicating temperature dependency, an influence parameter m indicating neutron irradiation dose dependency, and a proportional constant k which is the sum of the effects of environmental data factors. The four. The influence parameters n, Q, and m express the relative degree of influence of each factor on the bolt damage time, and k is a proportionality constant summarizing these.

Incidentally, the baffle former bolt damage time prediction means 10 can predict the bolt damage time of the baffle former bolt of an arbitrary reactor if there is structural data of the reactor. For this reason, the baffle former bolt damage time is predicted for the reactors for which the maintenance plan is to be prepared, and if one reactor is to be subjected to the maintenance plan, the inspection results of other reactors When data is obtained, the baffle former bolt damage time of this other reactor will be predicted using the structural data of this other reactor in order to use the inspection performance data of this other reactor. ing.

Damage baffle former bolt number calculation means 2
0 is to calculate the number of damaged baffle former bolts with respect to the operating time of the reactor from the baffle former bolt damage time predicted for each baffle former bolt by the baffle former bolt damage time predicting means 10.

Specifically, the estimated number of damaged baffle former bolts is calculated by accumulating the number of bolts whose bolt damage time has arrived at each time point from the start of operation of the nuclear reactor.

The estimated number of damaged baffle former bolts may be the number of bolts or the ratio of the number of bolts for which damage is predicted to the total number of bolts (damage rate).

The damaged baffle former bolt number calculating means 20 also calculates the estimated number of damaged baffle former bolts for an arbitrary reactor if the baffle former bolt damage time predicting means 10 predicts the bolt damage time for each bolt. Can be calculated.

If some or all of the baffle former bolts have been replaced since the start of the operation of the reactor, the damaged baffle formers for the bolts at the beginning of the operation of the reactor if this replacement was not carried out An estimated bolt count is calculated. The estimated number of damaged baffle former bolts calculated in this way is used for comparison between the prediction and the actual result in the influence parameter setting means 40 described later. This is done in order to make an appropriate comparison with the condition of the prediction and the result being equal.

On the other hand, in a state where an appropriate influence parameter is set by comparing the prediction and the actual result of the damage of the baffle former bolt, the inspection necessity determining means 60 is required.
When the management time is predicted in the above, the damaged baffle former bolt number predicting means 20 calculates the predicted number of damaged baffle former bolts using the predicted bolt damage time starting from the replacement time for the replaced bolt. I do. The predicted number of damaged baffle former bolts calculated in this way is a prediction in which bolt replacement is taken into consideration and is in accordance with the actual situation, so that a more appropriate maintenance plan can be provided.

The inspection result data acquiring means 30 is for acquiring inspection result data when a damage inspection of the baffle former bolt is performed.

The inspection result data includes at least the operation time of the reactor at the time of the inspection and the actual number of damaged baffle former bolts indicating the number of damaged baffle former bolts during the operation time. .

Inspection results obtained in actual damage inspection include the tightening force of each bolt, manufacturing errors, and the like, and it is unavoidable that there will be a deviation from the theoretical (predicted) damage time. However, the actual measurement number of the damaged baffle former bolts in which the number of damaged bolts is bundled cancels out the manufacturing error of each bolt, so that the deviation from the theoretical damage time (estimation) can be reduced.

This damage inspection may be performed by a general ultrasonic inspection test or other inspection methods. Further, a partial inspection may be performed in which only a part of the baffle former bolt included in the nuclear reactor is inspected by utilizing the symmetry of the core.

As a form of acquisition of the inspection result data, the data may be directly received from the inspection / inspection device, data recorded on an arbitrary recording medium may be read, or a keyboard input by an operator may be used. Depending on the form, the inspection result data acquiring means 30 includes a receiving means for receiving data from an inspection / inspection device or the like, a reading means for reading data recorded on a recording medium or the like, or an input means or interface for receiving data input. Is configured as

The actual number of damaged baffle former bolts may be the number of bolts or the ratio of the number of damaged bolts to the total number of bolts (damage rate).

The inspection result data acquiring means 30 sequentially acquires not only the inspection result data of the reactor for which the maintenance plan is made but also the inspection results of other reactors. Is available. The inspection result data of the other reactors includes at least information for specifying the reactor that has been checked or information for specifying the structural conditions (structural data) of the reactor that has been checked. Further, for a reactor in which baffle former bolts have been replaced, information indicating the number of replaced bolts and the replacement time is also included.

