Method for determining the rate of temperature reduction of reactor water during reactor shutdown

Abstract

The present invention provides a method for determining the rate of cooling of reactor water during reactor shutdown in a nuclear power plant, which can suppress the rise in radioactivity concentration of reactor water even when the rate of cooling of reactor water is high during reactor shutdown. [Solution] The method for determining the rate of cooling of reactor water during reactor shutdown according to the present invention comprises: a first step S21 of obtaining the feedwater iron concentration during operation of the nuclear power plant; a second step S22 of obtaining information on the fuel assembly replacement history; a third step S23 of estimating the amount of fuel cladding, which is the amount of cladding attached to the surface of the fuel rods, using the feedwater iron concentration and the replacement history; and a fourth step S24 to S26 of determining the rate of cooling of reactor water during reactor shutdown based on the estimated amount of fuel cladding.

Description

[Technical Field]

[0001] The present invention relates to a method for determining the rate of temperature reduction of reactor water when a reactor is shut down in a nuclear power plant. [Background technology]

[0002] Periodic inspections of nuclear power plants are conducted after shutting down the reactor, which is operating at high temperature and pressure, and returning it to a normal temperature and pressure state. During this shutdown operation, it is known that some of the radioactive nuclides attached to the surface of the fuel rods are released into the reactor water, causing an increase in the radioactivity concentration of the reactor water. The radioactive nuclides released into the reactor water adhere to the surfaces of pipes and other components, causing an increase in the surrounding radiation levels.

[0003] Patent Document 1 describes an example of a method for reducing radiation dose during reactor shutdown. In the shutdown method for a boiling water reactor described in Patent Document 1, the starting temperature of the residual heat removal system for carbon steel piping is set to 130°C or lower. This lowers the temperature at which radioactive nuclides released into the reactor water react on the surface of the carbon steel piping, thereby suppressing the incorporation of radioactive nuclides into the surface oxide film.

[0004] Furthermore, the relationship between the rate at which the reactor water temperature decreases during shutdown operations and the radioactivity of the reactor water is also being investigated. From the perspective of the integrity of the reactor materials, the rate at which the reactor water temperature decreases during shutdown is required to be below 55°C / h.

[0005] Non-Patent Document 1 evaluates the reactor water cooling rate and reactor water radioactivity in an actual reactor, and shows that a low cooling rate is effective in maintaining a low concentration of reactor water radioactivity. Non-Patent Document 2 reports that reducing the reactor water cooling rate suppresses the detachment of cladding (a general term for solid corrosion products that do not dissolve in reactor water) containing radioactive materials attached to the fuel surface from the fuel surface.

[0006] On the other hand, in recent years, in order to improve the operating rate of nuclear power plants, studies have been conducted to shorten the period of scheduled maintenance, and studies have been conducted to increase the rate at which the reactor water cools down in order to shorten the time until the reactor is shut down. For example, if the rate of cooling down from the rated temperature of 280°C to the cold shutdown temperature of 50°C is set at 20°C / h, the time until shutdown is 11.5 hours. In contrast, if the rate of cooling down is 40°C / h, the time until shutdown is 5.75 hours, which is 5.75 hours shorter than when the rate of cooling down is 20°C / h.

[0007] Patent documents 2 and 3 describe examples of mass balance models for reactor water quality. [Prior art documents] [Patent Documents]

[0008] [Patent Document 1] Japanese Patent Application Publication No. 9-230088 [Patent Document 2] Japanese Patent Application Publication No. 01-063894 [Patent Document 3] Japanese Patent Application Publication No. 06-289179 [Non-patent literature]

[0009] [Non-Patent Document 1] Water chemistry for nuclear reactor systems 4, BNES, London, 1987, P.55-P.58, “Investigation on transport and activation of corrosion products in the 2 X 1300 Mwe twin boiling water reactors of Gundremmingen” [Non-Patent Document 2] Water chemistry for nuclear reactor systems 5, BNES, London, 1989, P.315-P.316, “Data survey on radionuclides during BWR shutdown”

Summary of the Invention

Problems to be Solved by the Invention

[0010] Conventionally, when increasing the cooling rate of reactor water during reactor shutdown, it has been considered that the radioactive concentration of reactor water increases, and radionuclides adhere to the surfaces of equipment in the system where reactor water circulates, increasing the dose rate, which is a concern leading to an increase in the exposure dose during subsequent regular inspections. However, in current nuclear power plants that maintain a lower feed water iron concentration (the concentration of iron in the feed water supplied to the reactor) compared to the 1980s when the data of Non-Patent Document 1 and Non-Patent Document 2 were obtained, the relationship between the cooling rate of reactor water during reactor shutdown and the radioactive concentration of reactor water is not necessarily clear. Therefore, in current nuclear power plants, even if the cooling rate of reactor water is increased during reactor shutdown to shorten the regular inspection period, there is a possibility of suppressing the increase in the radioactive concentration of reactor water.

