CPR1000 unit refueling method, device and spent fuel assembly based on spent fuel assembly

Abstract

The present application relates to the technical field of nuclear power fuel management and loading, and discloses a CPR1000 unit refueling method, device and spent fuel assembly based on spent fuel assembly, by simulating the neutron transport process around the core according to the preset threshold value of the counting rate of the power detector and the preset simulation model to output the simulation result, determining the burnup depth boundary value required for the spent fuel assembly corresponding to the preset threshold value of the counting rate to be set at the preset core position according to the simulation result, refueling the unit according to the burnup depth boundary value so that the power detection result of the unit is within the preset power detection range, realizing the effective replacement of the secondary neutron source assembly with the spent fuel assembly, improving the unit refueling efficiency, and ensuring the smooth operation of the nuclear power plant unit.

Description

Technical Field

[0001] The present invention relates to the technical field of nuclear power fuel management and loading, and in particular to a CPR1000 unit fuel replacement method and device based on spent fuel assemblies, and a spent fuel assembly. Background Art

[0002] With the development of nuclear power technology, more and more problems related to nuclear power plant safety have emerged. The mainstream technology for non-first cycle fuel loading and reactor startup of pressurized water reactor nuclear power units at home and abroad adopts the secondary neutron source assembly arranged in the core. For example, Ling'ao Nuclear Power Plant Units 3 and 4 (CPR1000 core) have been using the secondary neutron source arranged in the core for fuel loading and reactor startup since commercial operation.

[0003] The activated secondary neutron source loaded into the core ensures that the core can always be monitored by the neutron detector when there is a nuclear fuel assembly. However, the secondary neutron source assembly is affected by the high temperature, high pressure, and high radiation environment of the core during its service life in the reactor. The cladding of the secondary neutron source rod is prone to radiation swelling, breakage, and other accidents in the core. Summary of the invention

[0004] Based on this, it is necessary to provide a CPR1000 unit refueling method, device and spent fuel assembly based on spent fuel assembly to address the above technical problems.

[0005] A CPR1000 unit refueling method based on spent fuel assemblies is applied to a core system, wherein the core system is provided with a power detector; the refueling method comprises:

[0006] Simulating the neutron transport process around the core according to a preset threshold value of the count rate of the power detector and a preset simulation model to output a simulation result;

[0007] Determine, according to the simulation result, a burnup depth boundary value required for setting the spent fuel assembly corresponding to the preset count rate threshold at a preset core position;

[0008] The unit is subjected to a fuel replacement process according to the burnup depth boundary value so that the power detection result of the unit is within a preset power detection range.

[0009] In one embodiment, determining, according to the simulation result, a burnup depth boundary value required for setting the spent fuel assembly corresponding to the preset count rate threshold at a preset core position includes:

[0010] Obtaining the enrichment of the spent fuel assembly and refueling cycle information of the unit;

[0011] A burnup depth boundary value required for setting the spent fuel assembly at a preset core position corresponding to the preset counting rate threshold is obtained according to the simulation result, the enrichment, and the refueling cycle information.

[0012] In one embodiment, the refueling cycle information includes that the unloading process and the loading process of the unit are in the same cycle period, and the unloading process and the loading process of the unit are respectively in two adjacent cycles; the burnup depth boundary value required for setting the spent fuel assembly corresponding to the preset threshold value of the counting rate at the preset core position according to the simulation result, the enrichment, and the refueling cycle information includes:

[0013] When the enrichment of the spent fuel assembly is a first preset enrichment and the unloading process and the loading process of the unit are in the same cycle, obtaining, according to the simulation result, a first burnup depth boundary value required for setting the spent fuel assembly at a preset core position corresponding to the preset counting rate threshold;

[0014] When the enrichment of the spent fuel assembly is a first preset enrichment and the unloading process and the loading process of the unit are respectively in two adjacent cycles, a second burnup depth boundary value required for setting the spent fuel assembly at a preset core position corresponding to the preset counting rate threshold is obtained according to the simulation result.

[0015] In one embodiment, the refueling cycle information includes that the unloading process and the loading process of the unit are in the same cycle period, and the unloading process and the loading process of the unit are respectively in two adjacent cycle periods; the neutron source intensity boundary value of the spent fuel assembly corresponding to the preset threshold value of the counting rate is obtained according to the simulation result, the enrichment, and the refueling cycle information, and further includes:

[0016] When the enrichment of the spent fuel assembly is a second preset enrichment and the unloading process and the loading process of the unit are in the same cycle, obtaining, according to the simulation result, a third burnup depth boundary value required for setting the spent fuel assembly at a preset core position corresponding to the preset counting rate threshold;

[0017] When the enrichment of the spent fuel assembly is a second preset enrichment and the unloading process and the loading process of the unit are respectively in two adjacent cycles, a fourth burnup depth boundary value required for setting the spent fuel assembly at a preset core position corresponding to the preset counting rate threshold is obtained according to the simulation result.

