Method and system for evaluating service life of reactor pressure vessel

Abstract

The invention discloses a method and a system for evaluating the service life of a reactor pressure vessel. Relates to the technical field of reactor pressure vessel management. The method is improved on the basis of a traditional reactor pressure vessel service life estimation technology, and an irradiation reference sample and an auxiliary irradiation original sample which belong to the same reactor pressure vessel to be evaluated are constructed while an irradiation original sample is constructed; the service life of the reactor pressure vessel to be evaluated is evaluated on the basis of the irradiation reference sample, the irradiation original sample and the auxiliary irradiation original sample, and the service life of the reactor pressure vessel to be evaluated is evaluated from multiple dimensions; the angle between the original sample and the reference sample in the reactor pressure vessel to be evaluated and the angle between the original sample in the reactor pressure vessel to be evaluated and the reference samples in other reactor pressure vessels can improve the accuracy of the service life evaluation result of the reactor pressure vessel.

Description

Technical Field

[0001] The invention relates to the technical field of reactor pressure vessel management, and in particular to a reactor pressure vessel life assessment method and system. Background Art

[0002] Reactor pressure vessels are important large components of pressurized water reactor nuclear power plants. Since low-alloy ferritic steel used in reactor pressure vessels will experience an increase in ductile-brittle transition temperature and a decrease in impact toughness after neutron irradiation, nuclear power plant reactor pressure vessels usually use irradiation monitoring tests to study and evaluate the neutron irradiation embrittlement effect. The irradiation monitoring method is to seal the material samples taken from the reactor pressure vessel in a monitoring tube, and place multiple monitoring tubes between the thermal shield layer in the reactor and the inner wall of the reactor pressure vessel for irradiation with the reactor. By regularly extracting an irradiation monitoring tube, the test is carried out to monitor the changes in the mechanical properties of the reactor pressure vessel material caused by irradiation.

[0003] Life assessment of reactor pressure vessels involves the safety and management of nuclear power plants. Traditional methods of reactor pressure vessel life assessment mainly analyze and estimate material samples taken from reactor pressure vessels in monitoring tubes, which results in large errors. Summary of the invention

[0004] The technical problem to be solved by the present invention is that the traditional reactor pressure vessel life assessment method mainly analyzes and estimates the material samples taken from the reactor pressure vessel in the monitoring tube, which has a large error; the purpose of the present invention is to provide a reactor pressure vessel life assessment method and system, which improves the method on the basis of the traditional reactor pressure vessel life estimation technology, and constructs an irradiated reference sample and an auxiliary irradiated original sample belonging to the same reactor pressure vessel to be assessed while constructing an irradiated original sample, and the life of the reactor pressure vessel to be assessed is assessed based on the irradiated reference sample, the irradiated original sample and the auxiliary irradiated original sample, and the life of the reactor pressure vessel to be assessed is assessed from multiple dimensions: the angle between different original samples in the reactor pressure vessel to be assessed, the angle between the original sample in the reactor pressure vessel to be assessed and the reference sample, the angle between the original sample in the reactor pressure vessel to be assessed and the reference sample in other reactor pressure vessels, so as to improve the accuracy of the reactor pressure vessel life assessment result.

[0005] The present invention is achieved through the following technical solutions:

[0006] This scheme provides a reactor pressure vessel life assessment method, including:

[0007] Put the reference sample and the original sample together in the pressure vessel of the reactor to be evaluated and irradiate with the reactor to construct the irradiated reference sample and the irradiated original sample; put the reference sample in the pressure vessel of the auxiliary reactor and irradiate with the reactor to construct the auxiliary irradiated reference sample;

[0008] Extracting an irradiated reference sample, an irradiated original sample, and an auxiliary irradiated original sample, and performing an impact test on the irradiated reference sample, the irradiated original sample, and the auxiliary irradiated original sample to obtain impact performance data of each sample;

[0009] The life of the reactor pressure vessel to be evaluated is estimated based on the impact performance data of each sample.