[0011] An object of the present invention is to provide a method for determining the cooling rate of reactor water during reactor shutdown in a nuclear power plant that can suppress an increase in the radioactive concentration of reactor water even when the cooling rate of reactor water is increased during reactor shutdown.

Means for Solving the Problems

[0012] The method for determining the cooling rate of reactor water during reactor shutdown according to the present invention includes a first step of obtaining the feed water iron concentration during the operation of a nuclear power plant, a second step of obtaining information on the replacement history of fuel assemblies, a third step of estimating the amount of fuel cladding, which is the amount of cladding attached to the surface of fuel rods, using the feed water iron concentration and the replacement history, and a fourth step of determining the cooling rate of reactor water during reactor shutdown based on the estimated amount of fuel cladding.

Effects of the Invention

[0013] According to the present invention, a method for determining the rate of cooling of reactor water during reactor shutdown can be provided in a nuclear power plant, which can suppress the rise in radioactivity concentration of reactor water even when the rate of cooling of reactor water is high during reactor shutdown. [Brief explanation of the drawing]

[0014] [Figure 1] This figure shows an example of experimental results determining the relationship between fuel cladding volume and Co-60 separation rate for two different reactor water cooling rates. [Figure 2] This flowchart shows the procedure for determining the temperature reduction rate of furnace water according to Example 1 of the present invention. [Figure 3] This is a diagram showing the configuration of a BWR plant. [Figure 4A] This diagram shows an example of the fuel assembly replacement history, illustrating the number of fuel assemblies loaded into the core during each operating cycle of a BWR plant. [Figure 4B] This figure shows examples of the amount of iron brought into the reactor by feedwater during each operating cycle, and examples of the amount of fuel cladding in the fuel assemblies at the end of each operating cycle. [Figure 4C] This figure shows an example of the fuel cladding amount of a fuel assembly at the start of each operating cycle. [Figure 5] This figure shows examples of the relationship between fuel cladding and Co-60 detachment for multiple reactor water temperature reduction rates obtained during past reactor shutdowns in actual reactors. [Figure 6] This diagram illustrates a mass balance model that describes the behavior of metal corrosion products being transferred to reactor water. [Modes for carrying out the invention]

[0015] In this invention, it is possible to determine the rate at which the temperature of the reactor water is lowered during reactor shutdown in a nuclear power plant, so as to suppress the rise in the radioactivity concentration of the reactor water even when the rate of temperature reduction of the reactor water is high during reactor shutdown. By increasing the rate of temperature reduction of the reactor water during reactor shutdown, for example, the period of periodic inspections of nuclear power plants can be shortened.

[0016] In this invention, the amount of fuel cladding, which is the amount of cladding attached to the surface of the fuel rods, is estimated using the feedwater iron concentration during operation of the nuclear power plant and the fuel assembly replacement history. Based on the estimated amount of fuel cladding, the rate of reactor water cooling during reactor shutdown is determined. For example, to determine the rate of cooling, the relationship between the amount of fuel cladding and the rate or amount of Co-60 detachment from the fuel rods can be used for multiple rates of cooling.

[0017] First, we will explain the general behavior of radionuclides in the cooling water of a boiling water reactor (BWR) using Figure 3.

[0018] Figure 3 is a diagram showing the configuration of a BWR plant.

[0019] Reactor 1 has a reactor core 3, composed of multiple nuclear fuel assemblies, inside a reactor pressure vessel 2 (hereinafter referred to as RPV2). Reactor water supplied to the core 3 by recirculation pumps 4 and jet pumps 5 is heated by the heat generated by the nuclear fission of the nuclear fuel material, and a portion of it turns into steam. This steam is guided to the turbine 7 through the main steam pipe 6 and rotates the turbine 7. The steam discharged from the turbine 7 is condensed back into water in the condenser 8. This water is supplied to the reactor 1 as feedwater.

[0020] To suppress the generation of radioactive corrosion products within the RPV2, the feedwater is purified by a condensate purification device 9 (filtration and demineralization device) located downstream of the condenser 8, which primarily removes metallic impurities. Corrosion products that form the basis of radioactive corrosion products are generated on the surfaces where BWR plant components, such as the RPV2 and recirculation system piping 10, come into contact with reactor water. For this reason, corrosion-resistant steels such as stainless steel and nickel-based alloys, which have low corrosion properties, are used for the main components of the primary system. In addition, the RPV2, which is made of low-alloy steel, has a stainless steel overlay on its inner surface to prevent the low-alloy steel from coming into direct contact with the reactor water. Furthermore, a portion of the reactor water is purified by a reactor water purification device 11 (filtration and demineralization device) in the reactor water purification system, actively removing even small amounts of metallic impurities present in the reactor water.