[0018] In one embodiment, the refueling process includes a loading process and an unloading process; and the refueling process for the unit according to the burnup depth boundary value includes:

[0019] When the unit is undergoing a fuel loading process, starting the fuel loading process at the preset core position according to the burnup depth boundary value;

[0020] When the unit is performing unloading processing, the unloading processing is ended at the preset core position according to the burnup depth boundary value.

[0021] A CPR1000 unit refueling device based on a spent fuel assembly is applied to a core system, wherein the core system is provided with a power detector; the refueling device comprises:

[0022] A simulation module, used for simulating the neutron transport process around the core according to a preset threshold value of the counting rate of the power detector and a preset simulation model to output a simulation result;

[0023] a burnup analysis module connected to the simulation module, and configured to determine, based on the simulation result, a burnup depth boundary value required for the spent fuel assembly corresponding to the preset count rate threshold to be set at a preset core position;

[0024] A material replacement module is connected to the burnup analysis module and is used to perform a material replacement process on the unit according to the burnup depth boundary value so that the power detection result of the unit is within a preset power detection range.

[0025] A spent fuel assembly for refueling a core system unit, wherein a preset core position of the spent fuel assembly is distributed at the outermost periphery of the core, and when the unit is undergoing a refueling process, an initial enrichment of the spent fuel assembly is 4% or 4.45%.

[0026] In one embodiment, when the initial enrichment of the spent fuel assembly is 4% and within the first preset unloading time, the burnup depth boundary value of the spent fuel assembly is 25000 MWd / tU;

[0027] When the initial enrichment of the spent fuel assembly is 4% and within the second preset unloading time, the burnup depth boundary value of the spent fuel assembly is 30000 MWd / tU.

[0028] In one embodiment, when the initial enrichment of the spent fuel assembly is 4.45% and within the first preset unloading time, the neutron source intensity boundary value of the spent fuel assembly is 26000MWd / tU;

[0029] When the initial enrichment of the spent fuel assembly is 4.45% and within the second preset unloading time, the burnup depth boundary value of the spent fuel assembly is 32000 MWd / tU.

[0030] A computer device comprises a memory and a processor, wherein the memory stores a computer program, and the processor implements the steps of the above method when executing the computer program.

[0031] A computer-readable storage medium stores a computer program, which implements the steps of the above method when executed by a processor.

[0032] The above-mentioned CPR1000 unit refueling method, device and spent fuel assembly based on spent fuel assemblies simulate the neutron transport process around the core according to the preset threshold value of the counting rate of the power detector and the preset simulation model to output the simulation result, determine the burnup depth boundary value required for the spent fuel assembly corresponding to the preset threshold value of the counting rate to be set at the preset core position according to the simulation result, refuel the unit according to the burnup depth boundary value so that the power detection result of the unit is within the preset power detection range, realize the effective replacement of secondary neutron source assemblies with spent fuel assemblies, improve the unit refueling efficiency, and ensure the smooth operation of the nuclear power plant unit. BRIEF DESCRIPTION OF THE DRAWINGS

[0033] In order to more clearly illustrate the embodiments of the present application or the technical solutions in the prior art, the drawings required for use in the embodiments or the description of the prior art will be briefly introduced below. Obviously, the drawings described below are only some embodiments of the present application. For ordinary technicians in this field, other drawings can be obtained based on these drawings without paying creative work.

[0034] Figure 1 It is a schematic flow chart of a unit refueling method based on spent fuel assemblies in one embodiment;

[0035] Figure 2 is a specific flow chart of step 104 in one embodiment;

[0036] Figure 3 This is a schematic diagram of a specific flow chart of step 204 in one embodiment;

[0037] Figure 4 This is a schematic diagram of a specific flow chart of step 204 in one embodiment;

[0038] Figure 5 This is a specific flow chart of step 106 in one embodiment;

[0039] Figure 6 A schematic diagram of a spent fuel assembly loading process in one embodiment;

[0040] Figure 7 is a front view of a baffle assembly in one embodiment;

[0041] Figure 8is a top view of a baffle assembly in one embodiment;

[0042] Fig. 9 A schematic diagram of a spent fuel assembly unloading process in one embodiment;

[0043] Fig.10 The figure is a schematic diagram of the structure of a CPR1000 unit refueling device based on a spent fuel assembly in one embodiment. DETAILED DESCRIPTION

[0044] In order to make the purpose, technical solution and advantages of the present application more clearly understood, the present application is further described in detail below in conjunction with the accompanying drawings and embodiments. It should be understood that the specific embodiments described herein are only used to explain the present application and are not used to limit the present application.

