Disposal Methods for Decommissioned Control Rod Assemblies of Pressurized Water Reactors

Abstract

The present invention relates to a method for processing decommissioned control rod assemblies of pressurized water reactors, comprising the following steps: S1. Underwater disassembly and division of the control rod assembly into multiple structural units; each structural unit corresponds to a type of radioactive solid waste based on its radioactive activity; S2. Separately encapsulating the multiple structural units in different shielding containers, each of which is placed in a storage grid within a spent fuel pool; S3. Determining whether the shielding container meets conditions for external transportation; if so, transporting the shielding container to a predetermined location for corresponding treatment based on the type of radioactive solid waste to which the structural units within the shielding container belong; if not, leaving the shielding container in the storage grid. The control rod assembly is disassembled, divided, and classified, and subsequently different external transportation treatment methods are selected based on the type of radioactive solid waste to which each structural unit belongs, thereby providing a scientific and reasonable external transportation treatment method for the control rod assembly stored in the spent fuel pool.

Description

Technical Field

[0001] The present invention relates to the technical field of nuclear power plant reactor internals, and in particular to a method for processing a decommissioned control rod assembly of a pressurized water reactor. Background Art

[0002] In existing pressurized water reactor (PWR) nuclear power plants, control rod assemblies are inserted above the fuel assemblies to control the safe operation of the reactor. As PWR units age, they need to be replaced before decommissioning, leading to a gradual increase in the number of control rod assemblies removed from the core.

[0003] Retired control rod assemblies contain a large amount of radioactive elements and cannot be directly transported for disposal. The existing method for handling retired control rod assemblies is temporary storage: the control rod assemblies are temporarily plugged into spent fuel assemblies in the spent fuel pool and then disposed of together when the reactor is decommissioned. However, the space within the spent fuel pool is limited, especially the limited space for the storage racks used to store spent fuel assemblies. As the number of control rod assemblies increases, temporary storage becomes less suitable. Furthermore, the increasing number of control rod assemblies requires a significant number of spent fuel assembly and control rod assembly plugging operations in the spent fuel pool, increasing the risk of fuel handling accidents.

[0004] In summary, given that retired control rod assemblies contain significant amounts of radioactive elements, and existing technologies lack research on how to safely and effectively transport them, direct disposal is not feasible. Furthermore, as the number of control rod assemblies unloaded from the core increases, temporary storage of retired control rod assemblies in the spent fuel pool will no longer be feasible. Therefore, a new method for handling retired control rod assemblies is needed. Summary of the Invention

[0005] The technical problem to be solved by the present invention is to provide an improved method for processing decommissioned control rod assemblies of pressurized water reactors.

[0006] The technical solution adopted by the present invention to solve the technical problem is to provide a method for processing a decommissioned control rod assembly of a pressurized water reactor, which comprises the following steps:

[0007] S0. Establishing a decommissioned control rod assembly radioactivity analysis model for measuring the radioactivity of decommissioned control rod assemblies;

[0008] S1. Disassemble and divide the control rod assembly into multiple structural units underwater;

[0009] The radioactivity of each of the structural units is calculated using the radioactivity analysis model of the decommissioned control rod assembly; each structural unit corresponds to a type of radioactive solid waste according to its radioactivity;

[0010] S2. Encapsulating the plurality of structural units in different shielded containers, respectively, and placing each shielded container containing the structural units in a storage rack within the spent fuel pool;

[0011] S3. Detecting the radioactive dose equivalent on the outer surface of each shielding container; calculating the radioactive activity of each structural unit using the radioactive activity analysis model of the decommissioned control rod assembly; each structural unit corresponds to a type of radioactive solid waste according to its radioactive activity;

[0012] Determining whether the shielded container in the storage grid in the spent fuel pool meets the conditions for external transportation; the conditions for external transportation are: the radioactive dose equivalent on the outer surface of the shielded container is ≤2.0 mSv / h, and the structural units encapsulated therein belong to a preset type of radioactive solid waste;

[0013] If the conditions for transportation are met, the shielding container will be transported to a preset location for corresponding treatment based on the type of radioactive solid waste to which the structural unit in the shielding container belongs; the preset location includes a low- and medium-level radioactive solid waste disposal site, a radioactive metal waste smelting and processing plant, and an abandoned mine.