[0010] Working principle of this scheme: The traditional reactor pressure vessel life assessment method mainly analyzes and estimates the material samples taken from the reactor pressure vessel in the monitoring tube, which has a large error; the purpose of the present invention is to provide a reactor pressure vessel life assessment method and system, which improves the method on the basis of the traditional reactor pressure vessel life estimation technology, and constructs an irradiated reference sample and an auxiliary irradiated original sample belonging to the same reactor pressure vessel to be assessed while constructing the irradiated original sample, and evaluates the life of the reactor pressure vessel to be assessed based on the irradiated reference sample, the irradiated original sample and the auxiliary irradiated original sample, and evaluates the life of the reactor pressure vessel to be assessed from multiple dimensions: the angle between different original samples in the reactor pressure vessel to be assessed, the angle between the original sample and the reference sample in the reactor pressure vessel to be assessed, and the angle between the original sample in the reactor pressure vessel to be assessed and the reference sample in other reactor pressure vessels, so as to improve the accuracy of the reactor pressure vessel life assessment result.

[0011] A further optimization scheme is that the original sample is sampled from a forging of a reactor pressure vessel A, and the sampling direction of the original sample is perpendicular or parallel to the main forging direction of the forging; the reactor pressure vessel A is the reactor pressure vessel to be evaluated;

[0012] The reference sample and the original sample are sampled from the forging of the reactor pressure vessel B; the sampling direction of the reference sample is consistent with that of the original sample; the reactor pressure vessel B and the reactor pressure vessel A are different nuclear reactors.

[0013] A further optimization scheme is to place the reference sample and the original sample together in the reactor pressure vessel to be evaluated before irradiation, and also includes:

[0014] The reference sample and the original sample are subjected to impact tests to obtain benchmark impact performance data.

[0015] A further optimization scheme is that the method for obtaining the impact performance data / benchmark impact performance data includes:

[0016] Impact tests were performed on each sample at different temperatures;

[0017] Obtain the impact absorption energy value, lateral expansion and brittle section percentage of each sample;

[0018] The ductile-brittle transition temperature curve of each sample was fitted;

[0019] The upper platform energy and characteristic temperature value of each sample are determined based on the ductile-brittle transition temperature curve.

[0020] A further optimization scheme is that the fitting method of the ductile-brittle transition temperature curve includes:

[0021] The impact absorption energy value, lateral expansion and brittle section percentage of each sample were input into the hyperbolic tangent function empirical equation to fit the ductile-brittle transition temperature curve of each sample;

[0022] The hyperbolic tangent function empirical equation is:

[0023] E=A1+(A2-A1)[1+tanh(T-T0) / C] / 2;

[0024] Among them, E represents the impact absorption energy, A1 represents the lower platform energy, A2 represents the upper platform energy, T represents the test temperature, T0 represents the ductile-brittle transition temperature, and C represents the half-width of the transition interval.

[0025] A further optimization scheme is that the characteristic temperature values ​​include: the marked temperature T corresponding to the absorption energy of 41J 41J , the absorption energy is 56J corresponding to the marked temperature T 56J , the marked temperature T corresponding to the lateral expansion of 0.89 mm 0.89mm The temperature corresponding to the brittle fracture percentage of 50% is FATT50.

[0026] A further optimization scheme is that the life of the reactor pressure vessel to be evaluated based on the impact performance data is evaluated, including the following method:

[0027] First, a radiation embrittlement assessment is performed based on the baseline impact performance data and the impact performance data of each sample to obtain a first radiation embrittlement assessment result;

[0028] Then, the first radiation embrittlement assessment result of the reference sample placed in the reactor pressure vessel to be assessed is compared horizontally with the first radiation embrittlement assessment result of the original sample to obtain a second radiation embrittlement assessment result;

[0029] Finally, the first radiation embrittlement assessment result of the reference sample placed in the reactor pressure vessel to be assessed is compared horizontally with the first radiation embrittlement assessment result of the reference sample placed in the auxiliary reactor pressure vessel to obtain a third radiation embrittlement assessment result;

[0030] The life of the reactor pressure vessel to be evaluated is obtained according to the first radiation embrittlement evaluation result, the second radiation embrittlement evaluation result and the third radiation embrittlement evaluation result.

[0031] A further optimization scheme is that the first radiation embrittlement evaluation result is obtained by evaluating the baseline impact performance data and the impact performance data of each sample, including the method:

[0032] The method of obtaining a first radiation embrittlement evaluation result by evaluating the baseline impact performance data and the impact performance data of each sample includes:

[0033] The ductile-brittle transition temperature of each sample was calculated according to the following formula:

[0034] RT NDT =RT NDTi +△T cv

[0035] Among them, RT NDT RT is the ductile-brittle transition temperature of the material after irradiation; NDTi is the initial reference transformation temperature; obtained according to the initial drop hammer result when the material is manufactured, △T cv Indicates the change in ductile-brittle transition temperature, which is determined by T before and after irradiation. 56J and T 0.89mm The change value is large.