[0045] It is understood that the terms "first", "second", etc. used in this application may be used herein to describe various elements, but these elements are not limited by these terms. These terms are only used to distinguish a first element from another element. For example, without departing from the scope of this application, a first client may be referred to as a second client, and similarly, a second client may be referred to as a first client. Both the first client and the second client are clients, but they are not the same client.

[0046] See also Figure 1 , is a schematic flow chart of a unit refueling method based on spent fuel assemblies in one embodiment.

[0047] In this embodiment, the unit refueling method based on the spent fuel assembly is applied to the core system of a nuclear power plant, and the core system is provided with a power detector. Figure 1 As shown, the unit refueling method based on spent fuel assemblies includes steps 102 to 106.

[0048] Step 102 , simulating the neutron transport process around the core according to a preset threshold value of the count rate of the power detector and a preset simulation model to output a simulation result.

[0049] Optionally, the power detector may be a source range detector, used to measure the nuclear power of the reactor starting from a subcritical shutdown state to a critical state; the preset threshold value of the counting rate of the power detector may be the sensitivity of the source range detector, used to indicate the ability of the source range detector to capture leaked neutrons; the preset simulation model may be a Monte Carlo simulation algorithm; the neutron transport process around the core refers to the transport process of neutrons from the inside to the outside of the reactor.

[0050] Optionally, the method for obtaining the preset threshold value of the counting rate of the power detector can be obtained according to the factory sensitivity of the power detector. For example, the factory sensitivity of the power detector is: 9±10% (c / s) / (n·cm -2 ·s -1 ), considering the reserved conservative margin, the sensitivity of the power detector is actually calculated as: 7(c / s) / (n·cm -2 ·s -1 ), taking into full account the ability of the power detector to capture leaked neutrons.

[0051] Step 104 , determining, based on the simulation results, a burnup depth boundary value required for setting the spent fuel assembly corresponding to the preset count rate threshold at the preset core position.

[0052] Optionally, the burnup depth boundary value refers to the burnup depth boundary value reached by the spent fuel assembly unloaded from the core; the method for determining the burnup depth boundary value may be to obtain a preset neutron source intensity value corresponding to a preset counting rate threshold value based on the simulation results, obtain a mapping relationship between the neutron source intensity and the burnup depth of the spent fuel assembly, and obtain the burnup depth boundary value required for the spent fuel assembly to be set at a preset core position based on the preset neutron source intensity value and the mapping relationship.

[0053] Optionally, the preset neutron source intensity value refers to the number of neutrons emitted by the neutron source per unit time; the method for obtaining the preset neutron source intensity value can be obtained by theoretical calculation using the ORIGEN-S program or the Monte Carlo particle transport program.

[0054] Optionally, the method for obtaining the mapping relationship between the neutron source intensity and the burnup depth of the spent fuel assembly can be obtained through a large amount of historical data of the spent fuel assembly; within a certain range, the neutron source intensity of the spent fuel assembly is mainly related to factors such as the initial enrichment of the fuel assembly, the burnup depth and the shutdown decay time, and a large number of theoretical results show that: the smaller the initial enrichment of the spent fuel assembly, the greater the neutron source intensity; the greater the burnup depth of the spent fuel assembly, the greater the neutron source intensity; the longer the decay time of the spent fuel assembly, the smaller the neutron source intensity; in the early stage of decay, the neutron source intensity decays faster, and as the decay time increases, the neutron source intensity decays gradually.

[0055] Optionally, the preset core position may be a position outside the core system of a nuclear power plant unit.

[0056] Step 106: Perform fuel replacement processing on the unit according to the burnup depth boundary value so that the power detection result of the unit is within a preset power detection range.

[0057] Optionally, the refueling process includes loading and unloading of spent fuel assemblies in the core of a nuclear power plant; the power detection result is within a preset power detection range, which may be a spent fuel assembly burnup depth boundary value that meets the power detector count rate requirement.

[0058] The CPR1000 unit refueling method based on spent fuel assemblies provided in this embodiment simulates the neutron transport process around the core according to a preset threshold value of the count rate of the power detector and a preset simulation model to output a simulation result, determines the burnup depth boundary value required for setting the spent fuel assembly corresponding to the preset threshold value of the count rate at a preset core position according to the simulation result, refuels the unit according to the burnup depth boundary value so that the power detection result of the unit is within the preset power detection range, realizes the effective replacement of the secondary neutron source assembly with the spent fuel assembly, improves the unit refueling efficiency, and ensures the smooth operation of the nuclear power plant unit.

[0059] See also Figure 2 , which is a specific flow chart of step 104 in an embodiment.

[0060] In this embodiment, if Figure 2 As shown, step 104 includes sub-steps 202 to 204 .

[0061] Step 202, obtaining the enrichment of the spent fuel assembly and the refueling cycle information of the unit.

[0062] Step 204, based on the simulation results, enrichment, and refueling cycle information, a burnup depth boundary value required for setting the spent fuel assembly corresponding to the preset counting rate threshold at the preset core position is obtained.

