Method and system for optimizing decay cooling time of fuel assembly after reactor subcriticality

Abstract

This invention discloses a method and system for optimizing the decay cooling time of fuel assemblies after subcritical reactor operation. The optimization method includes: obtaining the shortened time to resume fuel operation after reactor shutdown; analyzing the radioactive consequences of a fuel operation accident through modeling based on the time to resume fuel operation after shutdown, and determining whether a first preset condition is met; analyzing the cooling capacity of the spent fuel pool through modeling based on the time to resume fuel operation after shutdown, and determining whether a second preset condition is met; analyzing the radiation shielding of the fuel building through modeling, and determining whether a third preset condition is met; and when all three preset conditions are met, using the time to resume fuel operation after shutdown as the decay cooling time of the fuel assemblies after subcritical operation. Implementing this invention allows for the direct reduction of the fuel assembly decay cooling time requirement before overhaul unloading without requiring physical modifications.

Description

Technical Field

[0001] This invention relates to the field of nuclear power plant operation, and in particular to a method and system for optimizing the decay cooling time of fuel assemblies after subcritical reactors. Background Technology

[0002] According to the technical specifications, during the refueling process of a pressurized water reactor nuclear power unit, moving fuel assemblies is not permitted for a period of time after subcriticality. This period allows the fuel assemblies to decay and cool, reducing the accumulation of fission products and decay heat power. The minimum required time is defined as the post-subcritical fuel assembly decay cooling time.

[0003] Currently, most pressurized water reactor types require a 100-hour decay cooling time for fuel assemblies after subcriticality. Current methods for determining this decay cooling time primarily consider the radioactive consequences of fuel handling accidents, assuming that this time requirement is sufficient to ensure adequate decay time for fission products within the fuel assembly.

[0004] During a unit overhaul, if the preparatory procedures before fuel handling cannot be completed within 100 hours, the fuel assembly decay cooling time requirement will not affect the overhaul schedule. However, when the preparation time can be shortened to within 100 hours, this time requirement will become a critical path affecting the unit's overhaul schedule. In recent years, with the continuous improvement of overhaul management levels of various nuclear power groups, the completion time of preparatory procedures before unloading has been continuously shortened. Recently, there have been several instances where overhaul unloading has been affected by waiting for the fuel assembly decay cooling time to meet the 100-hour requirement after subcriticality. Summary of the Invention

[0005] The technical problem to be solved by the present invention is to provide a method and system for optimizing the decay cooling time of fuel assemblies after subcritical reactors, in view of the above-mentioned defects in the prior art.

[0006] The technical solution adopted by this invention to solve its technical problem is: constructing an optimization method for the decay cooling time of fuel assemblies after subcritical reactors, comprising:

[0007] Step S10: Obtain the shortened time to start fuel operation after shutdown to be determined;

[0008] Step S20: Based on the time of fuel operation after reactor shutdown, analyze the radioactive consequences of the fuel operation accident through modeling and determine whether the first preset condition is met.

[0009] Step S30: Based on the time for refueling to resume after reactor shutdown, analyze the cooling capacity of the spent fuel pool by modeling and determine whether the second preset condition is met.

[0010] Step S40: Based on the time for refueling to resume after reactor shutdown, analyze the radiation shielding of the fuel plant through modeling and determine whether the third preset condition is met.

[0011] Step S50: When the first preset condition, the second preset condition, and the third preset condition are all satisfied, the time after the reactor shutdown to start fuel operation is taken as the subcritical post-fuel assembly decay cooling time.

[0012] Preferably, step S20 includes:

[0013] Calculate the amount of fission products accumulated in the largest burnup component based on the time after the reactor shutdown resumes fuel operation.

[0014] Obtain a computational model pre-established based on each stage of the radioactive material migration path, wherein the computational model is used to describe the diffusion behavior and impact process of the radioactive material;

[0015] Based on the calculation model and the amount of fission products accumulated in the maximum burnup component, the radioactive consequences of fuel handling accidents are analyzed, and it is determined whether the preset environmental radioactivity conditions are met.

[0016] Based on the calculation model and the accumulation of fission products of the maximum fuel consumption component, the personnel dosage in the main control room of the fuel operation accident is analyzed, and it is determined whether the preset conditions for the main control room to remain are met.

[0017] Preferably, the calculation of the fission product accumulation of the maximum burnup component includes:

[0018] The nuclide density at each time step was calculated using the Bateman differential equation.

[0019] The accumulation of fission products in the maximum burnup component during a fuel handling accident is calculated by simulating the nuclide density change process.

[0020] Preferably, step S30 includes:

[0021] Calculate the maximum decay heat of the fuel assembly based on the unit's fuel management information;

[0022] Based on the design information of the spent fuel pool, the unit fuel management information, and the maximum decay heat of the fuel assemblies, the maximum heat load of the spent fuel pool under different operating conditions is determined.