[0014] If the transport conditions are not met, the shielding container will be left in the storage grid in the spent fuel pool.

[0015] Preferably, step S0 includes:

[0016] Step S01: using radioactivity detection equipment to detect and record the radioactivity of a plurality of retired control rod assemblies corresponding to different conditional variable parameters, and establishing a database of radioactivity of retired control rod assemblies;

[0017] Step S02: Combining the radioactivity database of the decommissioned control rod assemblies with the radionuclide decay theory, a radioactivity analysis model for the decommissioned control rod assemblies is established.

[0018] Preferably, the conditional variable parameters include the service time in the reactor and the storage time in the spent fuel pool after decommissioning.

[0019] Preferably, the radioactivity detection equipment is a gamma dose rate meter or a nuclide identification spectrometer.

[0020] Preferably, step S1 includes:

[0021] Step S10: pre-dividing the entire control rod assembly into a plurality of structural units according to the metal composition of each part of the control rod assembly;

[0022] Step S11, calculating the radioactivity of each of the structural units according to the radioactivity analysis model of the decommissioned control rod assembly;

[0023] Step S12: when the radioactivity of one of the structural units is calculated to be less than or equal to a preset value, the structural unit is sheared and divided underwater;

[0024] Step S13: Repeat step S12 until the control rod assembly is disassembled and divided into multiple structural units.

[0025] Preferably, the preset value is 5 Sv / h.

[0026] Preferably, in step S3, the preset radioactive solid waste types include intermediate radioactive solid waste, low radioactive solid waste, and very low radioactive solid waste.

[0027] Preferably, in step S3, when a shielding container meets the external transport conditions, the type of radioactive solid waste to which the structural unit in the shielding container belongs is determined:

[0028] When the structural units in the shielding container are intermediate-level radioactive solid waste or low-level radioactive solid waste, the shielding container is transported to the low- and intermediate-level radioactive solid waste disposal site for surface landfill treatment;

[0029] When the structural units in the shielding container are extremely low-level radioactive solid waste, the shielding container is transported to the radioactive metal waste smelting and processing plant for smelting and recovery treatment or transported to the abandoned mine for centralized landfill treatment.

[0030] Preferably, in step S1, the control rod assembly is disassembled and divided into three structural units underwater;

[0031] The three structural units are: absorber rod unit (1), stainless steel rod unit (2), and upper frame unit (3);

[0032] The three structural units correspond to three shielding containers respectively, and the three shielding containers are: a first shielding container, a second shielding container, and a third shielding container.

[0033] Preferably, in step S2, each shielding container is provided corresponding to the type of radioactive solid waste to which the structural unit to be encapsulated therein belongs.

[0034] The present invention has at least the following beneficial effects: by disassembling and segmenting the control rod assembly and classifying it, different transport and disposal methods are selected based on the type of radioactive solid waste to which each structural unit belongs, providing a scientific and rational transport and disposal method for the control rod assemblies stored in the spent fuel pool. The type of radioactive solid waste to which each structural unit belongs is indirectly determined through a radioactivity analysis model for retired control rod assemblies, eliminating numerous underwater measurements and improving safety and convenience. BRIEF DESCRIPTION OF THE DRAWINGS

[0035] The present invention will be further described below with reference to the accompanying drawings and embodiments, in which:

[0036] Figure 1 1 is a system schematic diagram of a method for processing a decommissioned control rod assembly of a pressurized water reactor according to an embodiment of the present invention;

[0037] Figure 2 It is a structural schematic diagram of a control rod assembly corresponding to a method for processing a decommissioned control rod assembly of a pressurized water reactor according to an embodiment of the present invention. DETAILED DESCRIPTION

[0038] In order to have a clearer understanding of the technical features, purposes and effects of the present invention, specific embodiments of the present invention are now described in detail with reference to the accompanying drawings.

[0039] The terms "first," "second," and "third" are used solely to facilitate description of the present technical solution and should not be construed as indicating or implying relative importance or implicitly specifying the quantity of the technical features indicated. In other cases, detailed descriptions of well-known systems, devices, and methods are omitted so as not to obscure the description of the present invention with unnecessary detail.