[0063] Optionally, enrichment refers to the enrichment generally refers to the mass fraction of U235 in nuclear fuel, that is, the conversion of the abundance of U235 in nuclear fuel rods (abundance refers to the ratio of the number of nucleons); the refueling cycle information includes that the unloading processing and loading processing of the unit are in the same cycle period, that is, the refueling operation and the unloading processing are within the same nuclear power plant unit overhaul cycle; the refueling cycle information also includes that the unloading processing and loading processing of the unit are respectively located in two adjacent cycle periods, that is, the unloading processing is within the previous nuclear power plant unit overhaul cycle and the loading processing is within the current nuclear power plant unit overhaul cycle.

[0064] See also Figure 3 , which is a specific flow chart of step 204 in an embodiment.

[0065] In this embodiment, if Figure 3 As shown, step 204 also includes sub-steps 302 to 304 .

[0066] Step 302, when the enrichment of the spent fuel assembly is a first preset enrichment and the unloading process and the loading process of the unit are in the same cycle, a first burnup depth boundary value required for setting the spent fuel assembly corresponding to the preset counting rate threshold at the preset core position is obtained according to the simulation result.

[0067] Specifically, when a spent fuel assembly is replaced in a nuclear power plant unit, when the initial enrichment of the spent fuel assembly used in the preset core position is 4%, and the unloading and loading processes of the unit are in the same cycle, the first neutron source intensity boundary value of the spent fuel assembly is greater than 25000MWd / tU.

[0068] Step 304, when the enrichment of the spent fuel assembly is the first preset enrichment and the unloading process and the loading process of the unit are respectively in two adjacent cycles, a second burnup depth boundary value required for setting the spent fuel assembly corresponding to the preset counting rate threshold at the preset core position is obtained according to the simulation results.

[0069] Specifically, when a spent fuel assembly is replaced in a nuclear power plant unit, when the initial enrichment of the spent fuel assembly used in the preset core position is 4%, and the unloading process and the loading process of the unit are respectively located in two adjacent cycles, the second neutron source intensity boundary value of the spent fuel assembly is greater than 30,000 MWd / tU.

[0070] See also Figure 4 , which is a specific flow chart of step 204 in an embodiment.

[0071] In this embodiment, if Figure 4 As shown, step 204 also includes sub-steps 402 to 404 .

[0072] Step 402, when the enrichment of the spent fuel assembly is a second preset enrichment and the unloading process and the loading process of the unit are in the same cycle, a third burnup depth boundary value required for setting the spent fuel assembly corresponding to the preset counting rate threshold at the preset core position is obtained according to the simulation results.

[0073] Specifically, when a spent fuel assembly is replaced in a nuclear power plant unit, when the initial enrichment of the spent fuel assembly used in the preset core position is 4.45%, and the unloading and loading processes of the unit are in the same cycle, the third neutron source intensity boundary value of the spent fuel assembly is greater than 26000MWd / tU.

[0074] Step 404, when the enrichment of the spent fuel assembly is the second preset enrichment and the unloading process and the loading process of the unit are respectively in two adjacent cycles, a fourth burnup depth boundary value required for setting the spent fuel assembly corresponding to the preset counting rate threshold at the preset core position is obtained according to the simulation results.

[0075] Specifically, when a spent fuel assembly is replaced in a nuclear power plant unit, when the initial enrichment of the spent fuel assembly used in the preset core position is 4.45%, and the unloading and loading processes of the unit are respectively located in two adjacent cycles, the fourth neutron source intensity boundary value of the spent fuel assembly is greater than 32000MWd / tU.

[0076] See also Figure 5 , which is a specific flow chart of step 106 in one embodiment.

[0077] In this embodiment, if Figure 5 As shown, step 106 includes sub-steps 502 to 504 .

[0078] Step 502, when the unit is performing a fuel loading process, the fuel loading process is started at a preset core position according to the burnup depth boundary value.

[0079] Optionally, when performing spent fuel assembly loading processing on a nuclear power plant unit, spent fuel assemblies that meet the burnup depth boundary value are loaded preferentially at preset core positions, and then spent fuel assemblies are loaded at other core positions of the unit.

[0080] Specifically, Figure 6 The figure is a schematic diagram of the spent fuel assembly loading process in one embodiment. When the spent fuel assembly loading process is performed on the nuclear power plant unit, the core positions are preset, including position A09 and position R09; also including position A07 and position R07; Figure 6 The numbers corresponding to the core positions indicate the order of loading processing, that is, the spent fuel assemblies that meet the burnup depth boundary value are loaded at the A09 (or A07) position fuel assembly and the R09 (or R07) position first, and then the spent fuel assemblies are loaded at other core positions of the unit.