The inspection result data acquisition means 30 stores and holds the acquired inspection result data.

The influence parameter setting means 40 is based on the prediction by the baffle former bolt damage time predicting means 10 and the damage baffle former bolt number calculating means 20 and the inspection result acquired by the inspection result data acquiring means 30 to determine the above-mentioned influence parameter. The setting of is updated.

The processing by the influence parameter setting means 40 is the update of the influence parameter because the prediction by the baffle former bolt damage time predicting means 10 and the damage baffle former bolt number calculating means 20 is based on the influence already set before the update. This is because it was performed by parameters.

The update of the influence parameter is performed by comparing the actual number of damaged baffle former bolts (actual number) during the operation time of the acquired inspection result data with the predicted number of damaged baffle former bolts (estimated) during the operation time. The predicted number of baffle former bolts matches the measured number of damaged baffle former bolts, or the degree of mismatch is within an allowable range (hereinafter, simply referred to as “match”).
It is determined by changing the setting of the influence parameter such that the estimated number of damaged baffle former bolts matches (envelopes) the actually measured number of damaged baffle former bolts.

As will be described later, when the prediction does not match the actual result, the influence parameter is changed and temporarily set, and the temporary setting is repeated until the prediction based on the temporarily set influence parameter matches the actual result. To do it.

If the inspection performance data relates to the reactor in which some of the baffle former bolts have been replaced, the replaced baffle former bolts are considered to be those in which the baffle former bolts at the beginning of the reactor operation were damaged. The actual number of damaged baffle former bolts for all bolts at the beginning of operation is used for comparison between treatment, prediction and actual results. On the other hand, as the predicted number of damaged baffle former bolts, the predicted number of damaged baffle former bolts for all bolts at the start of operation is used. By doing this,
Forecast and actual conditions can be compared equally,
It can contribute to the setting of appropriate influence parameters.

If the actual number of damaged baffle former bolts in the inspection result data is 0 (zero),
Assuming that one damaged bolt was generated immediately after the inspection,
Update the setting of the influence parameter. This is desirable because it is possible to realize damage prediction on the safer side.

In updating the setting of the influence parameter, the estimated number of damaged baffle former bolts, which is a combination of the damage prediction and the actual damage of all the baffle former bolts, is compared with the actually measured number of damaged baffle former bolts. By comparing and examining each bolt, it is possible to reduce the unavoidable difference between the predicted and actual values. As a result, it is possible to update the setting of the appropriate influence parameter by effectively using the small number of inspection results.

In updating the influence parameters, not only the inspection result data acquired most recently but also all the inspection result data previously acquired and stored in the inspection result data acquiring means 30 are to be compared and examined. Is done.

A plurality of inspection results (the actual number of damaged bolts in the inspection result data) for the same reactor all match the prediction (the estimated number of damaged baffle former bolts) for the reactor (envelope). It is determined whether or not to do so. The same applies to reactors of the same type having the same structural conditions (structural data).

A plurality of inspection results for reactors having different structural conditions (structural data) are determined for each reactor of the same type as to whether or not the results match predictions. The prediction is updated to the matching impact parameter.

As described above, in this embodiment, in the prediction equation (1), the stress σ, the temperature T, and the neutron irradiation amount Φ for each baffle former bolt are assumed as the environmental conditions (environmental data). As the influence parameters, four parameters are set: n indicating stress dependency, Q indicating temperature dependency, m indicating neutron irradiation dose dependency, and proportional constant k summing up these.

As described above, when a plurality of influence parameters are set, the influence parameter setting means 40 previously obtains the order of the magnitude of the influence of each of the influence parameters on the bolt damage time by a parameter survey, and By using the results and updating the setting of the influence parameters that have a large effect on the bolt damage time due to the fluctuation,
An appropriate set of influence parameters can be set early.

Specifically, the influence parameters n, Q, m, k
Among them, the proportional constant k is the overall proportional constant, and its variation is directly reflected in the bolt damage time, and is the most influential parameter.

As for the other influence parameters, some values are set within the range assumed for the influence parameter to be considered while fixing the other than the influence parameter to be studied, and the bolt damage time prediction is performed. When the influence of the influence parameter is examined, the influence of the influence parameter Q is relatively large.