[0040] Figure 1 This is a schematic diagram of a pressurized water reactor decommissioned control rod assembly processing system corresponding to a method for processing a pressurized water reactor decommissioned control rod assembly according to an embodiment of the present invention. Figure 1 , starting from the control rod assembly and following the arrow direction on the right, the method for processing a decommissioned control rod assembly of a pressurized water reactor according to one embodiment of the present invention comprises the following steps:

[0041] S0. Establish a decommissioned control rod assembly radioactivity analysis model (hereinafter sometimes referred to as the analysis model) for measuring the radioactivity of decommissioned control rod assemblies. The analysis model is actually a data analysis model.

[0042] Further, please refer to Figure 1 , starting from the control rod assembly and following the arrow direction on the left, step S0 includes:

[0043] Step S01: Using radioactivity detection equipment, detect and record the radioactivity of multiple retired control rod assemblies corresponding to different conditional variable parameters, and establish a retired control rod assembly radioactivity database to provide a basis for the subsequent establishment of a retired control rod assembly radioactivity analysis model.

[0044] Radioactivity detection equipment can include a gamma dose rate meter or a radionuclide identification spectrometer. For example, the AMP100 gamma dose rate meter can be used to remotely or underwater measure the radioactivity of radioactive materials. Alternatively, an ORTEC portable radionuclide identification spectrometer can be used. This device is portable and can be used on-site at nuclear power plants to directly and qualitatively measure the radioactivity in a pipe or object.

[0045] In this embodiment, the conditional variable parameters include the service time in the reactor and the storage time in the spent fuel pool after decommissioning. Of course, in other embodiments, the conditional variable parameters can also be other parameters, or other parameters can be included based on this embodiment.

[0046] In other words, each retired control rod assembly has a different in-reactor service time and different storage time in the spent fuel pool after decommissioning, which has a certain degree of impact on its radioactivity. Using existing retired control rod assemblies with different conditional variable parameters as data samples, their radioactivity is measured and recorded using radioactivity detection equipment. By recording a large amount of sample data, a database of retired control rod assembly radioactivity is established. This database records the radioactivity of the control rod assembly corresponding to a specific in-reactor service time, the radioactivity of the control rod assembly corresponding to a specific storage time in the spent fuel pool after decommissioning, and other information.

[0047] Not only can existing retired control rod assemblies be used as data samples, but subsequently generated retired control rod assemblies can also be used as data samples. The retired control rod assembly radioactivity database can continuously update the recorded data to increase its sample data volume. The larger the sample data volume, the more accurate the output results of the analysis model ultimately established will be.

[0048] Step S02: The retired control rod assembly radioactivity database is combined with radionuclide decay theory to establish a retired control rod assembly radioactivity analysis model for measuring the radioactivity of retired control rod assemblies.

[0049] Specifically, according to radionuclide decay theory, different radionuclides have different natural properties, such as half-lives. The metal elements in various components of the control rod assembly include some radionuclides. After a radionuclide's half-life has passed, its radioactivity is likely to decrease significantly. Therefore, natural properties such as the half-life of the radionuclides also have a certain degree of influence on the radioactivity of the control rod assembly and can therefore correspond to variable parameters in an analytical model. Specifically, in this embodiment, the independent variable parameters of the analytical model include conditional variables (in-reactor service time and time stored in the spent fuel pool after decommissioning) and metal composition variables.

[0050] Therefore, the decommissioned control rod assembly activity analysis model effectively reflects the influence of three independent variables (in-process service time, time stored in the spent fuel pool after decommissioning, and metal composition) on the output result (radioactivity). Simply by determining these three independent variables and inputting them into the decommissioned control rod assembly activity analysis model, a corresponding radioactivity value can be obtained. Of course, as mentioned above, the conditional variable parameters can also include other parameters.

[0051] The radioactivity analysis model for retired control rod assemblies is actually a data analysis model. Those skilled in the art can establish a database of radioactivity of retired control rod assemblies and a radioactivity analysis model for retired control rod assemblies based on existing data analysis and data analysis modeling technologies.