[0036] A further optimization scheme is that the method of horizontal comparison includes: integrating reference sample data from different nuclear power plants, comparing the change values ​​of the ductile-brittle transition temperature of the reference samples under different neutron injections and different operating conditions, and analyzing the degree of radiation embrittlement.

[0037] The present invention also provides a reactor pressure vessel life assessment system, which is used to implement the above-mentioned reactor pressure vessel life assessment method, and the system includes:

[0038] The sample construction module is used to place the reference sample and the original sample together in the pressure vessel of the reactor to be evaluated for irradiation with the reactor, so as to construct the irradiated reference sample and the irradiated original sample; and to place the reference sample in the pressure vessel of the auxiliary reactor for irradiation with the reactor, so as to construct the auxiliary irradiation reference sample;

[0039] A test module, used for extracting an irradiated reference sample, an irradiated original sample and an auxiliary irradiated original sample, and performing an impact test on the irradiated reference sample, the irradiated original sample and the auxiliary irradiated original sample to obtain impact performance data of each sample;

[0040] The evaluation module is used to evaluate the life of the reactor pressure vessel to be evaluated based on the impact performance data of each sample.

[0041] Compared with the prior art, the present invention has the following advantages and beneficial effects:

[0042] The present invention provides a reactor pressure vessel life assessment method and system, which improves the method on the basis of the traditional reactor pressure vessel life estimation technology. While constructing the irradiated original sample, an irradiated reference sample and an auxiliary irradiated original sample belonging to the same reactor pressure vessel to be assessed are constructed, and the life of the reactor pressure vessel to be assessed is assessed based on the irradiated reference sample, the irradiated original sample and the auxiliary irradiated original sample. The life of the reactor pressure vessel is assessed from multiple dimensions: the angle between different original samples in the reactor pressure vessel to be assessed, the angle between the original sample and the reference sample in the reactor pressure vessel to be assessed, and the angle between the original sample in the reactor pressure vessel to be assessed and the reference sample in other reactor pressure vessels, so as to improve the accuracy of the reactor pressure vessel life assessment result. BRIEF DESCRIPTION OF THE DRAWINGS

[0043] In order to more clearly illustrate the technical solutions of the exemplary embodiments of the present invention, the following briefly introduces the drawings required for use in the embodiments. It should be understood that the following drawings only illustrate certain embodiments of the present invention and should not be regarded as limiting the scope. For those of ordinary skill in the art, other relevant drawings can be obtained based on these drawings without creative work. In the drawings:

[0044] Figure 1 This is a flow chart of the reactor pressure vessel life assessment method;

[0045] Figure 2 is the structural schematic diagram of the reference sample;

[0046] Figure 3 This is a schematic diagram of the structure of the irradiation monitoring tube;

[0047] Figure 4 It is a schematic diagram of the energy-temperature transformation curve;

[0048] Figure 5 is the lateral expansion-temperature transition curve;

[0049] Figure 6 is the brittle fracture percentage-temperature transformation curve;

[0050] Figure 7 This is a schematic diagram of the reactor pressure vessel life assessment system structure. DETAILED DESCRIPTION

[0051] In order to make the objectives, technical solutions and advantages of the present invention more clearly understood, the present invention is further described in detail below in conjunction with embodiments and drawings. The exemplary embodiments of the present invention and their description are only used to explain the present invention and are not intended to limit the present invention.

[0052] Life assessment of reactor pressure vessels is related to the safety and management of nuclear power plants. Traditional methods for assessing the life of reactor pressure vessels mainly analyze and estimate material samples taken from the reactor pressure vessels in the monitoring tube, which results in large errors. In view of this, the present invention provides the following embodiments to solve the above technical problems:

[0053] Example 1

[0054] This embodiment provides a method for evaluating the life of a reactor pressure vessel. Figure 1 As shown, including:

[0055] Step 1: Place the reference sample and the original sample together in the reactor pressure vessel to be evaluated and irradiate with the reactor to construct an irradiated reference sample and an irradiated original sample; Place the reference sample in the auxiliary reactor pressure vessel and irradiate with the reactor to construct an auxiliary irradiated reference sample;

[0056] The original sample is sampled from a forging of a reactor pressure vessel A, and the sampling direction of the original sample is perpendicular or parallel to the main forging direction of the forging; the reactor pressure vessel A is the reactor pressure vessel to be evaluated;

[0057] The reference sample and the original sample are sampled from the forging of the reactor pressure vessel B; the sampling direction of the reference sample is consistent with that of the original sample; the reactor pressure vessel B and the reactor pressure vessel A are different nuclear reactors.