[0081] In one embodiment, the secondary neutron sources in the fuel assemblies at positions C08 and N08 of the core are replaced with choke plug assemblies, such as Figure 7 and Figure 8 The front view and top view of the choke plug assembly are shown. The choke plug assembly is divided into an upper clamping system and a lower stainless steel rod. The upper clamping system is consistent with the design of the secondary neutron source assembly. Optionally, the lower part contains 24 choke plug rods (short stainless steel rods). The length of the upper clamping system is 111.38 mm, and the length of the lower stainless steel rod is 201.9 mm.

[0082] Step 504, when the unit is performing unloading processing, the unloading processing is ended at a preset core position according to the burnup depth boundary value.

[0083] Optionally, when unloading spent fuel assemblies of a nuclear power plant unit, unloading is not performed at a preset core position first, and unloading is performed at the preset core position after unloading at other core positions of the unit is completed.

[0084] Specifically, Fig. 9 The figure is a schematic diagram of spent fuel assembly unloading processing in one embodiment. When the spent fuel assembly unloading processing is performed on the nuclear power plant unit, the core positions are preset, including position A09 and position R09; also including position A07 and position R07; Figure 6 The numbers corresponding to the core positions indicate the order of unloading processing. The fuel assemblies at positions A09 (or A07) and R09 (or R07) remain unchanged; after all other fuel assemblies are unloaded, the fuel assemblies at positions A09 (or A07) and R09 (or R07) are unloaded.

[0085] According to theoretical calculations, it can be ensured that after the implementation of loading and startup without secondary neutron source, the source range detector still has an effective counting rate during the evaporator maintenance and refueling, cold shutdown, hot shutdown, and transition from hot shutdown to cold shutdown. The implementation of loading and startup without secondary neutron source will not affect the conclusion of boron dilution accident analysis. Since loading and startup without secondary neutron source does not affect the physical and thermal parameters of the core, for the subcritical uncontrolled rod lifting accident, if the nuclear power plant's external power range detector is available, the conclusion of the accident analysis is still valid.

[0086] After the nuclear power plant core is fully loaded, the count rate of the source range detector without faults must be greater than 2C / s, which meets the requirements of the startup physical test procedures. Therefore, the implementation of the non-secondary neutron source loading and the criticality reaching mode after startup and the zero-power physical experiment mode remain unchanged, and have no impact on the startup physical test. The impact of the non-secondary neutron source on the relevant data during the startup physical test is limited to the source range detector count rate and the intermediate current value level during the hot shutdown and criticality period. After the core reaches criticality, the fission contribution to the source range detector count rate is much greater than that of neutrons. Therefore, the impact of the non-secondary neutron source on the source range and intermediate range after reaching criticality can be ignored.

[0087] Most of the tritium produced in the coolant is discharged into the environment through liquid effluent (accounting for more than 90%). The annual emission of liquid tritium can reflect the amount of tritium produced in the coolant. Referring to the operating practice of other reactor types that have eliminated the secondary neutron source, it is shown that the elimination of the secondary neutron source has a great contribution to the impact of tritium emissions. Combined with theoretical tritium emission predictions, for example, after the secondary neutron source was eliminated in the CPR1000 core, the tritium emissions of the unit were reduced by more than 30%.

[0088] The following table lists the results of the on-site source range response test for the refueling process of the nuclear power plant unit L310 based on steps 102 to 108.

[0089]

[0090] From the data in the table, it can be seen that the source range 1 (L3RPN014MA) △C = 41% and the source range 2 (L3RPN024MA) △C = 67% of the downline measurement of the L310 overhaul of the nuclear power plant unit. Although the △C of source range 2 is greater than 60%, it indicates that the probe has begun to age seriously and needs to be replaced in this round of overhaul. From the response test results, it can be seen that when the burnup depth of the fuel assemblies unloaded during this overhaul is 20200MWd / tU, the source range can be guaranteed to obtain a count greater than 2C / s; for the fuel assemblies unloaded during the previous overhaul, the source range count can reach 5C / s when the burnup depth is 35936MWd / tU, which is basically consistent with the theoretical calculation. It ensures that the core is in a state of continuous monitoring during the loading process, and also proves the correctness and conservatism of the theoretical demonstration of loading and startup without secondary neutron sources.

[0091] It should be understood that although Figures 1 to 5 The steps in the flowchart are shown in sequence as indicated by the arrows, but the steps are not necessarily executed in the order indicated by the arrows. Unless otherwise specified in this document, there is no strict order restriction for the execution of the steps, and the steps can be executed in other orders. Moreover, Figures 1 to 5 At least part of the steps in the above method may include multiple sub-steps or multiple stages, and these sub-steps or stages are not necessarily executed at the same time, but can be executed at different times. The execution order of these sub-steps or stages is not necessarily sequential, but can be executed in turn or alternately with other steps or at least part of the sub-steps or stages of other steps. It should be noted that the above different embodiments can be combined with each other.