[0023] Based on the time to start fuel operation after reactor shutdown, the maximum decay heat of the fuel assembly, and the maximum heat load of the spent fuel pool, the cooling capacity of the spent fuel pool under different operating conditions is analyzed, and it is determined whether the requirements for removing the decay heat of the fuel assembly are met.

[0024] Preferably, the operating conditions include: power operation condition, normal overhaul unloading condition, unplanned full core unloading condition, design basis accident condition, and beyond design basis accident condition.

[0025] Preferably, step S40 includes:

[0026] Obtain pre-stored component modeling information, wherein the component modeling information is generated by pre-modeling a single fuel assembly based on the structure of the fuel assembly;

[0027] Obtain pre-stored area modeling information, wherein the area modeling information is generated by pre-modeling a specific area based on the layout information of fuel assemblies stored in the spent fuel pool, the loading and unloading workflow information, and the structural information around the fuel plant and transfer channel;

[0028] Based on the component modeling information and area modeling information, the radiation dose rate at specific locations around the fuel plant and transfer channel is calculated, and it is determined whether the conditions meet the corresponding radiation zoning requirements. In addition, a calculation model for the radiation dose rate at the operating platform during the operation of the mobile fuel assembly is established, and the minimum shielding water layer thickness and the minimum distance between the fuel assembly and the transfer gate are calculated based on the radiation dose rate limit at the operating platform to meet the radiation zoning requirements.

[0029] Preferably, specific locations around the fuel plant and transfer channels include: above the water surface of the spent fuel pool, outside the wall of the spent fuel pool, outside the transfer gate, and outside the loading gate.

[0030] Preferably, the minimum thickness of the shielding water layer is 270cm; the minimum distance between the fuel assembly and the transfer gate is 210cm.

[0031] Preferably, it further includes:

[0032] Based on the optimized post-subcritical fuel assembly decay cooling time, the corresponding power plant documents were modified.

[0033] The present invention also constructs an optimization system for the decay cooling time of fuel assemblies after subcritical reactors, including a processor and a memory storing a computer program. When the processor executes the computer program, it implements the steps of the above-described optimization method for the decay cooling time of fuel assemblies after subcritical reactors.

[0034] Through the technical solution of this invention, for a shortened time to start fuel operation after reactor shutdown that needs to be determined, the radioactive consequences of fuel operation accidents, the cooling capacity of spent fuel pools, and the radiation shielding of fuel buildings are analyzed by modeling to determine whether the time to start fuel operation after reactor shutdown can be used as the decay cooling time of fuel assemblies after subcriticality. Thus, the decay cooling time of fuel assemblies after subcriticality can be optimized. Therefore, the requirement to directly shorten the decay cooling time of fuel assemblies before overhaul unloading can be obtained without implementing physical modifications. Attached Figure Description

[0035] To more clearly illustrate the embodiments of the present invention, the accompanying drawings used in the description of the embodiments will be briefly introduced below. Obviously, the drawings described below are only some embodiments of the present invention. For those skilled in the art, other drawings can be obtained based on these drawings without creative effort. In the drawings:

[0036] Figure 1 This is a flowchart of Embodiment 1 of the method for optimizing the decay cooling time of fuel assemblies after subcritical reactors according to the present invention;

[0037] Figure 2 yes Figure 1 The flowchart of Example 1, which analyzes the radioactive consequences of a fuel handling accident. Detailed Implementation

[0038] The technical solutions of the embodiments of the present invention will be clearly and completely described below with reference to the accompanying drawings. Obviously, the described embodiments are only some embodiments of the present invention, and not all embodiments. Based on the embodiments of the present invention, all other embodiments obtained by those skilled in the art without creative effort are within the scope of protection of the present invention.

[0039] First, it should be noted that shortening the decay cooling time of subcritical post-critical fuel assemblies will have the following direct and indirect effects. The direct effects are: a shorter decay cooling time results in a larger accumulation of fission products in the fuel assemblies. If the fuel assemblies break during loading and unloading, it will lead to more severe accident consequences, including radioactive impacts on the environment and radiation exposure to personnel in the main control room. Furthermore, a shorter decay cooling time results in fuel assemblies with greater radioactivity during unloading, thus increasing the radiation dose received by fuel operators during loading and unloading. The indirect effects are: a shorter decay cooling time also results in higher decay heat levels in the fuel assemblies. Although the peak heat load of the spent fuel pool occurs at the completion of unloading and is not directly determined by the start time of unloading, an earlier start to the unloading operation means a correspondingly earlier completion time. Therefore, shortening the decay cooling time of subcritical post-critical fuel assemblies will increase the heat load of the spent fuel pool, posing a challenge to the cooling capacity of the reactor pool and the spent fuel pool cooling and treatment system (PTR). Therefore, a newly determined time for resuming fuel operation after reactor shutdown needs to be analyzed and verified in order to determine whether this time can be used as the subcritical post-fuel assembly decay cooling time.