[0052] S1. Disassemble the control rod assembly into multiple structural units underwater.

[0053] In existing technology, the radioactivity of control rod assemblies can be determined through underwater measurements. However, this involves operational risks and is limited by on-site factors, making it inconvenient and unsafe. Therefore, the present invention pre-establishes a radioactivity analysis model for retired control rod assemblies to measure the radioactivity of retired control rod assemblies, eliminating the need for many underwater measurements and providing greater safety and convenience.

[0054] In this embodiment, step S1, i.e., disassembling and dividing the control rod assembly into multiple structural units underwater, specifically includes the following steps:

[0055] Step S10: pre-dividing the entire control rod assembly into a plurality of structural units based on the metal composition of each part of the control rod assembly (or the metal composition and the distance between the control rod assembly and the fuel assembly).

[0056] The distance from the fuel assembly is introduced as a basis for judgment because the closer a part of the control rod assembly is to the fuel assembly, the greater its radioactivity; the farther away it is from the fuel assembly, the smaller its radioactivity. This allows us to roughly judge the radioactivity of each part of the control rod assembly and perform artificial pre-classification.

[0057] Furthermore, using metal composition as a conditional variable parameter in the analysis model facilitates the division of the control rod assembly into multiple structural units. Specifically, a portion of the control rod assembly with the same metal composition is considered a structural unit, from which multiple structural units can be divided. These structural units belong to the same control rod assembly and therefore share the same characteristics of in-reactor service life and post-decommissioning storage in the spent fuel pool, but differ in their metal composition. Therefore, the radioactivity analysis model for decommissioned control rod assemblies can be used to determine the radioactivity of each structural unit by inputting the metal composition characteristics of different parts of the control rod assembly into the analysis model. Therefore, the entire control rod assembly can be pre-divided into multiple structural units based on varying radioactivity.

[0058] Step S11: Calculate the radioactivity of each structural unit according to the radioactivity analysis model of the decommissioned control rod assembly.

[0059] Step S12: When it is calculated that the radioactivity of one of the structural units is less than or equal to a preset value, the structural unit is sheared and divided underwater.

[0060] Specifically, shearing and shrinking equipment can be used to shear and split underwater structural units. The preset value can be determined based on the specific conditions of each nuclear power plant. The shearing and splitting operation can only be started when the radioactivity of a structural unit has dropped to a safe preset value, minimizing the risk to equipment and personnel. This enhances the safety of the shearing and splitting operation.

[0061] Furthermore, the preset value may be 5 Sv / h. It should be noted that, based on historical underwater measurements of radioactive activity of discharged burnable poison assemblies, when the measured radioactive activity on the surface of the burnable poison assembly is less than or equal to 5 Sv / h, it is a relatively safe time for underwater operations.

[0062] Step S13: Repeat step S12 until the control rod assembly is disassembled and divided into multiple structural units.

[0063] S2. Encapsulate the multiple structural units in different shielding containers respectively, and place each shielding container containing the structural units in a storage grid in the spent fuel pool.

[0064] Specifically, the shielding container may have functions such as gamma ray shielding, sealing and waterproofing, and drainage, and may be designed with some cooperating parts for lifting to facilitate lifting and transportation.

[0065] S3. Use a dose equivalent rate meter or other testing equipment to measure the radioactive dose equivalent on the outer surface of each shielding container. Calculate the radioactivity of each structural unit using the radioactivity analysis model for decommissioned control rod assemblies. Each structural unit is assigned a type of radioactive solid waste based on its radioactivity.

[0066] According to current national standards, radioactive solid waste types include high-level, intermediate-level, low-level, and very-low-level radioactive solid waste. Each type of radioactive solid waste corresponds to a range of radioactivity. High-level, intermediate-level, low-level, and very-low-level radioactive solid wastes are assigned to a range of radioactivity, in descending order of their radioactivity. Therefore, once the radioactivity of each structural unit is known, it can be assigned to a radioactive solid waste type based on its radioactivity.

[0067] Determine whether the shielded containers in the storage grids within the spent fuel pool meet the conditions for external transport. The conditions for external transport are: the radiation dose equivalent on the outer surface of the shielded container is ≤ 2.0 mSv / h, and the structural units enclosed within it belong to the designated radioactive solid waste type.