[0058] In the specific implementation process, this embodiment uses the reactor pressure vessel A508-3 steel forging as the reference material, and samples are processed at the inner and outer 1 / 4 thickness of the reference material to form a batch of Charpy impact specimens as reference specimens; the sampling position and direction of the reference specimens are consistent with the original impact specimens in the monitoring tube. The structural schematic diagram of the reference specimen is shown in FIG. Figure 2 As shown, the irradiation monitoring tube is as follows Figure 3 shown.

[0059] The reference sample and the original sample are respectively placed in irradiation monitoring tubes, and are placed in the pressure vessel of the reactor to be evaluated for irradiation with the reactor, so as to construct an irradiation reference sample and an irradiation original sample; the reference sample is respectively placed in irradiation monitoring tubes, and is placed in the auxiliary reactor pressure vessel for irradiation with the reactor, so as to construct an auxiliary irradiation reference sample;

[0060] Before placing the reference sample and the original sample together in the reactor pressure vessel to be evaluated for irradiation, the following steps are also included:

[0061] An impact test is performed on the reference sample and the original sample to obtain benchmark impact performance data; the specific acquisition method includes:

[0062] Impact tests were performed on reference specimens / original specimens at different temperatures;

[0063] Obtain the impact absorption energy value, lateral expansion and brittle fracture percentage of the reference sample / original sample;

[0064] Fit the ductile-brittle transition temperature curve of the reference sample / original sample; the fitting method of the ductile-brittle transition temperature curve includes:

[0065] The impact absorption energy value, lateral expansion and brittle section percentage of the reference sample / original sample are input into the hyperbolic tangent function empirical equation to fit the ductile-brittle transition temperature curve of each sample;

[0066] The hyperbolic tangent function empirical equation is:

[0067] E=A1+(A2-A1)[1+tanh(T-T0) / C] / 2;

[0068] Wherein, E represents the impact absorption energy, A1 represents the lower platform energy, A2 represents the upper platform energy, T represents the test temperature, T0 represents the ductile-brittle transition temperature, C represents the half width of the transition interval; tanh represents the hyperbolic tangent function.

[0069] The energy-temperature transition curve, lateral expansion-temperature transition curve and brittle fracture percentage-temperature transition curve of this embodiment are shown in FIG. Figure 4-Figure 6 As shown. Based on the ductile-brittle transition temperature curve, the upper platform energy and characteristic temperature value of each sample are determined. The characteristic temperature value includes: the marking temperature T corresponding to the absorption energy of 41J 41J , the absorption energy is 56J corresponding to the marked temperature T 56J , the marked temperature T corresponding to the lateral expansion of 0.89 mm 0.89mm The temperature corresponding to the brittle fracture percentage of 50% is FATT50.

[0070] Conduct pendulum impact tests on a group (usually 15 to 18) of unirradiated reference samples to establish material benchmark performance data, including: impact absorption energy, lateral expansion, brittle section percentage, ductile-brittle transition curve, upper platform energy, characteristic temperature value, etc. at a series of temperatures;

[0071] Step 2, extracting an irradiated reference sample, an irradiated original sample and an auxiliary irradiated original sample, and performing an impact test on the irradiated reference sample, the irradiated original sample and the auxiliary irradiated original sample to obtain impact performance data of each sample;

[0072] Impact tests are performed on irradiated reference specimens / irradiated original specimens at different temperatures;

[0073] Obtain the impact absorption energy value, lateral expansion and brittle fracture percentage of the irradiated reference sample / irradiated original sample;

[0074] Fitting the ductile-brittle transition temperature curve of the irradiated reference sample / irradiated original sample; the fitting method of the ductile-brittle transition temperature curve includes:

[0075] The impact absorption energy value, lateral expansion and brittle section percentage of the irradiated reference sample / irradiated original sample are respectively input into the hyperbolic tangent function empirical equation to fit the ductile-brittle transition temperature curve of each sample;