[0040] Figure 1 This is a flowchart of an embodiment of the optimization method for the decay cooling time of fuel assemblies after subcritical reactors according to the present invention. The optimization method of this embodiment includes:

[0041] Step S10: Obtain the shortened time to start fuel operation after shutdown to be determined;

[0042] In this step, the time for starting fuel operation after reactor shutdown can be determined based on the completion time of the preparation process before fuel operation, for example, shortened from 100 hours to 80 hours.

[0043] Step S20: Based on the time of fuel operation after reactor shutdown, analyze the radioactive consequences of the fuel operation accident through modeling and determine whether the first preset condition is met.

[0044] Step S30: Based on the time for refueling to resume after reactor shutdown, analyze the cooling capacity of the spent fuel pool by modeling and determine whether the second preset condition is met.

[0045] Step S40: Based on the time for refueling to resume after reactor shutdown, analyze the radiation shielding of the fuel plant through modeling and determine whether the third preset condition is met.

[0046] Step S50: When the first preset condition, the second preset condition, and the third preset condition are all satisfied, the time after the reactor shutdown to start fuel operation is taken as the subcritical post-fuel assembly decay cooling time.

[0047] Regarding this embodiment, it should be noted that steps S20, S30, and S40 are not in any particular order. In practical applications, steps S20, S30, and S40 can be executed in any order.

[0048] This embodiment's technical solution addresses a shortened post-shutdown fuel operation start time by analyzing the radioactive consequences of a fuel operation accident, the cooling capacity of the spent fuel pool, and the radiation shielding of the fuel building through modeling. This analysis determines whether the post-shutdown fuel operation start time can be used as the subcritical fuel assembly decay cooling time, thereby optimizing the subcritical fuel assembly decay cooling time. Therefore, without physical modifications, the requirement for directly shortening the fuel assembly decay cooling time before overhaul unloading can be obtained. Taking a nuclear power plant with six 1,000 MW pressurized water reactor units as an example, if this embodiment's technical solution shortens the overhaul start time requirement from 100 hours to 80 hours, a maximum of 20 hours of overhaul time can be saved per overhaul, generating considerable economic benefits. Assuming a loss of 10 million yuan per day per unit shutdown, and a maximum reduction of 20 hours per overhaul, the six units can increase power generation time by 120 hours every 18 months, generating approximately 50 million yuan in economic benefits.

[0049] Further, step S20 includes:

[0050] Step S21: Calculate the amount of fission products accumulated in the maximum burnup component based on the fuel operation start time after the reactor shutdown.

[0051] Step S22: Obtain a calculation model pre-established based on each link in the radioactive material migration path, wherein the calculation model is used to describe the diffusion behavior and impact process of radioactive materials;

[0052] Step S23: Based on the calculation model and the amount of fission products accumulated in the maximum burnup component, analyze the radioactive consequences of the fuel operation accident and determine whether the preset environmental radioactivity conditions are met.

[0053] Step S24: Based on the calculation model and the amount of fission products accumulated in the maximum fuel consumption component, analyze the personnel dosage in the main control room of the fuel operation accident and determine whether the preset conditions for the main control room to remain are met.

[0054] Regarding this embodiment, it should be noted that fuel handling accidents are a direct limiting factor for the decay cooling time of the fuel assembly after subcriticality. This time requirement is used to ensure that the fission products within the fuel assembly have sufficient decay time before fuel loading and unloading operations to reduce the radioactive consequences of fuel handling accidents. If the fuel assembly decay cooling time is shortened, for example from 100 hours to 80 hours, combined with... Figure 2The environmental radioactive consequences of fuel handling accidents and the doses to personnel in the main control room still meet some limit requirements. For example, the environmental radioactive consequences of fuel handling accidents meet the limit requirements of GB6249-2011, and the doses to personnel in the main control room meet the limit requirements of HAD002 / 01-2019.

[0055] Further, step S21 includes:

[0056] The nuclide density at each time step was calculated using the Bateman differential equation.

[0057] The accumulation of fission products in the maximum burnup component during a fuel handling accident is calculated by simulating the change process of the nuclide density.

[0058] In one specific embodiment, the calculation of the accumulation of fission products in a fuel assembly essentially involves calculating the compositional changes of radionuclides within the fuel assembly during irradiation and decay. From the time the fuel assembly is loaded into the reactor until it is shut down and removed from the core, the nuclides within the assembly undergo fission, neutron capture, and decay reactions within the core. The change in the nuclear density of radionuclides within the assembly over time is the result of the combined effects of these factors. The nuclear density at each moment can be solved using the Bateman differential equation:

[0059]

[0060] In the formula:

[0061] dN i dt nuclide i nuclear density change rate

[0062]

[0063] ∑γ ji σ f,j N j The rate at which φ nuclide fission produces nuclide i

[0064] σ c,i-1 N i-1 φ The rate at which other nuclides generate nuclide i through neutron capture reactions

[0065] ∑λ k→i N k The rate at which nuclide k decays to produce nuclide i

[0066] σ f,i N i φ The rate at which nuclide i decreases due to fission reaction

[0067] σ c,i N i φ The rate at which nuclide i decreases due to neutron capture reaction

[0068] λ i N i The rate at which nuclide i decreases due to decay

[0069] Taking a second-generation pressurized water reactor nuclear power unit in China as an example, the specific assumptions used in the calculation process include:

[0070] 1. The damaged fuel assembly is exactly the assembly to be unloaded with the highest fuel consumption in all cycles. For example, the unloading fuel consumption of the damaged fuel assembly is 52000 MWD / tU.