[0068] If the conditions for transport are met, the shielding container will be transported to a designated location for disposal based on the type of radioactive solid waste contained within the structural unit. Predetermined locations include low- and intermediate-level radioactive solid waste disposal sites, radioactive metal waste smelting plants, and abandoned mines. Of course, other feasible and legal disposal locations may also be considered. In other words, when considering transporting a shielding container, each container will be classified based on the type of radioactive solid waste contained within.

[0069] This involves disassembling and segmenting the control rod assembly, classifying it, and subsequently selecting different methods for transport and disposal based on the type of radioactive solid waste to which each structural unit belongs. This differentiated treatment approach offers the benefits of targeted, cost-effective treatment, and promising engineering applications.

[0070] The preset radioactive solid waste types may include intermediate-level radioactive solid waste, low-level radioactive solid waste, and very-low-level radioactive solid waste. That is, only when the structural units enclosed in the shielding container are at least intermediate-level radioactive solid waste will they be considered for external disposal. Of course, in other embodiments, the external disposal standards may be raised, and the preset radioactive solid waste types may include only low-level radioactive solid waste and very-low-level radioactive solid waste. Only when the structural units enclosed in the shielding container are at least low-level radioactive solid waste will they be considered for external disposal.

[0071] If the conditions for external transport are not met, the shielded container will be left in the storage racks within the spent fuel pool. After a period of retention, as the radioactive elements in the structural units within the shielded container decay, the radioactivity gradually decreases. The radioactivity on the outer surface of the shielded container also gradually decreases, and the type of radioactive solid waste to which the structural units within it belong will also change. Over time, when the dose equivalent on the outer surface of the retained shielded container is ≤2.0mSv / h and the structural units enclosed within it belong to the predetermined radioactive solid waste type, the conditions for external transport are met and the container will be transported.

[0072] The core idea and corresponding beneficial effects of the method for processing the decommissioned control rod assembly of a pressurized water reactor of the present invention are: (1) the control rod assembly is disassembled, divided and classified, and then different external transportation treatment methods are selected according to the type of radioactive solid waste to which each structural unit belongs. The benefits brought by the classified treatment are strong pertinence and good safety. By adopting corresponding external transportation treatment methods for different types of radioactive solid waste, the final disposal method of each structural unit can be scientifically and reasonably arranged, providing a scientific and reasonable external transportation treatment method for the control rod assembly stored in the spent fuel pool, and essentially improving the safety of the spent fuel assembly and the decommissioned control rod assembly. In addition, the classified treatment can reasonably allocate radioactive solid waste treatment resources, thereby reducing the cost of radioactive solid waste treatment, and safety is guaranteed, which has a prospect for engineering practice application.

[0073] (2) The type of radioactive solid waste to which each structural unit belongs is indirectly calculated through the radioactivity analysis model of the decommissioned control rod assembly, which eliminates the need for many underwater measurement operations and is safer and more convenient.

[0074] Furthermore, in this embodiment, in step S3, when a shielding container meets the conditions for external transportation, the type of radioactive solid waste to which the structural unit in the shielding container belongs is determined:

[0075] If the structural units within the shielding container are intermediate-level or low-level radioactive solid waste, the shielding container will be transported to a low- and intermediate-level solid waste disposal site for surface landfill. Currently, several low- and intermediate-level solid waste disposal sites are licensed for operation in China.

[0076] When the structural units in the shielding container are extremely low-level radioactive solid waste, the shielding container will be transported to a radioactive metal waste smelting and processing plant for smelting and recovery treatment or transported to an abandoned mine for centralized landfill treatment.

[0077] The current mainstream international treatment technology for the smelting and recycling of low-level radioactive metal waste is metal smelting and recycling. China possesses both the technology and the conditions for smelting low-level radioactive waste. When transporting waste to abandoned mines for centralized landfill, it can be disposed of along with other low-pollution metals generated by nuclear power plants.

[0078] Of course, in other embodiments, different processing methods may be adopted according to the specific conditions of different nuclear power plants.