[0076] Step 3: Evaluate the life of the reactor pressure vessel to be evaluated based on the impact performance data of each sample; specifically, the method includes:

[0077] First, a radiation embrittlement assessment is performed based on the baseline impact performance data and the impact performance data of each sample to obtain a first radiation embrittlement assessment result; the specific method includes:

[0078] The ductile-brittle transition temperature of each sample was calculated according to the following formula:

[0079] RT NDT =RT NDTi +△T cv

[0080] Among them, RT NDT RT is the ductile-brittle transition temperature of the material after irradiation; NDTi is the initial reference transformation temperature; obtained according to the initial drop hammer result when the material is manufactured; △T cv Indicates the change in ductile-brittle transition temperature, which is T before and after irradiation 56J and T 0.89mm The one with the larger change value.

[0081] Then, the first radiation embrittlement assessment result of the reference sample placed in the reactor pressure vessel to be assessed is compared horizontally with the first radiation embrittlement assessment result of the original sample to obtain a second radiation embrittlement assessment result;

[0082] Finally, the first radiation embrittlement assessment result of the reference sample placed in the reactor pressure vessel to be assessed is compared horizontally with the first radiation embrittlement assessment result of the reference sample placed in the auxiliary reactor pressure vessel to obtain a third radiation embrittlement assessment result;

[0083] The method for horizontal comparison includes: integrating reference sample data of different nuclear power plants, comparing the ductile-brittle transition temperature change values ​​of the reference samples under different neutron injections and different operating conditions, and analyzing the degree of radiation embrittlement.

[0084] The life of the reactor pressure vessel to be evaluated is obtained according to the first radiation embrittlement evaluation result, the second radiation embrittlement evaluation result and the third radiation embrittlement evaluation result.

[0085] Specifically, the first radiation embrittlement assessment result, the second radiation embrittlement assessment result and the third radiation embrittlement assessment result are used as important influencing factors for evaluating the life of the reactor pressure vessel. Weights are set for each radiation embrittlement assessment result to calculate the life of the reactor pressure vessel, wherein the weight of the first radiation embrittlement assessment result is greater than that of the second radiation embrittlement assessment result, and the weight of the second radiation embrittlement assessment result is greater than that of the third radiation embrittlement assessment result.

[0086] By comparing the reference sample data of different monitoring tubes and different nuclear power plants, a horizontal comparison is conducted to complete the irradiation effect analysis of the material under different neutron injections and different operating conditions.

[0087] By combining the data of different monitoring tubes of the same nuclear power plant and reference samples of different nuclear power plants, a horizontal comparison is made to obtain the irradiation embrittlement of the reference material under different neutron injections and different operating conditions. When the data of a certain pressure vessel supervision test is abnormal, the reliability of different batches of irradiation supervision tests of the nuclear power plant can be verified through the results of the reference sample, and the problem of difficulty in comparing irradiation supervision data of different nuclear power plants is solved.

[0088] In this embodiment, the reference sample to be irradiated is encapsulated in a monitoring tube together with other samples in the monitoring tube, and is placed in a nuclear power plant reactor pressure vessel for irradiation with the stack. After each monitoring tube is extracted from the stack, the pendulum impact test of the reference sample before and after irradiation is completed. The impact performance data of the reference sample before and after irradiation is compared to complete the irradiation embrittlement evaluation. The reference sample data of different nuclear power plants are compared horizontally to complete the analysis of the irradiation effect of the material under different neutron injections and different operating conditions. The problem that the existing reactor pressure vessel irradiation monitoring test data is single, the reliability is difficult to verify, and the comparison between different nuclear power plants cannot be solved.