[0092] See also Fig.10 , is a schematic structural diagram of a CPR1000 unit refueling device based on a spent fuel assembly in one embodiment.

[0093] In this embodiment, the unit refueling device based on the spent fuel assembly is applied to the core system, and the core system is provided with a power detector. Fig.10 As shown, the unit refueling device based on spent fuel assemblies includes a simulation module 1010 , a burnup analysis module 1020 and a refueling module 1030 .

[0094] The simulation module 1010 is used to simulate the neutron transport process around the core according to the preset threshold value of the counting rate of the power detector and the preset simulation model to output the simulation result.

[0095] The burnup analysis module 1020 is connected to the simulation module 1010 and is used to determine the burnup depth boundary value required for setting the spent fuel assembly corresponding to the preset count rate threshold at the preset core position according to the simulation results.

[0096] The fuel replacement module 1030 is connected to the fuel consumption analysis module 1020, and is used to perform a fuel replacement process on the unit according to the fuel consumption depth boundary value so that the power detection result of the unit is within a preset power detection range.

[0097] In this embodiment, each module is used to execute Figure 1 For details of the steps in the corresponding embodiments, please refer to Figure 1 as well as Figure 1 The relevant descriptions in the corresponding embodiments are not repeated here.

[0098] The unit refueling device based on spent fuel assemblies provided in this embodiment simulates the neutron transport process around the core according to the preset threshold value of the counting rate of the power detector and the preset simulation model through the simulation module 1010 to output the simulation result. The burnup analysis module 1020 connected to the simulation module 1010 determines the burnup depth boundary value required for the spent fuel assembly corresponding to the preset threshold value of the counting rate to be set at the preset core position according to the simulation result. The refueling module 1030 connected to the burnup analysis module 1020 refuels the unit according to the burnup depth boundary value so that the power detection result of the unit is within the preset power detection range, thereby realizing the effective replacement of the secondary neutron source assembly with the spent fuel assembly, improving the unit refueling efficiency, and ensuring the smooth operation of the nuclear power plant unit.

[0099] In one embodiment, a spent fuel assembly for core system unit refueling is provided. When the preset core position of the spent fuel assembly is distributed at the outermost periphery of the core, the initial enrichment of the spent fuel assembly is 4% or 4.45%.

[0100] In one embodiment, a spent fuel assembly for refueling of a CPR1000 core system unit is provided, wherein when the initial enrichment of the spent fuel assembly is 4% and within a first preset unloading time, the neutron source intensity boundary value of the spent fuel assembly is 25000MWd / tU or 30000MWd / tU; when the initial enrichment of the spent fuel assembly is 4% and within a second preset unloading time, the burnup depth boundary value of the spent fuel assembly is 30000MWd / tU; when the initial enrichment of the spent fuel assembly is 4.45% and within the first preset unloading time, the burnup depth boundary value of the spent fuel assembly is 26000MWd / tU; when the initial enrichment of the spent fuel assembly is 4.45% and within the second preset unloading time, the burnup depth boundary value of the spent fuel assembly is 32000MWd / tU.

[0101] Optionally, the first preset unloading time may be that the time from the reactor core to the spent fuel pool is less than 60 days; the second preset unloading time may be that the time from the reactor core to the spent fuel pool is less than the sum of one refueling cycle and 60 days.

[0102] Specifically, when the initial enrichment of the spent fuel assembly is 4%, the time for unloading from the reactor core to the spent fuel pool is less than 60 days, and the unloading and loading processes of the unit are in the same cycle, the neutron source intensity boundary value of the spent fuel assembly is greater than 25000MWd / tU; when the initial enrichment of the spent fuel assembly is 4%, the time for unloading from the reactor core to the spent fuel pool is less than the sum of one refueling cycle and 60 days, and the unloading and loading processes of the unit are in two adjacent cycles, the neutron source intensity boundary value of the spent fuel assembly is greater than 30000MWd / tU; when When the initial enrichment of the spent fuel assembly is 4.45%, and the unloading and loading processes of the unit are in the same cycle, the neutron source intensity boundary value of the spent fuel assembly is greater than 26000MWd / tU, and the time for unloading from the reactor core to the spent fuel pool is less than 60 days; when the initial enrichment of the spent fuel assembly is 4.45%, the time for unloading from the reactor core to the spent fuel pool is less than the sum of one refueling cycle and 60 days, and the unloading and loading processes of the unit are respectively in two adjacent cycles, the neutron source intensity boundary value of the spent fuel assembly is greater than 32000MWd / tU.

[0103] The division of each module in the above-mentioned CPR1000 unit refueling method, device and spent fuel assembly based on spent fuel assembly is only for illustration. In other embodiments, the CPR1000 unit refueling method, device and spent fuel assembly based on spent fuel assembly can be divided into different modules as needed to complete all or part of the functions of the above-mentioned CPR1000 unit refueling method, device and spent fuel assembly based on spent fuel assembly.