[0071] 2. Assume that the fuel assembly is irradiated with a specific power of 40 MW / tU in the first few cycles and with a specific power of 60 MW / tU in the last cycle; in the last cycle, the fuel consumption increment of the damaged fuel assembly is 28000 MWd / tU, which is the result of rounding up the maximum value of the fuel consumption increment of a single cycle experienced by a component in all cycles.

[0072] 3. The accident assumes that all fuel rods in the assembly are damaged.

[0073] 4. Uncertainty is not considered.

[0074] Then, by using specialized programs to simulate the nuclide density change process, the amount of fuel assembly fission products accumulated in the nuclear power unit due to a fuel operation accident can be obtained.

[0075] In one specific embodiment, the analysis of environmental radioactive consequences adopts the same assumptions as those in the fuel operation accident in the FSAR (Final Safety Analysis Report). Accident source terms for radioactive consequences analysis under different operating conditions are calculated, and the environmental radioactive consequences of the nuclear power unit fuel operation accident are recalculated based on the latest known values ​​of atmospheric dispersion factor and isotope dose conversion factor.

[0076] In a specific example, when the fuel assembly decay cooling time is shortened, for example to 80 hours, the calculation model shows that the maximum effective dose for the public at the non-residential boundary 2 hours after the accident and the public at the boundary outside the planned restricted area 12 hours after the accident are 2.17 mSv and 0.647 mSv, respectively; the maximum thyroid equivalent dose is 34.6 mSv and 10.3 mSv, respectively. According to the design basis condition (DBC) classification of this nuclear power unit, the fuel operation accident is a limit accident. The "Regulations for Environmental Radiation Protection of Nuclear Power Plants" (GB6249-2011) stipulates that "in the event of a limit accident, the effective dose that the public at the non-residential boundary may receive within 2 hours after the accident and the public at the boundary outside the planned restricted area throughout the entire duration of the accident should be controlled below 0.1 Sv, and the thyroid equivalent dose should be controlled below 1 Sv." Therefore, when the fuel assembly decay cooling time is shortened to 80 hours, the environmental radioactive consequences of the fuel operation accident still meet the limits required by GB6249-2011.

[0077] In one specific embodiment, the assessment of the habitability of the main control room considers that personnel in the main control room may receive doses through the following exposure pathways during an accident: external gamma irradiation from outdoor radioactive plumes; external irradiation from radioactive material deposition; and irradiation from air pollution within the main control room, including inhalation and external irradiation. However, since the walls, floor, and ceiling provide effective shielding between the radiation source and the operating personnel in the control room, external gamma irradiation from outdoor radioactive plumes and external irradiation from radioactive material deposition can be neglected. Only the effective dose and thyroid equivalent dose from contaminated air within the main control room are calculated.

[0078] A comprehensive calculation model was established based on the method for calculating the dose of personnel in the main control room during a large-break loss-of-coolant accident in a nuclear power unit, detailing the migration path and impact process of radioactive materials. When the fuel assembly decay cooling time is shortened, for example to 80 hours, without considering air leakage, the effective dose for personnel in the habitable area of ​​the main control room is 1.09 mSv, and the thyroid equivalent dose is 1.60 mSv; considering air leakage, the effective dose for personnel in the habitable area of ​​the main control room is 3.32 mSv, and the thyroid equivalent dose is 45.7 mSv. The "Emergency Preparedness and Response of Nuclear Power Plant Operating Units" (HAD002 / 01-2019) stipulates that "the habitability criteria for important emergency facilities such as the main control room should be met as follows: during the established continuous emergency response period (generally 30 days), the effective dose received by personnel should not exceed 50 mSv, and the thyroid equivalent dose should not exceed 500 mSv." Therefore, when the fuel assembly decay cooling time is shortened to 80 hours, the dose received by personnel in the main control room during a fuel operation accident still meets the limit requirements of HAD002 / 01-2019.

[0079] Further, in an optional embodiment, step S30 includes:

[0080] Step S31: Calculate the maximum decay heat of the fuel assembly based on the unit fuel management information. For example, calculate the maximum core decay heat for the first cycle, flexible cycle, and long and short cycles under different decay cooling times, and obtain the maximum decay heat curve through envelope processing.