Therefore, in the influence parameter setting means 40, when the prediction and the actual result do not match, first, the influence parameter k is adjusted with priority over the others to roughly match the predicted and actual results. Fine adjustment is performed based on the influence parameters Q, n, and m so that the prediction matches the actual result.

As described above, even if there are a plurality of unknown influence parameters, by temporarily updating the setting with the priority given to the influence parameter having a large influence, it is possible to quickly estimate the bolt damage suitable for the inspection result. The effect parameters to be realized can be set.

The influence parameter setting means 40 stores and holds the updated influence parameters.

The updating of the setting of the influence parameter is as follows.
It is preferable to perform this every time the inspection result data acquisition means 30 newly acquires inspection result data. This is because the acquired inspection results are reflected in the maintenance plan at an early stage, thereby contributing to more appropriate maintenance plan execution.

As will be described later, the influence parameter setting means 40 also performs initial temporary setting of the influence parameter when the influence parameter has not been set yet.

FIG. 7 is a graph showing an example of the estimated number of damaged baffle former bolts thus obtained. In addition,
The vertical axis is the estimated number of damaged baffle former bolts, expressed as a ratio of the number of damaged bolts to the total number of bolts. The circles in the figure plot the actual number of damaged baffle former bolts based on the actual inspection data of the reactor for which damage was predicted, and the triangles indicate the reactors for which damage was predicted. It plots the actual number of damaged baffle former bolts based on the inspection data of the same type of reactor.

FIG. 8 shows that new inspection result data is obtained.
It is a graph which shows the change of the damage baffle Formabolt prediction number when the setting update of an influence parameter is performed.

In this figure, the curve indicated by the solid line indicates the estimated number of damaged baffle former bolts calculated based on the influence parameters before the acquisition of the actual inspection data.

At this time, the damage inspection of the baffle former bolt is performed at the time indicated by the triangle in the figure, and if the inspection result data is obtained, the setting of the influence parameter is updated according to the inspection result data. A predicted number of damaged baffle former bolts that enclose this inspection result data is obtained.

Specifically, if the actual number of damage baffle former bolts in the inspection result data is A in the figure, the damage prediction envelops the inspection result, and the prediction is not corrected.

On the other hand, if the actual number of damaged baffle former bolts in the inspection result data is B in FIG.
Since the number of damaged bolts is 0 (zero), the influence parameter is set and updated so that the predicted number of damaged bolts envelops this B while standing on the safe side that the damage of bolts occurs immediately after this inspection. The estimated number of baffle former bolts is corrected to a dashed line in the figure.

Further, if the actual number of damaged bolts in the inspection result data is C in the figure, the influence parameter is updated so that the predicted number of damaged bolts envelops this C,
The estimated number of damaged baffle former bolts is corrected to a two-dot chain line in the figure.

The weld damage time predicting means 50 predicts the damage time of the core weld. When there are a plurality of welds, the prediction is made for each weld.

[0099] Damage to the core weld was caused by irradiation-induced stress corrosion cracking (Irra
diation Assisted Stress Corrosion Cracking: IAS
CC). For this reason, the weld damage time prediction means 50 has substantially the same configuration as the baffle former bolt damage time prediction means 10 described above, and is set by the influence parameter setting means 40 based on the baffle former bolt inspection results. Using the influence parameter, the damage timing of the core welded portion is predicted by the above-mentioned prediction formula (1).

The weld damage time is not predicted as a specific time, but rather at a specific time, for example, a time later than the end of the total scheduled operation time of the reactor.
It may be predicted with a range.

The soundness evaluating means 55 evaluates whether or not the functional soundness of the nuclear reactor is ensured even if a crack is assumed to occur in the welded portion of the core.

The evaluation of the soundness is performed by a simulation assuming that a crack is generated in the core welded portion.
If the crack propagation assumed in the result of the simulation is stopped and the functional integrity of the reactor is ensured by the remaining weld, the reactor is evaluated as having soundness.

In this evaluation, various knowledge affecting the simulation results obtained at the time of evaluation, such as the magnitude and distribution of residual stress determined by the welding state, material properties, and behavior when stress is applied, etc. In such cases, the latest knowledge is applied.