[0079] like Figure 2 As shown, in this embodiment, in step S1, the control rod assembly is disassembled and divided into three structural units underwater according to the metal composition of each part of the control rod assembly and the distance between each part and the fuel assembly.

[0080] Please refer to Figure 2 The control rod assembly of a pressurized water reactor (PWR) consists primarily of an upper frame unit 3 (star-shaped frame) and 24 control rods in the lower unit. The upper frame unit 3 is brazed together from a central tube, wing plates, and cylindrical finger rods. Made of austenitic stainless steel, the central tube is hollow and contains a spring system. The buffer springs are made of Inconel 718. The control rods consist of a cladding tube and a pellet housed within it. The cladding tube is made of AISI316L nitrided stainless steel. The lower unit contains 24 control rods, which are further divided into absorber rods and stainless steel rods. The absorber rods' cladding tubes contain Ag-In-Cd alloy pellets and an austenitic stainless steel compression spring. The stainless steel rods contain stainless steel alloy pellets and an austenitic stainless steel compression spring.

[0081] Therefore, the absorber rods are primarily composed of Ag, In, Cd, and AISI316L nitrided stainless steel. They are located closer to the bottom of the control rod assembly and have a higher radioactivity. Compared to the absorber rods and stainless steel rods, the upper frame unit 3 is located closer to the top of the control rod assembly and has a lower radioactivity. Therefore, the control rod assembly can be pre-divided into three structural units.

[0082] The three structural units are the absorber rod unit 1, the stainless steel rod unit 2, and the upper frame unit 3.

[0083] The three structural units correspond to three shielding containers respectively, and the three shielding containers are: a first shielding container, a second shielding container, and a third shielding container.

[0084] It is understandable that in other embodiments, the control rod assembly may be divided into two structural units, four structural units, five structural units, or a greater number of structural units. When divided into two structural units, the cladding tube may be regarded as one structural unit, and the core block therein may be regarded as another structural unit.

[0085] Furthermore, in this embodiment, each shielding container is configured to correspond to the type of radioactive solid waste to which the structural unit to be encapsulated therein belongs.

[0086] In other words, once the in-process service life of a control rod assembly, the time it has been stored in the spent fuel pool after decommissioning, and the metal composition of each structural unit are determined, the analysis model can be used to separately calculate the radioactive activity of absorber rod unit 1, stainless steel rod unit 2, and upper frame unit 3. Consequently, each absorber rod unit 1, stainless steel rod unit 2, and upper frame unit 3 can be categorized as a corresponding type of radioactive solid waste.

[0087] The three structural units correspond to three shielding containers respectively, and the three shielding containers are: a first shielding container, a second shielding container, and a third shielding container.

[0088] Furthermore, in step S2, each shielding container may be configured to correspond to the type of radioactive solid waste to which the structural unit to be encapsulated therein belongs.

[0089] Specifically, for example, multiple shielding containers are designed for high-level, intermediate-level, low-level, and very-low-level radioactive solid waste types, with each type of radioactive solid waste corresponding to the radioactivity level of the structural unit. Consequently, multiple shielding containers can be designed with varying shielding performance based on the radioactivity level of the structural units to be enclosed. Structural units with higher radioactivity levels can be designed with shielding containers with higher shielding performance; correspondingly, structural units with lower radioactivity levels can be designed with shielding containers with lower performance standards.

[0090] For example, a shielding container with higher shielding performance can be designed in a multi-layer form.

[0091] Please refer to Figure 1 The specific implementation process of this embodiment is provided below:

[0092] Step S01: Detect and record the radioactivity of multiple retired control rod assemblies using radioactivity detection equipment, and establish a database of retired control rod assemblies radioactivity.

[0093] The multiple retired control rod assemblies correspond to in-reactor service times of 5 years, 10 years, and 15 years, etc.; and the multiple retired control rod assemblies correspond to storage times in the spent fuel pool after decommissioning of 3 years, 6 years, and 9 years, etc.

[0094] For example, the database records that the radioactivity of a decommissioned control rod assembly with a service time of 15 years in the reactor was 11 Sv / h when it was first unloaded into the spent fuel pool.