[0081] Step S32: Determine the maximum heat load of the spent fuel water pool under different operating conditions based on the design information of the spent fuel water pool, the unit fuel management information, and the maximum decay heat of the fuel assembly.

[0082] Step S33: Based on the time to start fuel operation after reactor shutdown, the maximum decay heat of the fuel assembly, and the maximum heat load of the spent fuel pool, analyze the cooling capacity of the spent fuel pool under different operating conditions, and determine whether it meets the requirements for removing the decay heat of the fuel assembly.

[0083] In this embodiment, spent fuel pool cooling is an indirect limiting factor for the fuel assembly decay cooling time after subcriticality. The maximum heat load of the spent fuel pool considered in the analysis corresponds to D1 (the time interval from refueling overhaul shutdown to core reloading completion), D2 (the time interval from refueling overhaul shutdown to unloading completion), and D3 (the time interval from the previous cycle unit overhaul shutdown to the shutdown and unloading completion due to an abnormal reason during this cycle's power operation), which is not directly related to the unloading start time. When the fuel assembly decay cooling time is shortened, for example, by 20 hours, the cooling capacity is evaluated under different operating conditions according to the PTR (Reactor Pool and Spent Fuel Pool Cooling and Treatment System) system design benchmarks. Specifically:

[0084] Under normal power operation conditions, the heat load of the spent fuel pool depends on the core reloading completion time and is independent of the unloading start time. To ensure that the cooling capacity of the spent fuel pool meets design specifications under normal power operation conditions, the reloading completion time should be limited to no earlier than the earliest permissible time, which can be calculated from the supply water temperature of the RRI (Component Cooling System).

[0085] Under normal refueling overhaul conditions, the heat load of the spent fuel pool depends on the unloading completion time and is independent of the unloading start time. To ensure that the cooling capacity of the spent fuel pool meets design standards under normal refueling overhaul conditions, the unloading completion time should be limited to no earlier than the earliest permissible time. This time can be calculated from the RRI supply water temperature, taking into account different PTR operating modes.

[0086] Under unplanned full-core unloading conditions, with the PTR system operating in a two-pump, two-replacement mode and considering an RRI temperature of 35°C, the temperature of the spent fuel pool in the nuclear power unit can be maintained below 60°C, and the cooling capacity of the spent fuel pool can meet the design benchmark.

[0087] Under the design basis accident condition, assuming the PTR system operates in single-pump, single-refueling mode, maintaining the spent fuel pool temperature below 80°C, and the RRI supply water temperature not exceeding 35°C, the PTR heat exchanger's heat discharge power is greater than 13.16 MW. This value exceeds the spent fuel pool heat load determined under the initial conditions of normal power operation, normal reactor shutdown and refueling, and unplanned full-core unloading, considering the post-subcritical fuel assembly decay cooling time to 80 hours. Therefore, with a shortened post-subcritical fuel assembly decay cooling time, for example, to 80 hours, the spent fuel pool cooling capacity under the design basis accident condition can meet the design basis.

[0088] Under conditions exceeding design basis, the spent fuel pool can be replenished with water through multiple methods at a rate greater than 23 t / h. Considering a replenishment temperature of 40°C, the maximum heat loss from pool water evaporation is 13.93 MW. This value exceeds the spent fuel pool heat load determined under the initial operating conditions of normal power operation, normal reactor shutdown and refueling, and unplanned full core unloading, taking into account the 80-hour post-subcritical fuel assembly decay cooling time. Therefore, the PTR system possesses sufficient replenishment capacity to compensate for pool water evaporation losses.

[0089] In one specific embodiment, the calculation of the maximum decay heat includes the following three items regarding the residual thermal power after reactor shutdown: the residual fission heat power (hereinafter referred to as "Item A"); the decay heat power of captured products U-239 and Np-239 (hereinafter referred to as "Item B"); and the decay heat power of fission products and actinides other than U-239 and Np-239 (hereinafter referred to as "Item C"). During reactor shutdown, Item A originates from the effects of transient and delayed neutrons. The residual fission heat power has a significant impact on the residual core thermal power within 50 seconds after shutdown, but after 600 seconds, Item A becomes negligible compared to Items B and C. Therefore, when evaluating the cooling capacity of the spent fuel pool, only Items B and C are considered.

[0090] The thermal power in the B+C term depends on the amount of fission products and actinides stored in the core at the time of shutdown. The composition of nuclides in the core is primarily influenced by key fuel management parameters such as the enrichment of new fuel assemblies, the number of newly added assemblies, the enrichment and burnup of old fuel assemblies, and the power history of the fuel assemblies. Therefore, the impact of changes in fuel management on the calculation results must be considered in the calculation of decay heat.

[0091] For conservative reasons, the final decay heat calculation result is determined using the envelope method, taking the maximum decay heat at the same shutdown time during the first cycle, subsequent balancing cycles, and flexible cycles. For example, the same assumptions as in the core decay heat calculation report for the flexible refueling project can be used to finally calculate the decay heat result for item B+C.