Specifically, in this simulation, a structural portion including a welded portion is modeled by a finite element method from the structural data of the nuclear reactor, a virtual crack is given in the model, and a residual portion of the welded portion is formed. Simulate crack growth by applying stress, thermal stress acting on each part during operation of the reactor, and stress due to its own weight, and confirm whether this stops.

Further, a simulation for inputting an assumed seismic wave in a state where the propagation of the crack has stopped is performed, and it is evaluated whether or not the soundness of the reactor is secured in the remaining welded portion.

By performing such a soundness evaluation, it is confirmed whether or not the functional soundness of the reactor is secured even if the welded portion is damaged. Prevention and safety can be ensured while contributing to rational maintenance planning.

The inspection necessity judging means 60 judges whether or not it is necessary to inspect the welded portion based on whether or not the welded portion is damaged during the total scheduled operation time (operating life) of the reactor. It is.

Further, even if damage to the weld occurs during the total scheduled operation time of the reactor or if it is confirmed that the integrity of the reactor can be ensured even if a crack occurs in the weld. Only, it is determined that damage inspection of the weld is not necessary.

More specifically, the weld damage time predicted by the weld damage time predicting means described above is predicted with respect to the total scheduled operation time of the reactor, which is input in advance or input by an operator or the like. It is determined whether the vehicle arrives during the scheduled operation time.

FIG. 9 is an explanatory diagram for determining whether or not a weld is damaged during the total scheduled operation time of the reactor.

In this figure, the solid line shows the SCC generation time when the initial stress including the stress remaining in the welded portion by welding is assumed to have various magnitudes. On the other hand, the broken line indicates the magnitude of the initial stress acting on the actual core weld. That is, in the welded portion of the nuclear reactor, the predicted damage time of the welded portion is a time at which the gradually decreasing solid line and the broken line intersect.

On the other hand, the total scheduled operation time of this reactor is set to 60
If the time is 10,000 hours, the damage prediction time of the welded portion is much earlier than the time when 600,000 hours have elapsed. That is, it is considered that no damage to the weld occurs during the total scheduled operation time of the reactor.

If it is predicted from the damage prediction of the welded portion that no damage will occur within the total scheduled operation time of the reactor, the stability against the virtual crack by the soundness evaluation means 55 is further confirmed, and the stability is confirmed. If the property is confirmed, the operator or the like is notified that there is no need for inspection.

The structural data storage unit 70 stores the structural data of the nuclear reactor.

The structural data is data expressing the structural conditions of the nuclear reactor, and is structural data necessary for obtaining environmental data (environmental conditions) in the baffle former bolt damage time prediction means 10 and the weld damage time prediction means 50. At the same time, the soundness evaluation means 55 includes structural data used for a simulation for confirming the functional soundness of the nuclear reactor.

More specifically, the structural data includes shape data and material data of the furnace internal structure, and data such as a load and a vibration acting upon an assumed earthquake or accident.

The structural data storage unit 70 stores structural data of all the reactors for which the inspection result data is acquired, in addition to the structural data of the reactor for which the maintenance plan is to be prepared. I have. These structural data can be called up based on the information specifying the reactor or the structure of the reactor acquired by the inspection result data acquiring means 30.

Next, a procedure for preparing a maintenance plan in such a maintenance plan support system will be described.

First, the procedure for setting and updating the influence parameter from the inspection results of the baffle former bolt is described with reference to FIG.
This will be described with reference to the flowchart of FIG.

The setting update of the influence parameter is performed every time the inspection result data acquiring means 30 acquires a new inspection result.

In setting the influence parameters, first, the influence parameters are assumed by the influence parameter setting means 40 (step S10). If there is an already set influence parameter and this is to be updated, the influence parameter stored in the influence parameter setting means 40 is used. If the influence parameter has not been set yet, the influence parameter setting means 40 An appropriate value is assumed according to the physical significance of each influence parameter.

Subsequently, based on the structural data stored in the structural data storage unit 70, environmental data is calculated for each baffle former bolt by the baffle former bolt damage time prediction means 10 (step S12), and environmental data of each bolt is obtained. The bolt damage time is predicted for each baffle former bolt based on the above-mentioned influence parameter assumed to be (step S14). The calculation of the environmental data in step S12 and the prediction of the baffle former bolt damage time in step S14 may be repeatedly performed for each baffle former bolt.