[0095] Step S02: Combining the retired control rod assembly radioactivity database with radionuclide decay theory, that is, introducing a metal element variable parameter, establishes a retired control rod assembly radioactivity analysis model for measuring the radioactivity of retired control rod assemblies.

[0096] For example, consider the primary radionuclides in the control rod assembly (CRA), such as Ag-110m, Cd-110, Cd-114, Sn-119, and Co-60, with corresponding half-lives ranging from 60 days to 5 years. Therefore, the radioactivity of the portion of the CRA containing Ag and Cd metal elements will likely decrease significantly after storage for more than 5 years. The portion of the CRA containing Ag and Cd can be mapped to absorber rod unit 1.

[0097] Step S10: The control rod assembly is pre-divided into absorber rod units 1, stainless steel rod units 2, and upper frame units 3 based on the metal composition of each part of the control rod assembly and the distance between the parts and the fuel assembly.

[0098] Step S11: Using absorber rod unit 1 as an example, the decommissioned control rod assembly radioactivity analysis model is used. The independent variable parameters are: 15 years of in-reactor service, 10 years of post-decommission storage in the spent fuel pool, and Ag, In, Cd, and Co-60 (the main metal components of absorber rod unit 1). The decommissioned control rod assembly radioactivity analysis model outputs a theoretical radioactivity value of 5 Sv / h for absorber rod unit 1 corresponding to these parameters.

[0099] Step S12: The radioactivity of the absorber rod unit 1 is calculated to be 5 Sv / h, which satisfies the condition of being less than or equal to 5 Sv / h. The absorber rod unit 1 is sheared and divided underwater.

[0100] S2. Encapsulate the absorber rod unit 1 in a first shielding container and place the first shielding container in a storage grid within the spent fuel pool. The absorber rod unit 1 has a high radioactivity level, so the first shielding container adopts a double-layer design.

[0101] S3. Actual measurement or calculation using an analytical model indicates that the radiation dose equivalent on the outer surface of the first shielding container is 1.8 mSv / h, meeting the requirement of less than or equal to 2.0 mSv / h. Furthermore, the theoretical radioactivity of the absorber rod unit 1 within the first shielding container is 5 Sv / h, categorizing it as intermediate-level radioactive solid waste and meeting the conditions for external transport. The first shielding container is then transported to a low- and intermediate-level solid waste disposal site for surface landfill.

[0102] Similarly, steps S11 , S12 , S2 , and S3 are repeated for the stainless steel rod unit 2 , the upper frame unit 3 , and their corresponding second and third shielding containers.

[0103] Measurements showed that the dose equivalent on the outer surfaces of the second and third shielding containers was 1.9 mSv / h, meeting the requirement of less than or equal to 2.0 mSv / h. Furthermore, the stainless steel rod unit 2 and upper frame unit 3 contained within them were classified as very low-level radioactive solid waste, thus meeting the requirements for external transport. The second and third shielding containers were then transported to a radioactive metal waste smelting and processing plant for smelting and recovery.

[0104] The steps of establishing a decommissioned control rod assembly radioactivity database or establishing a decommissioned control rod assembly radioactivity analysis model described in conjunction with the embodiments disclosed herein can be directly implemented using computer software. The corresponding computer software can be stored in random access memory (RAM), internal memory, read-only memory (ROM), electrically programmable ROM, electrically erasable programmable ROM, registers, a hard disk, a removable disk, a CD-ROM, or any other storage medium known in the art.

[0105] Professionals may further appreciate that the methods and steps of measurement, modeling, and judgment described in conjunction with the embodiments disclosed herein can be implemented using computer software.

[0106] It is understandable that the above embodiments only express preferred embodiments of the present invention, and their descriptions are relatively specific and detailed, but they should not be understood as limiting the scope of the patent of the present invention. It should be pointed out that for ordinary technicians in this field, without departing from the concept of the present invention, the above technical features can be freely combined, and several variations and improvements can be made, all of which fall within the scope of protection of the present invention. Therefore, all equivalent changes and modifications made to the scope of the claims of the present invention should fall within the scope of coverage of the claims of the present invention.