[0092] In one specific embodiment, the calculation of the maximum heat load of the spent fuel pool considers filling the pool as much as possible and analyzes the maximum heat load under normal power operation, normal overhaul unloading, and abnormal conditions. Under normal power operation, the heat load of the spent fuel pool consists of two parts: the decay heat power of N batches of normally unloaded spent fuel assemblies stored in the pool; and the decay heat power of the (N+1)th batch of unloaded spent fuel after a shutdown day D1, where N is the number of spent fuel batches that can be stored in the pool. Under normal overhaul unloading, the heat load of the spent fuel pool consists of two parts: the decay heat power of N batches of normally unloaded spent fuel assemblies stored in the pool; and the core decay heat power after a shutdown day D2 at the end of the fuel cycle life. Under abnormal operating conditions, the heat load of the spent fuel pool consists of three parts: the decay heat power of the N-1 batches of normally discharged spent fuel assemblies stored in the spent fuel pool; the decay heat of the Nth batch of discharged spent fuel assemblies stored in the spent fuel pool after a shutdown of D3 days; and the total core decay heat power after a shutdown of D4 days due to abnormal reasons. Based on the above operating condition definitions, the design reference heat load of the spent fuel pool under different operating conditions can be calculated by summing the decay heat of the fuel assemblies in the spent fuel pool.

[0093] In one specific embodiment, regarding the analysis of the cooling capacity of the spent fuel pool, under power operation conditions, since the heat load of the spent fuel pool under normal power operation conditions is determined by the time interval (D1) from the reactor shutdown during refueling overhaul to the completion of core reloading, and under power operation conditions, the PTR system should be able to maintain the average temperature of the spent fuel pool not exceeding 50°C in the single pump single refueling operation mode, the heat load of the spent fuel pool at the moment of completion of refueling of the nuclear power plant can be calculated for different values ​​of D1, and the RRI supply water temperature required to meet the requirement that the pool water temperature does not exceed 50°C under this heat load.

[0094] Under normal overhaul unloading conditions, the spent fuel pool heat load is determined by the time interval (D2) from reactor shutdown to unloading completion. The PTR system design basis requires that the PTR system maintain an average spent fuel pool temperature not exceeding 50°C under normal overhaul unloading conditions. The spent fuel pool heat load at the moment of unloading completion was calculated for different values ​​of D2, and the maximum RRI supply water temperature required to maintain a pool water temperature not exceeding 50°C under different PTR system operating modes was also calculated.

[0095] Under unplanned full-core unloading conditions, the heat load of the spent fuel pool is determined by the time interval (D3) between the unit's overhaul shutdown in the previous cycle and the shutdown due to an abnormal reason during the current cycle's power operation, and the unplanned full-core unloading time (D4). The PTR system design basis requires that the PTR system can maintain the average temperature of the spent fuel pool not exceeding 60°C under unplanned full-core unloading conditions.

[0096] Further, in an optional embodiment, step S40 includes:

[0097] Step S41: Obtain pre-stored component modeling information, wherein the component modeling information is generated by pre-modeling a single fuel assembly based on the structure of the fuel assembly;

[0098] Step S42: Obtain pre-stored area modeling information, wherein the area modeling information is generated by pre-modeling a specific area based on the layout information of fuel assemblies stored in the spent fuel pool, the loading and unloading workflow information, and the structural information around the fuel plant and transfer channel.

[0099] Step S43: Based on the component modeling information and area modeling information, calculate the radiation dose rate at specific locations around the fuel plant and transfer channel, and determine whether it meets the corresponding radiation zoning requirements. Also, establish a calculation model for the radiation dose rate at the operating platform during the operation of the mobile fuel assembly, and calculate the minimum shielding water layer thickness and the minimum distance between the fuel assembly and the transfer gate required to meet the radiation zoning requirements based on the radiation dose rate limit at the operating platform.

[0100] In this embodiment, radiation shielding during fuel loading and unloading is also a direct limiting factor for the post-subcritical fuel assembly decay cooling time. This embodiment performs radiation shielding calculations and evaluations for two scenarios: when no fuel assembly movement is performed and when the fuel assembly is moved.

[0101] When no fuel assembly movement is performed, radioactivity primarily originates from the fuel assemblies stationary at the bottom of the spent fuel pool. Therefore, the dose rate around the fuel plant and transfer routes was recalculated. The results show that the dose rate above the surface of the spent fuel pool is extremely low when no fuel assembly movement is performed. Reducing the fuel assembly decay cooling time to 80 hours has a limited impact on the dose rate around the fuel plant and transfer routes, as well as on the operating platform where fuel operators are located; these areas still meet the requirements of the original radiation zoning.