When the bolt damage time of each baffle former bolt is predicted in this way, the damaged bolt number calculating means 20
Thereby, the estimated number of damaged baffle former bolts with respect to the operation time of the reactor is calculated (step S16).

The reactors for which the prediction of the damage time of the baffle former bolt and the calculation of the predicted number of damaged baffle former bolts are all the reactors for which the actual inspection data have been acquired.

The influence parameter setting means 40 determines whether or not the estimated number of damaged baffle former bolts thus calculated envelops the inspection result data acquired by the inspection result data acquiring means 30 (step S1).
8). Note that, as the inspection result data to be determined, not only the inspection result data that has just been acquired, but also all the inspection result data that has already been acquired. In this determination, it is determined whether or not each inspection result data is enveloped in the estimated number of damaged baffle formal bolts of each reactor for each reactor targeted by the inspection result data.

If the estimated number of damaged baffle former bolts does not enclose the inspection result data (N in step S18)
O) Returning to step S10, the influence parameters are assumed again. At this time, the influence parameter (specifically, Q) which is given priority from the factor having a large influence (proportional constant k) using the result of the parameter survey is assumed to be corrected.

If the predicted number of damaged baffle former bolts encloses the inspection result data (YE in step S18)
S), the influence parameter at this time is updated and registered as a future influence parameter by the influence parameter setting means 40 (step S20).

For different types of reactors, commonly used influence parameters are set and updated so as to bridge the difference between the results and predictions studied for each type of reactor. This makes it possible to increase the data by sharing the actual data obtained for a plurality of reactors of different types, and to set the influence parameters with higher accuracy.

Next, the entire flow of the maintenance plan drafting procedure including the procedure for updating the setting of the influence parameters will be described.
This will be described with reference to FIG.

The purpose of this maintenance plan is to determine whether or not the core weld is required to be inspected, and to determine whether damage can occur to the core weld within the total scheduled operation time of the reactor. It consists of a damage possibility study stage to determine and a soundness study stage to determine whether damage will affect the soundness of the reactor, even if damage has occurred.

In the damage possibility examination stage, the apparatus is in a standby state until new inspection result data is obtained (NO in step S50). When new inspection result data is acquired (YES in step S50), the core welding portion is obtained. The likelihood of damage occurring is reviewed. At the start of the maintenance plan drafting, the initially obtained inspection result data is treated as new inspection result data, and step S50 is YES.
Becomes

When the new inspection result data is obtained, the setting of the influence parameter is updated according to the above-described procedure (step S52).

As a result, if the influence parameter has been changed (YES in step S54), the welding part damage time prediction means 50 based on the changed new influence parameter.
Thereby, damage prediction of the core welding portion is performed (step S56). At the start of the maintenance planning, the influence parameters set for the first time are treated as changed, and the result in step S54 is YES.

The inspection necessity judging means 60 judges whether or not the predicted weld damage prediction time arrives within the total scheduled operation time of the reactor (step S58). If the result of this determination is that there is no damage to the welded portion within the total scheduled operation time of the reactor (NO in step S58), there is no need for inspection, so that an operator or the like is notified (step S60). The examination results will be maintained until new inspection result data is obtained.

After it is once determined that there is no need for inspection, new inspection result data is obtained (YES in step S50), and the setting of the influence parameter is updated taking into account the new inspection result data. Also (step S52), if the influence parameter has not been changed (NO in step S54), the result does not change even if the damage prediction of the core weld is performed using the same influence parameter as before, so that it is notified that the inspection is unnecessary. The process returns to step S50.

On the other hand, if it is determined that the core weld is damaged (YES in step S58), the process proceeds to the soundness examination stage.

In this soundness examination stage, first, the soundness evaluation means 55 assumes that cracks will be generated in the core welded portion, and stops the growth of the cracks. It is evaluated whether or not the soundness of the nuclear reactor is ensured against an assumed earthquake or the like (step S62).

As a result of the soundness evaluation, if it is determined that there is no soundness (NO in step S64), the operator or the like is notified that an inspection is required (step S6).
8). At this time, based on the predicted damage time, a predetermined time sufficiently earlier than that may be proposed as the inspection time.

On the other hand, if it is determined that there is soundness (step S64), even if the welded portion is damaged as expected in the nuclear reactor, the reactor is still in a state in which the damaged portion is damaged. Since it is possible to operate the vehicle while ensuring the above, the operator or the like is notified that the inspection is unnecessary (step S66).