Claims

1. A method for processing a decommissioned control rod assembly of a pressurized water reactor, characterized in that: The following steps are involved: S0. Establishing a retired control rod assembly radioactivity analysis model for measuring the radioactivity of retired control rod assemblies. Step S0 includes: Step S01. Detecting and recording the radioactivity of multiple retired control rod assemblies corresponding to different conditional variable parameters using radioactivity detection equipment to establish a retired control rod assembly radioactivity database; Step S02. Combining the retired control rod assembly radioactivity database with radionuclide decay theory to establish the retired control rod assembly radioactivity analysis model. The conditional variable parameters include in-reactor service time and storage time in the spent fuel pool after decommissioning. S1. Disassemble the control rod assembly underwater into three structural units; the three structural units are: an absorber rod unit (1), a stainless steel rod unit (2), and an upper frame unit (3); S2. Encapsulating the plurality of structural units in different shielded containers, respectively, and placing each shielded container containing the structural units in a storage rack within the spent fuel pool; S3. Detecting the radioactive dose equivalent on the outer surface of each shielding container; calculating the radioactive activity of each structural unit using the radioactive activity analysis model of the decommissioned control rod assembly; each structural unit corresponds to a type of radioactive solid waste according to its radioactive activity; Determining whether the shielded container in the storage grid in the spent fuel pool meets the conditions for external transportation; the conditions for external transportation are: the radioactive dose equivalent on the outer surface of the shielded container is ≤2.0 mSv / h, and the structural units encapsulated therein belong to a preset type of radioactive solid waste; If the transport conditions are met, the shielding container is transported to a predetermined location for corresponding treatment based on the type of radioactive solid waste to which the structural unit in the shielding container belongs; the predetermined location includes a low- and intermediate-level radioactive solid waste disposal site, a radioactive metal waste smelting and processing plant, and an abandoned mine; If the transport conditions are not met, the shielding container will be left in the storage grid in the spent fuel pool.

2. The method for processing a decommissioned control rod assembly of a pressurized water reactor according to claim 1, characterized in that: The radioactivity detection equipment is a gamma dose rate meter or a nuclide identification spectrometer.

3. The method for processing a decommissioned control rod assembly of a pressurized water reactor according to claim 1, wherein: Step S1 includes: Step S10: pre-dividing the entire control rod assembly into a plurality of structural units according to the metal composition of each part of the control rod assembly; Step S11, calculating the radioactivity of each of the structural units according to the radioactivity analysis model of the decommissioned control rod assembly; Step S12: when the radioactivity of one of the structural units is calculated to be less than or equal to a preset value, the structural unit is sheared and divided underwater; Step S13: Repeat step S12 until the control rod assembly is disassembled and divided into multiple structural units.

4. The method for processing a decommissioned control rod assembly of a pressurized water reactor according to claim 3, characterized in that: The preset value is 5 Sv / h.

5. The method for processing a decommissioned control rod assembly of a pressurized water reactor according to claim 1, wherein: In step S3, the preset radioactive solid waste types include intermediate radioactive solid waste, low radioactive solid waste, and very low radioactive solid waste.

6. The method for processing a decommissioned control rod assembly of a pressurized water reactor according to claim 5, characterized in that: In step S3, when a shielding container meets the transport conditions, the type of radioactive solid waste to which the structural unit in the shielding container belongs is determined: When the structural units in the shielding container are intermediate-level radioactive solid waste or low-level radioactive solid waste, the shielding container is transported to the low- and intermediate-level radioactive solid waste disposal site for surface landfill treatment; When the structural units in the shielding container are extremely low-level radioactive solid waste, the shielding container is transported to the radioactive metal waste smelting and processing plant for smelting and recovery treatment or transported to the abandoned mine for centralized landfill treatment.

7. The method for processing a decommissioned control rod assembly of a pressurized water reactor according to any one of claims 1 to 6, characterized in that: The three structural units correspond to three shielding containers respectively, and the three shielding containers are: a first shielding container, a second shielding container, and a third shielding container.

8. The method for processing a decommissioned control rod assembly of a pressurized water reactor according to any one of claims 1 to 6, characterized in that: In step S2, each shielding container is set corresponding to the type of radioactive solid waste to which the structural unit to be encapsulated therein belongs.

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