[0104] For the specific definition of the CPR1000 unit refueling device based on spent fuel assemblies, please refer to the definition of the CPR1000 unit refueling method based on spent fuel assemblies in the above text, which will not be repeated here. Each module in the above-mentioned CPR1000 unit refueling device based on spent fuel assemblies can be implemented in whole or in part by software, hardware and a combination thereof. The above-mentioned modules can be embedded in or independent of the processor in the computer device in the form of hardware, or can be stored in the memory in the computer device in the form of software, so that the processor can call and execute the operations corresponding to the above modules.

[0105] A computer device is also provided in an embodiment of the present application, including a memory and a processor. A computer program is stored in the memory. When the computer program is executed by the processor, the processor executes the steps of the method in the above embodiment.

[0106] The present application also provides a computer-readable storage medium, one or more non-volatile computer-readable storage media containing computer-executable instructions, which, when executed by one or more processors, cause the processors to execute the steps of the CPR1000 unit refueling method based on spent fuel assemblies.

[0107] The CPR1000 unit refueling method, device and spent fuel assembly based on spent fuel assemblies provided in the above embodiments realize the effective replacement of secondary neutron source assemblies with spent fuel assemblies during the refueling process of nuclear power plant units, improve the efficiency of unit refueling, ensure the smooth operation of nuclear power plant units, and have important economic value and promotion and practical value.

[0108] Any reference to memory, storage, database or other medium used in this application may include non-volatile and / or volatile memory. Non-volatile memory may include read-only memory (ROM), programmable ROM (PROM), electrically programmable ROM (EPROM), electrically erasable programmable ROM (EEPROM) or flash memory. Volatile memory may include random access memory (RAM), which is used as an external cache memory. As an illustration and not limitation, RAM is available in many forms, such as static RAM (SRAM), dynamic RAM (DRAM), synchronous DRAM (SDRAM), double data rate SDRAM (DDR SDRAM), enhanced SDRAM (ESDRAM), synchronous link (Synchlink) DRAM (SLDRAM), memory bus (Rambus) direct RAM (RDRAM), direct memory bus dynamic RAM (DRDRAM), and memory bus dynamic RAM (RDRAM).

[0109] The technical features of the above-described embodiments may be arbitrarily combined. To make the description concise, not all possible combinations of the technical features in the above-described embodiments are described. However, as long as there is no contradiction in the combination of these technical features, they should be considered to be within the scope of this specification.

[0110] The above-mentioned embodiments only express several implementation methods of the present invention, and the descriptions thereof are relatively specific and detailed, but they cannot be understood as limiting the scope of the invention patent. It should be pointed out that, for ordinary technicians in this field, several variations and improvements can be made without departing from the concept of the present invention, and these all belong to the protection scope of the present invention. Therefore, the protection scope of the patent of the present invention shall be subject to the attached claims.

Claims

1. A CPR1000 unit refueling method based on spent fuel assemblies, characterized in that: Applied to a core system, the core system is provided with a power detector; the material replacement method comprises: The neutron transport process around the core is simulated according to the preset count rate threshold of the power detector and the preset simulation model to output a simulation result; the preset count rate threshold of the power detector is used to indicate the ability of the source range detector to capture leaked neutrons; the neutron transport process around the core refers to the transport process of neutrons from the inside to the outside of the reactor; Determine, according to the simulation result, a burnup depth boundary value required for setting the spent fuel assembly corresponding to the preset count rate threshold at a preset core position; Performing a fuel replacement process on the unit according to the burnup depth boundary value so that the power detection result of the unit is within a preset power detection range; Wherein, determining, according to the simulation result, a burnup depth boundary value required for setting the spent fuel assembly corresponding to the preset count rate threshold at a preset core position includes: A preset neutron source intensity value corresponding to the preset counting rate threshold is obtained according to the simulation result, a mapping relationship between the neutron source intensity and the burnup depth of the spent fuel assembly is obtained, and the burnup depth boundary value required for setting the spent fuel assembly at a preset core position is obtained according to the preset neutron source intensity value and the mapping relationship.

2. The material replacement method according to claim 1, characterized in that: The step of determining, according to the simulation result, a burnup depth boundary value required for setting the spent fuel assembly corresponding to the preset count rate threshold at a preset core position includes: Obtaining the enrichment of the spent fuel assembly and refueling cycle information of the unit; A burnup depth boundary value required for setting the spent fuel assembly at a preset core position corresponding to the preset counting rate threshold is obtained according to the simulation result, the enrichment, and the refueling cycle information.