In this state, if a new condition for soundness evaluation is obtained as new knowledge that may affect the result of soundness evaluation (YES in step S70), the process returns to step S62 to perform the soundness evaluation again. .

If new inspection result data is obtained (YES in step S72), there is no basis for the premise of the soundness examination stage that damage may occur in the core welding portion, so that the process returns to step S52 and damage is possible. Move to gender study stage.

On the other hand, if there is no input of new soundness evaluation conditions or acquisition of new inspection result data (step S70,
(NO in S72), and waits until these are newly obtained.

As described above, the present invention has been described in accordance with the embodiments. However, the maintenance plan support system for a core welded portion according to the present invention is not limited to the above embodiments, and the gist of the present invention is changed. Modifications can be made appropriately within a range not to be performed.

[0144]

As described above, according to the present invention, the baffle former bolt damage time for each baffle former bolt is determined based on environmental conditions that differ depending on the position of each baffle former bolt, and the influence parameters common to all baffle former bolts. The number of damaged baffle former bolts and the estimated number of damaged baffle former bolts with respect to the operation time of the reactor are calculated by combining the bolt damage times of all baffle former bolts. Since the setting of the influence parameter is updated based on the actual number of former bolts, the deviation between the actual inspection results and the predictions, including manufacturing errors, which cannot be avoided for each baffle former bolt, is reduced. This can be used effectively and is common to all baffle former bolts Reflecting the inspection results to the effect parameters used can be appropriately set. When the influence parameter used for predicting the damage time of the baffle former bolt with high reliability is set in this manner, the damage time of the welded portion of the core is predicted using the influence parameter, so that the reliable core So a prediction of the damage time of the weld can be obtained. Then, based on the predicted damage time of the core weld and the total scheduled operation time of the reactor, the possibility of damage occurring within the scheduled reactor operation period is determined,
Based on this, the necessity of the damage inspection of the core weld is determined, so that unnecessary damage inspection can be suppressed, the cost can be suppressed, and a reasonable maintenance plan can be supported.

[Brief description of the drawings]

FIG. 1 is a perspective explanatory view showing a reactor internal structure of a nuclear reactor.

FIG. 2 is a horizontal sectional view of a lower furnace internal structure.

FIG. 3 is an explanatory perspective view showing an attached state of a baffle plate.

FIG. 4 is a partially sectional perspective view of a reactor core.

FIG. 5 is an overall configuration diagram of a maintenance plan support system for a core welding portion according to an embodiment of the present invention.

FIG. 6 is a graph showing an example of a stress σ acting on each baffle former bolt, where (a) is a bolt tightening force,
(B) is baffle thermal deformation, (c) is swelling,
(D) shows the effect of neutron irradiation hardening, (e) shows the effect of creep under irradiation, and (f) shows the stress σ obtained by adding them.

FIG. 7 is a graph showing an example of a predicted number of damaged baffle former bolts.

FIG. 8 is a graph exemplifying a change in setting of an influence parameter.

FIG. 9 is an explanatory diagram for determining whether damage to a weld occurs during the total scheduled operation time of the reactor.

FIG. 10 is a flowchart showing a procedure for updating the setting of an influence parameter.

FIG. 11 is a flowchart showing an overall flow of a maintenance plan drafting procedure.

[Explanation of symbols]

 DESCRIPTION OF SYMBOLS 10 Baffle former bolt damage time prediction means 15 Barrel former bolt damage time prediction means 20 Damage baffle former bolt prediction number calculation means 25 Damage barrel former bolt prediction number calculation means 30 Inspection results data acquisition means 40 Influence parameter setting means 50 Weld joint damage time Prediction unit 55 Soundness evaluation unit 60 Inspection necessity judgment unit 70 Structural data storage unit 80 Reactor vessel 90 Lower reactor internal structure 91 Fuel assembly (core) 92 Core core 921 Core core welded part 93 Core baffle (baffle plate) 94) Core baffle mounting plate (former plate) 98 Baffle former bolt