[0089] Since the pressure vessel materials used in different nuclear power plants are not from the same batch, the manufacturers are different, and the initial performance is different, which has a great impact on the pressure vessel life assessment process, in order to verify the effectiveness of the irradiation supervision test of each unit in different nuclear power plants, a reference material with the same pressure vessel material can be selected. This reference material is made into a Charpy impact specimen (called a reference specimen), which is packaged in the irradiation supervision tube of the nuclear power plant together with other original specimens to receive irradiation, and is tested in the same way as other supervision specimens; by comparing the test results of the reference specimens, the reliability of the irradiation supervision tests of different batches of the nuclear power plant can be tested; in addition, the results of the reference specimens of the same material between different nuclear power plants can be compared for horizontal comparative analysis. This implementation makes methodological improvements based on the traditional reactor pressure vessel life estimation technology. While constructing the irradiated original sample, an irradiated reference sample and an auxiliary irradiated original sample belonging to the same reactor pressure vessel to be evaluated are constructed, and the life of the reactor pressure vessel to be evaluated is evaluated based on the irradiated reference sample, the irradiated original sample and the auxiliary irradiated original sample. The life of the reactor pressure vessel to be evaluated is evaluated from multiple dimensions: the angle between different original samples in the reactor pressure vessel to be evaluated, the angle between the original sample and the reference sample in the reactor pressure vessel to be evaluated, and the angle between the original sample in the reactor pressure vessel to be evaluated and the reference samples in other reactor pressure vessels, so as to improve the accuracy of the reactor pressure vessel life evaluation results.

[0090] Example 2

[0091] This embodiment provides a reactor pressure vessel life assessment system, which is used to implement a reactor pressure vessel life assessment method described in Example 1. Figure 7 As shown, the system comprises:

[0092] The sample construction module is used to place the reference sample and the original sample together in the pressure vessel of the reactor to be evaluated for irradiation with the reactor, so as to construct the irradiated reference sample and the irradiated original sample; and to place the reference sample in the pressure vessel of the auxiliary reactor for irradiation with the reactor, so as to construct the auxiliary irradiation reference sample;

[0093] The sample construction module includes relevant equipment for constructing irradiation reference samples, auxiliary irradiation reference samples and irradiation original samples, including equipment for preparing reference samples and original samples, as well as irradiation supervision tubes for placing reference samples and original samples in the reactor pressure vessel to be evaluated or the auxiliary reactor pressure vessel. The samples to be irradiated are respectively encapsulated in irradiation supervision tubes and placed in the reactor pressure vessel of the nuclear power plant. The irradiation supervision tubes are placed in the retaining frame outside the thermal shield of the reactor pressure vessel before the operation of the nuclear power plant and are irradiated together with the operation of the nuclear power plant.

[0094] A test module, used for extracting an irradiated reference sample, an irradiated original sample and an auxiliary irradiated original sample, and performing an impact test on the irradiated reference sample, the irradiated original sample and the auxiliary irradiated original sample to obtain impact performance data of each sample;

[0095] After each irradiation monitoring tube is extracted from the stack, it is dismantled and cut in the hot chamber, and the sample is taken out to complete the pendulum impact test of the reference sample after irradiation at a series of temperatures.

[0096] The evaluation module is used to evaluate the life of the reactor pressure vessel to be evaluated based on the impact performance data of each sample. The change values ​​before and after irradiation are compared to complete the irradiation embrittlement evaluation. The reference sample data of different monitoring tubes in the same nuclear power plant and different nuclear power plants are integrated to make a horizontal comparison to obtain the irradiation embrittlement of the reference material under different neutron injections and different operating conditions. Finally, the life of the reactor pressure vessel to be evaluated is obtained based on the irradiation embrittlement situation.

[0097] Example 3

[0098] This embodiment provides a computer-readable medium having a computer program stored thereon. The computer program is executed by a processor to implement a reactor pressure vessel life assessment method as described in Embodiment 1, specifically performing the following steps:

[0099] T1, obtain the impact performance data of the irradiated reference sample, the irradiated original sample and the auxiliary irradiated original sample after the impact test;

[0100] T2, evaluating the life of the reactor pressure vessel to be evaluated based on the impact performance data.

[0101] Specifically, impact tests are performed on irradiated reference specimens / irradiated original specimens at different temperatures;

[0102] Obtain the impact absorption energy value, lateral expansion and brittle fracture percentage of the irradiated reference sample / irradiated original sample;

[0103] Fitting the ductile-brittle transition temperature curve of the irradiated reference sample / irradiated original sample; the fitting method of the ductile-brittle transition temperature curve includes:

[0104] The impact absorption energy value, lateral expansion and brittle section percentage of the irradiated reference sample / irradiated original sample are respectively input into the hyperbolic tangent function empirical equation to fit the ductile-brittle transition temperature curve of each sample;

[0105] Calculate the ductile-brittle transition temperature RT of each sample NDT ;