[0102] When moving fuel assemblies, the primary concern is the radioactivity released by the moving assemblies. To evaluate radiation shielding during fuel assembly movement, this embodiment analyzes the minimum thickness of the shielding water layer. Calculations show that the minimum shielding water layer thickness required to maintain the green zone above the water surface during fuel assembly movement is 270 cm. Meeting this requirement ensures that the bottom of the moving fuel assembly remains 93.2 cm below the top of the fuel assemblies stored at the bottom of the pool, guaranteeing that the movement will not affect the assemblies stored at the bottom. Therefore, the current water depth in the spent fuel pool meets the required shielding water layer thickness.

[0103] Furthermore, considering that if the fuel assembly is moved too close to the transfer gate, the dose rate at the operating platform above the transfer gate may exceed the green zone limit, this embodiment also analyzed and calculated the minimum distance requirement between the assembly and the transfer gate. The results show that, during fuel assembly movement, to maintain the dose rate at the operating platform elevation +21.0m within the green zone limit, the minimum distance required between the assembly and the transfer gate is 210cm. The current spent fuel pool design meets this requirement.

[0104] Regarding this embodiment, it should also be noted that the radiation field is not uniform in any working area of ​​a nuclear power plant. The effective dose rate used in the shielding calculation is the maximum possible radiation field within the area accessible to personnel at the work location, or the effective dose rate at the contact point during equipment maintenance and repair. In other words, the effective dose rate used in the shielding calculation is relatively conservative. Nevertheless, considering the uncertainty of the source term calculation and the uncertainty of the neutron and photon transport calculation, a certain safety factor is taken into account for the effective dose rate level in the shielding calculation.

[0105] First, based on the structure of the fuel assembly, it is divided into three parts along its height: the upper end, the active section, and the lower end. By assigning material properties and gamma source terms to each of these three parts, a model of the individual fuel assembly is completed. Further, considering the arrangement of fuel assemblies stored in the spent fuel pool and the surrounding structures of the fuel plant and transfer channels, the entire area of ​​interest is modeled. Finally, dose rates are calculated at specific locations around the fuel plant and transfer channels to evaluate compliance with radiation zoning requirements. The minimum shielding water layer thickness and the minimum distance between the assembly and the transfer gate are given for meeting radiation zoning requirements at the operating platform during the operation of the moving fuel assembly.

[0106] To shorten the fuel assembly decay cooling time, for example to 80 hours, the dose rate around the fuel plant and transfer channels and the dose rate during the movement of fuel assemblies were calculated and compared with the case of 96 hours of fuel assembly decay cooling (the shortest time under the rapid unloading scenario considered in the original radiation shielding design report). The radiation dose rate received by fuel operators still meets the power plant's dose rate limit requirements for this area.

[0107] In one specific embodiment, the dose rate evaluation around the fuel plant and transfer channel can be carried out by selecting a total of 20 dose points located above the water surface of the spent fuel pool, outside the wall of the spent fuel pool, and outside the transfer gate and loading gate to calculate the dose rate around the fuel plant and transfer channel.

[0108] In one specific embodiment, regarding the minimum shielding water layer thickness requirement, it should be noted that since the dose rate above the spent fuel pool surface is extremely low when the fuel assembly is not being moved, and the dose rate above the spent fuel pool surface increases with the thickness of the shielding water layer above the fuel assembly during relocation, the minimum shielding water layer thickness required to maintain the area above the water surface in a green zone is calculated. Specifically, the selected dose points include those directly above the assembly and those offset from the assembly's central axis by 20cm to 90cm, encompassing the peak dose rate point above the moved assembly. In the height direction, two height planes, 5cm and 100cm above the spent fuel pool surface, are selected. According to calculations, the minimum shielding water layer thickness required to maintain the area above the water surface in a green zone (dose rate <25μSv / h) during fuel assembly relocation is 270cm. The required shielding water layer thickness increases slightly after shortening the decay cooling time, but does not exceed 10cm.

[0109] Regarding the minimum distance requirement between the fuel assembly and the transfer gate, if the assembly is too close to the transfer gate during fuel assembly movement, the dose rate at the operating platform above the transfer gate may exceed the green zone limit. Therefore, it is necessary to calculate the minimum distance between the assembly and the transfer gate required to maintain the operating platform elevation +21.0m within the green zone during fuel assembly movement. According to the calculation, the minimum distance required between the assembly and the transfer gate to maintain the dose rate at the operating platform elevation +21.0m within the green zone limit during fuel assembly movement is 210cm. Shortening the decay cooling time increases the minimum distance between the assembly and the transfer gate by 10cm. The current spent fuel pool design meets this requirement.

[0110] Furthermore, in an optional embodiment, after optimizing the subcritical post-fuel assembly decay cooling time, the method further includes: modifying the corresponding power plant documents based on the optimized subcritical post-fuel assembly decay cooling time. In this embodiment, power plant documents such as technical specifications and related operating procedures are reviewed and then adaptively modified.