 ──────────────────────────────────────────────────続 き Continued on the front page (71) Applicant 000230940 Japan Nuclear Power Co., Ltd. 1-1, Kanda-Midshiro-cho, Chiyoda-ku, Tokyo (71) Applicant 000006208 Mitsubishi Heavy Industries, Ltd. 2-5-1 Marunouchi, Chiyoda-ku, Tokyo (72) Inventor Masashi Kameyama 3-2-2, Nakanoshima, Kita-ku, Osaka-shi Kansai Electric Power Co., Inc. (72) Inventor Koji Hasegawa 3-2-2, Nakanoshima, Kita-ku, Osaka Kansai Electric Power Co., Ltd. 72) Inventor Hidetaka Shimizu 1-2-2 Odorihigashi, Chuo-ku, Sapporo City Inside Hokkaido Electric Power Co., Inc. Inventor Takeshi Yoshinaga 1-1-1, Kanda-Midshiro-cho, Chiyoda-ku, Tokyo Inside of Nippon Nuclear Power Co., Ltd. Kobe Shipyard, (72) Inventor Noboru Kubo 1-1-1, Wadasakicho, Hyogo-ku, Kobe City, Kobe Shipyard (72) Inventor Masaaki Katayama, Kobe, Kobe 1-1-1, Wadazakicho, Hyogo-ku, Tokyo F-term (reference) in Kobe Shipyard, Mitsubishi Heavy Industries, Ltd. 2G075 AA05 CA07 CA44 DA16 EA03 EA09 FB05 FB16 FB18 GA09

Claims (9)