[0106] Then, the first radiation embrittlement assessment result of the reference sample placed in the reactor pressure vessel to be assessed is compared horizontally with the first radiation embrittlement assessment result of the original sample to obtain a second radiation embrittlement assessment result;

[0107] Finally, the first radiation embrittlement assessment result of the reference sample placed in the reactor pressure vessel to be assessed is compared horizontally with the first radiation embrittlement assessment result of the reference sample placed in the auxiliary reactor pressure vessel to obtain a third radiation embrittlement assessment result;

[0108] Specifically, the first radiation embrittlement assessment result, the second radiation embrittlement assessment result and the third radiation embrittlement assessment result are used as important influencing factors for evaluating the life of the reactor pressure vessel. Weights are set for each radiation embrittlement assessment result to calculate the life of the reactor pressure vessel, wherein the weight of the first radiation embrittlement assessment result is greater than that of the second radiation embrittlement assessment result, and the weight of the second radiation embrittlement assessment result is greater than that of the third radiation embrittlement assessment result.

[0109] This scheme can carry out impact tests on reference samples before and after irradiation by loading an additional set of reference samples without changing the original supervision test method, so as to achieve the purpose of comparative analysis of supervision data between different batches of supervision tubes and different nuclear power plants.

[0110] Those skilled in the art will appreciate that the embodiments of the present application may be provided as methods, systems, or computer program products. Therefore, the present application may adopt the form of a complete hardware embodiment, a complete software embodiment, or an embodiment in combination with software and hardware. Moreover, the present application may adopt the form of a computer program product implemented in one or more computer-usable storage media (including but not limited to disk storage, CD-ROM, optical storage, etc.) that include computer-usable program code.

[0111] The present application is described with reference to the flowcharts and / or block diagrams of the methods, devices (systems), and computer program products according to the embodiments of the present application. It should be understood that each process and / or box in the flowchart and / or block diagram, as well as the combination of the processes and / or boxes in the flowchart and / or block diagram, can be implemented by computer program instructions. These computer program instructions can be provided to a processor of a general-purpose computer, a special-purpose computer, an embedded processor, or other programmable data processing device to generate a machine, so that the instructions executed by the processor of the computer or other programmable data processing device generate instructions for implementing the processes in the flowchart and / or block diagram. Figure 1 A process or multiple processes and / or boxes Figure 1 A device that provides the functions specified in a block or multiple blocks.

[0112] These computer program instructions may also be stored in a computer-readable memory capable of directing a computer or other programmable data processing device to operate in a specific manner, so that the instructions stored in the computer-readable memory produce an article of manufacture comprising an instruction device, which implements the process Figure 1 A process or multiple processes and / or boxes Figure 1 A function specified in one or more boxes.

[0113] The specific implementation methods described above further illustrate the objectives, technical solutions and beneficial effects of the present invention in detail. It should be understood that the above description is only a specific implementation method of the present invention and is not intended to limit the scope of protection of the present invention. Any modifications, equivalent substitutions, improvements, etc. made within the spirit and principles of the present invention should be included in the scope of protection of the present invention.

Claims

1. A method for assessing the life of a reactor pressure vessel, characterized in that: include: Put the reference sample and the original sample together in the pressure vessel of the reactor to be evaluated and irradiate with the reactor to construct the irradiated reference sample and the irradiated original sample; put the reference sample in the pressure vessel of the auxiliary reactor and irradiate with the reactor to construct the auxiliary irradiated reference sample; Extracting an irradiated reference sample, an irradiated original sample, and an auxiliary irradiated original sample, and performing an impact test on the irradiated reference sample, the irradiated original sample, and the auxiliary irradiated original sample to obtain impact performance data of each sample; The life of the reactor pressure vessel to be evaluated is estimated based on the impact performance data of each sample.

2. A reactor pressure vessel life assessment method according to claim 1, characterized in that: The original sample is sampled from a forging of a reactor pressure vessel A, and the sampling direction of the original sample is perpendicular or parallel to the main forging direction of the forging; the reactor pressure vessel A is the reactor pressure vessel to be evaluated; The reference sample and the original sample are sampled from the forging of the reactor pressure vessel B; the sampling direction of the reference sample is consistent with that of the original sample; the reactor pressure vessel B and the reactor pressure vessel A are different nuclear reactors.