3. The material replacement method according to claim 2, characterized in that: The refueling cycle information includes that the unloading process and the loading process of the unit are in the same cycle period, and the unloading process and the loading process of the unit are respectively in two adjacent cycle periods; the burnup depth boundary value required for setting the spent fuel assembly corresponding to the preset counting rate threshold value at the preset core position according to the simulation result, the enrichment, and the refueling cycle information includes: When the enrichment of the spent fuel assembly is a first preset enrichment and the unloading process and the loading process of the unit are in the same cycle, obtaining, according to the simulation result, a first burnup depth boundary value required for setting the spent fuel assembly at a preset core position corresponding to the preset counting rate threshold; When the enrichment of the spent fuel assembly is a first preset enrichment and the unloading process and the loading process of the unit are respectively in two adjacent cycles, a second burnup depth boundary value required for setting the spent fuel assembly at a preset core position corresponding to the preset counting rate threshold is obtained according to the simulation result.

4. The material replacement method according to claim 2, characterized in that: The refueling cycle information includes that the unloading process and the loading process of the unit are in the same cycle period, and the unloading process and the loading process of the unit are respectively in two adjacent cycle periods; the neutron source intensity boundary value of the spent fuel assembly corresponding to the preset threshold value of the counting rate is obtained according to the simulation result, the enrichment, and the refueling cycle information, and also includes: When the enrichment of the spent fuel assembly is a second preset enrichment and the unloading process and the loading process of the unit are in the same cycle, obtaining, according to the simulation result, a third burnup depth boundary value required for setting the spent fuel assembly at a preset core position corresponding to the preset counting rate threshold; When the enrichment of the spent fuel assembly is a second preset enrichment and the unloading process and the loading process of the unit are respectively in two adjacent cycles, a fourth burnup depth boundary value required for setting the spent fuel assembly at a preset core position corresponding to the preset counting rate threshold is obtained according to the simulation result.

5. The material replacement method according to claim 1, characterized in that: The refueling process includes loading process and unloading process; the refueling process of the unit according to the burnup depth boundary value includes: When the unit is undergoing a fuel loading process, starting the fuel loading process at the preset core position according to the burnup depth boundary value; When the unit is performing unloading processing, the unloading processing is ended at the preset core position according to the burnup depth boundary value.

6. A CPR1000 unit refueling device based on spent fuel assemblies, characterized in that: Applicable to a core system, the core system is provided with a power detector; the material changing device comprises: A simulation module, used for simulating the neutron transport process around the core according to the preset threshold value of the counting rate of the power detector and the preset simulation model to output the simulation result; the preset threshold value of the counting rate of the power detector is used to indicate the ability of the source range detector to capture the leaked neutrons; the neutron transport process around the core refers to the transport process of neutrons from the inside to the outside of the reactor; a burnup analysis module connected to the simulation module, and configured to determine, based on the simulation result, a burnup depth boundary value required for the spent fuel assembly corresponding to the preset count rate threshold to be set at a preset core position; a refueling module, connected to the burnup analysis module, and configured to perform a refueling process on the unit according to the burnup depth boundary value so that the power detection result of the unit is within a preset power detection range; Among them, the burnup analysis module is also used to obtain a preset neutron source intensity value corresponding to the preset counting rate threshold according to the simulation result, obtain a mapping relationship between the neutron source intensity and the burnup depth of the spent fuel assembly, and obtain the burnup depth boundary value required for the spent fuel assembly to be set at a preset core position according to the preset neutron source intensity value and the mapping relationship.

7. A spent fuel assembly for refueling a CPR1000 core system unit, which is used in the CPR1000 unit refueling method based on spent fuel assemblies as described in claim 1, characterized in that: The preset core position of the spent fuel assembly is distributed at the outermost periphery of the core. When the unit is undergoing a refueling process, the initial enrichment of the spent fuel assembly is 4% or 4.45%.

8. The spent fuel assembly according to claim 7, characterized in that: When the initial enrichment of the spent fuel assembly is 4% and within the first preset unloading time, the burnup depth boundary value of the spent fuel assembly is 25000 MWd / tU; When the initial enrichment of the spent fuel assembly is 4% and within the second preset unloading time, the burnup depth boundary value of the spent fuel assembly is 30000 MWd / tU.

9. The spent fuel assembly according to claim 7, characterized in that: When the initial enrichment of the spent fuel assembly is 4.45% and within the first preset unloading time, the neutron source intensity boundary value of the spent fuel assembly is 26000MWd / tU; When the initial enrichment of the spent fuel assembly is 4.45% and within the second preset unloading time, the burnup depth boundary value of the spent fuel assembly is 32000 MWd / tU.

10. A computer device comprising a memory and a processor, wherein the memory stores a computer program, characterized in that: When the processor executes the computer program, the steps of the method according to any one of claims 1 to 5 are implemented.

11. A computer-readable storage medium having a computer program stored thereon, characterized in that: When the computer program is executed by a processor, the steps of the method according to any one of claims 1 to 5 are implemented.

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