[Claims] 1. A maintenance plan support system for supporting the development of a maintenance plan for a core welding portion of a pressurized water reactor, comprising: a baffle former bolt constituting a reactor internal structure; A baffle former for predicting a baffle former bolt damage time for each baffle former bolt based on environmental conditions assumed from furnace structural conditions and an influence parameter indicating the degree of influence of the environmental conditions on the damage time of the baffle former bolts Bolt damage time prediction means, damage baffle former bolt prediction number calculation means for calculating a damage baffle former bolt prediction number with respect to reactor operation time from baffle former bolt damage time predicted for each baffle former bolt, baffle former When bolts are inspected for damage Inspection result data acquisition means for acquiring inspection result data including the operating time of the nuclear reactor and the actual number of damaged baffle former bolts indicating the number of baffle former bolts damaged during the operation time, and acquiring the inspection result data The influence parameter setting means for updating the setting of the influence parameter based on the actual number of damaged baffle former bolts and the predicted number of damaged baffle former bolts during the operation time of the inspection result data, and welding of the reactor core A weld portion damage time prediction means for predicting a weld damage time based on the environmental conditions and the influence parameters assumed from the structural conditions of the reactor, and the weld damage time and the total operation schedule of the reactor. An inspection necessity judging means for judging necessity of inspection from time. Maintenance planning support system of the contact portion. 2. The baffle former bolt damage time predicting means and the weld damage time predicting means use a plurality of factors as the environmental condition and use an influence parameter for each of the plurality of factors. Item 2. The maintenance plan support system for a core welded section according to Item 1. 3. A plurality of factors of the environmental conditions include stresses acting on each baffle former bolt and weld, temperatures to which each baffle former bolt and weld are exposed, and irradiation of each baffle former bolt and weld. 3. The maintenance plan support system for a core welding part according to claim 2, wherein the neutron irradiation amount is included. 4. The apparatus according to claim 1, wherein said influence parameter setting means is configured to prioritize and update the setting of the influence parameter of which variation greatly affects the baffle former bolt damage time among the plurality of influence parameters. The maintenance plan support system for a core welding part according to claim 2 or 3. 5. The baffle former bolt damage time prediction means is configured to predict a baffle former bolt damage time for each baffle former bolt in each reactor based on the influence parameter common to a plurality of reactors. The inspection result data acquisition unit is configured to sequentially acquire inspection result data in a plurality of reactors, and the influence parameter setting unit operates the inspection result data every time the inspection result data is acquired. Based on the measured number of damaged baffle former bolts at time and the predicted number of damaged baffle former bolts on the reactor on which the damage inspection of the inspection result data was performed, the setting of the influence parameter common to a plurality of reactors is performed. The furnace according to any one of claims 1 to 4, wherein the furnace is configured to be renewed. So weld maintenance planning support system. 6. A simulation in an operating state of a nuclear reactor is performed on the assumption that a crack is generated in a core weld,
The method further comprises a soundness evaluation means for evaluating whether or not the crack propagation assumed in the result of the simulation is stopped, and whether or not the functional soundness of the reactor is secured by the remaining welded portion. Means, even if the weld damage time is earlier than the end of the total scheduled operation time of the reactor, the necessity of inspection is required if soundness is confirmed by the soundness evaluation means. The maintenance plan support system for a core welded joint according to any one of claims 1 to 5, wherein the determination is made that there is no gap. 7. A maintenance plan supporting method for supporting a maintenance plan of a core welded portion of a pressurized water reactor, comprising: a baffle former bolt constituting an in-reactor structure; Estimating the baffle former bolt damage time for each baffle former bolt based on environmental conditions assumed from the furnace structural conditions and an influence parameter indicating the degree of influence of the environmental conditions on the damage time of the baffle former bolts; Calculating a predicted number of damaged baffle former bolts with respect to the operating time of the reactor from the baffle former bolt damage time predicted for each baffle former bolt; and The operating time and the baffle damaged during the operating time Acquiring inspection result data including the actual number of damaged baffle former bolts indicating the number of formabolts, and each time the inspection actual data is obtained, the actual number of damaged baffle former bolts during the operation time of the inspection actual data is obtained. Updating the setting of the influence parameter based on the estimated number of damaged baffle former bolts, and welding the welded portion of the core based on the environmental conditions assumed from the structural conditions of the reactor and the influence parameter. A step of predicting a part damage time, and a step of determining whether or not an inspection is necessary based on the weld part damage time and the total scheduled operation time of the reactor. Method. 8. A program for causing a computer to function as a maintenance plan support system for supporting the development of a maintenance plan for a core welded portion of a pressurized water reactor, comprising: a baffle former bolt constituting a reactor internal structure; At each baffle former bolt position, the baffle former bolt damage time is determined for each baffle based on the environmental conditions assumed from the reactor structural conditions and the influence parameter indicating the degree of influence of the environmental conditions on the damage time of the baffle former bolts. Damage baffle former bolt calculating means for predicting the number of damaged baffle former bolts with respect to the operating time of the reactor from the baffle former bolt damage time prediction means for predicting each baffle former bolt damage time and the baffle former bolt damage time predicted for each baffle former bolt A predicted number calculating means, Inspection results to obtain inspection result data including the operation time of the reactor when the damage inspection of the baffle former bolts was performed and the actual number of damaged baffle former bolts indicating the number of damaged baffle former bolts during the operation time Data acquisition means, each time the inspection result data is acquired, updates the setting of the influence parameter based on the measured number of damaged baffle former bolts and the estimated number of damaged baffle former bolts during the operation time of the inspection result data. Influence parameter setting means, a welded part damage time prediction means for predicting a welded part damage time, based on the environmental conditions assumed from the structural conditions of the reactor and the influence parameter, for the welded part of the reactor core, Inspection required to judge the necessity of inspection based on weld damage time and total scheduled operation time of reactor A determination unit, and a program for causing a computer to function with. 9. A computer-readable recording medium on which a program for causing a computer to function as a maintenance plan support system for supporting the development of a maintenance plan for a core welded portion of a pressurized water reactor is recorded. Regarding the baffle former bolts constituting the structure, based on the environmental conditions assumed from the structural conditions of the reactor at the position of each baffle former bolt, and the influence parameters indicating the degree of influence of the environmental conditions on the damage time of the baffle former bolts A baffle former bolt damage time prediction means for predicting the baffle former bolt damage time for each baffle former bolt, and a baffle former bolt damage time predicted for each baffle former bolt. Loss to calculate the estimated number of bolts A baffle former bolt estimated number calculating means, and an operation time of the reactor when the baffle former bolt is inspected for damage and an actual measured number of damaged baffle former bolts indicating the number of damaged baffle former bolts during the operation time. Inspection results data acquisition means for acquiring inspection results data including, every time the inspection results data is obtained, the actual number of damage baffle former bolts and the predicted number of damaged baffle former bolts in the operation time of the inspection results data An influence parameter setting means for updating the setting of the influence parameter based on the environmental parameters assumed from the structural conditions of the reactor and the influence parameter for the weld of the core, based on the influence parameter; Means for predicting the weld damage time, the weld damage time and the total operation of the reactor. And check necessity determining means for determining the necessity of inspection and a scheduled time, and a program for causing a computer to function by recording computer-readable recording medium.