[0111] The present invention also constructs an optimization system for the decay cooling time of fuel assemblies after subcritical reactors. The optimization system includes a processor and a memory storing a computer program. When the processor executes the computer program, it implements the steps of the optimization method for the decay cooling time of fuel assemblies after subcritical reactors described above.

[0112] The above description is merely a preferred embodiment of the present invention and is not intended to limit the invention. Various modifications and variations can be made to the present invention by those skilled in the art. Any modifications, equivalent substitutions, improvements, etc., made within the spirit and principles of the present invention should be included within the scope of the claims of the present invention.

Claims

1. A method for optimizing the decay cooling time of fuel assemblies after subcritical reactors, characterized in that, include: Step S10: Obtain the shortened time to start fuel operation after shutdown to be determined; Step S20: Based on the time of fuel operation after reactor shutdown, analyze the radioactive consequences of the fuel operation accident through modeling and determine whether the first preset condition is met. Step S30: Based on the time for refueling to resume after reactor shutdown, analyze the cooling capacity of the spent fuel pool by modeling and determine whether the second preset condition is met. Step S40: Based on the time for refueling to begin after reactor shutdown, analyze the radiation shielding of the fuel plant through modeling and determine whether the third preset condition is met. Step S50: When the first preset condition, the second preset condition, and the third preset condition are all satisfied, the time after the reactor shutdown to start fuel operation is taken as the subcritical post-fuel assembly decay cooling time.

2. The optimization method according to claim 1, characterized in that, Step S20 includes: Calculate the amount of fission products accumulated in the largest burnup component based on the time after the reactor shutdown resumes fuel operation. Obtain a computational model pre-established based on each stage of the radioactive material migration path, wherein the computational model is used to describe the diffusion behavior and impact process of the radioactive material; Based on the calculation model and the amount of fission products accumulated in the maximum burnup component, the radioactive consequences of fuel handling accidents are analyzed, and it is determined whether the preset environmental radioactivity conditions are met. Based on the calculation model and the accumulation of fission products of the maximum fuel consumption component, the personnel dosage in the main control room of the fuel operation accident is analyzed, and it is determined whether the preset conditions for the main control room to remain are met.

3. The optimization method according to claim 2, characterized in that, The calculation of the fission product accumulation of the maximum burnup component includes: The nuclide density at each time step was calculated using the Bateman differential equation; The accumulation of fission products in the maximum burnup component during a fuel handling accident is calculated by simulating the nuclide density change process.

4. The optimization method according to claim 1, characterized in that, Step S30 includes: Calculate the maximum decay heat of the fuel assembly based on the unit's fuel management information; Based on the design information of the spent fuel pool, the unit fuel management information, and the maximum decay heat of the fuel assemblies, the maximum heat load of the spent fuel pool under different operating conditions is determined. Based on the time to start fuel operation after reactor shutdown, the maximum decay heat of the fuel assembly, and the maximum heat load of the spent fuel pool, the cooling capacity of the spent fuel pool under different operating conditions is analyzed, and it is determined whether the requirements for removing the decay heat of the fuel assembly are met.

5. The optimization method according to claim 4, characterized in that, The operating conditions include: power operation, normal overhaul unloading, unplanned full core unloading, design basis accident, and beyond design basis accident.

6. The optimization method according to claim 1, characterized in that, Step S40 includes: Obtain pre-stored component modeling information, wherein the component modeling information is generated by pre-modeling a single fuel assembly based on the structure of the fuel assembly; Obtain pre-stored area modeling information, wherein the area modeling information is generated by pre-modeling a specific area based on the layout information of fuel assemblies stored in the spent fuel pool, the loading and unloading workflow information, and the structural information around the fuel plant and transfer channel; Based on the component modeling information and area modeling information, the radiation dose rate at specific locations around the fuel plant and transfer channel is calculated, and it is determined whether the conditions meet the corresponding radiation zoning requirements. In addition, a calculation model for the radiation dose rate at the operating platform during the operation of the mobile fuel assembly is established, and the minimum shielding water layer thickness and the minimum distance between the fuel assembly and the transfer gate are calculated based on the radiation dose rate limit at the operating platform to meet the radiation zoning requirements.

7. The optimization method according to claim 6, characterized in that, Specific locations around the fuel plant and transfer routes include: above the water surface of the spent fuel pool, outside the walls of the spent fuel pool, outside the transfer gate, and outside the loading gate.

8. The optimization method according to claim 6, characterized in that, The minimum thickness of the shielding water layer is 270cm; the minimum distance between the fuel assembly and the transfer gate is 210cm.

9. The optimization method according to claim 1, characterized in that, Also includes: Based on the optimized post-subcritical fuel assembly decay cooling time, the corresponding power plant documents were modified.

10. A system for optimizing the decay cooling time of fuel assemblies after subcritical reactor conditions, comprising a processor and a memory storing a computer program, characterized in that, When the processor executes the computer program, it implements the steps of the method for optimizing the decay cooling time of fuel assemblies after subcritical reactors as described in any one of claims 1-9.

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