3. A reactor pressure vessel life assessment method according to claim 1, characterized in that: Before placing the reference sample and the original sample together in the reactor pressure vessel to be evaluated for irradiation, the following steps are also included: The reference sample and the original sample are subjected to impact tests to obtain benchmark impact performance data.

4. A reactor pressure vessel life assessment method according to claim 3, characterized in that: The method for obtaining the impact performance data / benchmark impact performance data comprises: Impact tests were performed on each sample at different temperatures; Obtain the impact absorption energy value, lateral expansion and brittle section percentage of each sample; The ductile-brittle transition temperature curve of each sample was fitted; The upper platform energy and characteristic temperature value of each sample are determined based on the ductile-brittle transition temperature curve.

5. A reactor pressure vessel life assessment method according to claim 4, characterized in that: The fitting method of the ductile-brittle transition temperature curve comprises: The impact absorption energy value, lateral expansion and brittle section percentage of each sample were input into the hyperbolic tangent function empirical equation to fit the ductile-brittle transition temperature curve of each sample; The hyperbolic tangent function empirical equation is: E=A1+(A2-A1)[1+tanh(T-T0) / C] / 2; Wherein, E represents the impact absorption energy, A1 represents the lower platform energy, A2 represents the upper platform energy, T represents the test temperature, T0 represents the ductile-brittle transition temperature, C represents the half width of the transition interval; tanh represents the hyperbolic tangent function.

6. A reactor pressure vessel life assessment method according to claim 4, characterized in that: The characteristic temperature values ​​include: the marked temperature T corresponding to the absorption energy of 41J 41J , the absorption energy is 56J corresponding to the marked temperature T 56J , the marked temperature T corresponding to the lateral expansion of 0.89 mm 0.89mm The temperature corresponding to the brittle fracture percentage of 50% is FATT50.

7. A reactor pressure vessel life assessment method according to claim 6, characterized in that: The method of evaluating the life of a reactor pressure vessel to be evaluated based on the impact performance data includes: First, a radiation embrittlement assessment is performed based on the baseline impact performance data and the impact performance data of each sample to obtain a first radiation embrittlement assessment result; Then, the first radiation embrittlement assessment result of the reference sample placed in the reactor pressure vessel to be assessed is compared horizontally with the first radiation embrittlement assessment result of the original sample to obtain a second radiation embrittlement assessment result; Finally, the first radiation embrittlement assessment result of the reference sample placed in the reactor pressure vessel to be assessed is compared horizontally with the first radiation embrittlement assessment result of the reference sample placed in the auxiliary reactor pressure vessel to obtain a third radiation embrittlement assessment result; The life of the reactor pressure vessel to be evaluated is obtained according to the first radiation embrittlement evaluation result, the second radiation embrittlement evaluation result and the third radiation embrittlement evaluation result.

8. A reactor pressure vessel life assessment method according to claim 7, characterized in that: The method of obtaining a first radiation embrittlement evaluation result by evaluating the baseline impact performance data and the impact performance data of each sample includes: The ductile-brittle transition temperature of each sample was calculated according to the following formula: RT NDT =RT NDTi +△T cv Among them, RT NDT RT is the ductile-brittle transition temperature of the material after irradiation; NDTi is the initial reference transition temperature; △T cv Indicates the change value of ductile-brittle transition temperature.

9. A reactor pressure vessel life assessment method according to claim 1, characterized in that: The method for horizontal comparison includes: integrating reference sample data of different nuclear power plants, comparing the ductile-brittle transition temperature change values ​​of the reference samples under different neutron injections and different operating conditions, and analyzing the degree of radiation embrittlement.

10. A reactor pressure vessel life assessment system, characterized in that: A method for evaluating the life of a reactor pressure vessel according to any one of claims 1 to 9, the system comprising: The sample construction module is used to place the reference sample and the original sample together in the pressure vessel of the reactor to be evaluated for irradiation with the reactor, so as to construct the irradiated reference sample and the irradiated original sample; and to place the reference sample in the pressure vessel of the auxiliary reactor for irradiation with the reactor, so as to construct the auxiliary irradiation reference sample; A test module, used for extracting an irradiated reference sample, an irradiated original sample and an auxiliary irradiated original sample, and performing an impact test on the irradiated reference sample, the irradiated original sample and the auxiliary irradiated original sample to obtain impact performance data of each sample; The evaluation module is used to evaluate the life of the reactor pressure vessel to be evaluated based on the impact performance data of each sample.

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