[0021] However, even with the corrosion countermeasures described above, it is unavoidable that minute amounts of metallic impurities will be present in the reactor water. As a result, some of these metallic impurities adhere to the surface of the fuel rods contained in the fuel assemblies as metallic oxides. Some of the impurities (e.g., metallic elements) adhering to the surface of the fuel rods undergo nuclear reactions when irradiated with neutrons released by the nuclear fission of the nuclear fuel material within the fuel rods, becoming radioactive nuclides such as cobalt-60, cobalt-58, chromium-51, and manganese-54.

[0022] These radionuclides, mostly in the form of metal oxides, remain attached to the surface of the fuel rods contained within the fuel assemblies. However, some radionuclides dissolve into the reactor water as ions, depending on the solubility of the incorporated oxides, or are re-released into the reactor water as an insoluble solid of metal oxides called cladding. Radioactive materials in the reactor water are removed by a reactor water purification system connected to RPV2.

[0023] Radioactive materials not removed by the reactor water purification system accumulate on the surfaces of BWR plant components (e.g., piping) that come into contact with the reactor water as they circulate with the reactor water through recirculation systems and other means. As a result, radiation is emitted from the surfaces of BWR plant components, causing radiation exposure to workers performing routine inspections.

[0024] Furthermore, when reducing the output of a BWR plant to achieve a cold shutdown, control rods are inserted between the fuel assemblies in core 3 to stop the fission reaction of the nuclear fuel. Simultaneously, the residual heat from the nuclear fuel is used to evaporate the reactor water, removing it as latent heat of vaporization. The reactor water is then guided to a residual heat removal system branched off from the recirculation system for heat exchange and cooling. The following describes this procedure.

[0025] (1) When the reactor water temperature reaches the rated 280°C, control rods are inserted to stop the fission reaction of the nuclear fuel.

[0026] (2) When the reactor water temperature drops from 280°C to about 150°C, the reactor pressure is high and a large amount of steam is generated. Therefore, simultaneously with the insertion of the control rods, steam is sent from the main steam pipe 6 to the condenser 8 to cool the fuel assemblies by the heat of vaporization. After the turbine 7 is disconnected, the turbine bypass valve 17 of the turbine bypass pipe 12 is gradually opened to send steam to the condenser 8 to cool the fuel assemblies.

[0027] (3) Before starting operation of the residual heat removal system, a warming operation is performed to gradually raise the system temperature to approach the reactor water temperature in order to relieve thermal stress on the equipment and piping and maintain their integrity.

[0028] (4) When the reactor water temperature is below 150°C, the cooling efficiency due to the heat of vaporization decreases. Therefore, the reactor water is guided to the residual heat removal system heat exchanger 15 using residual heat removal system piping 13 and residual heat removal system pump 14, which branch off from the recirculation system, to cool it. The cooling rate at this time is adjusted by controlling the flow rate of water flowing through the residual heat removal system heat exchanger 15 and the flow rate of the bypass piping 16.

[0029] Conventionally, in order to suppress the dose rate of the residual heat removal system (RHR system), it has been considered effective to suppress the peeling off of the cladding containing radioactive materials adhering to the surface of the fuel rods by lowering the rate of cooling of the reactor water, as disclosed in Non-Patent Documents 1 and 2, thereby suppressing the increase in the concentration of radioactive cladding in the reactor water. Cladding is a general term for solid corrosion products that do not dissolve in reactor water.

[0030] However, lowering the rate of cooling of the reactor water has the disadvantage of taking longer to shut down the reactor, delaying the start of scheduled inspections, lengthening the inspection period, and thus lowering the operating rate of the BWR plant. For this reason, it is desirable to keep the rate of cooling as high as possible without affecting the radioactivity concentration of the reactor water at the time of reactor shutdown.

[0031] Therefore, considering the operating conditions of BWR plants in the 1980s, when the data disclosed in Non-Patent Documents 1 and 2 were obtained, and the current operating conditions, the inventors hypothesized that, given the improved water quality of feedwater during operation, the amount of cladding adhering to the surface of the fuel rods (fuel cladding amount) is now less, and that this reduction in fuel cladding amount may affect the detachment of the cladding from the surface of the fuel rods when the reactor water temperature is lowered. The inventors then conducted experiments to investigate how the fuel cladding amount and the rate of cooling of the reactor water affect the detachment of the fuel cladding when the reactor is shut down.

[0032] In the following, the cladding adhering to the surface of a fuel rod will be referred to as "fuel cladding," and the amount of cladding adhering to the surface of a fuel rod will be referred to as "fuel cladding volume." Fuel cladding mainly contains iron oxides and also contains radioactive materials.

[0033] The inventors conducted an experiment simulating reactor shutdown using an experimental loop that could simulate the temperature and pressure of the reactor water and the boiling conditions on the fuel rods. In this experiment, they investigated the relationship between the amount of fuel cladding, which correlates with the feedwater iron concentration (concentration of iron in the feedwater), the rate of cooling of the reactor water, and the rate of separation of radionuclides contained in the fuel cladding from the fuel rods. The inventors considered using the results of this experiment to determine a method for determining the rate of cooling of the reactor water. Generally, a higher feedwater iron concentration is associated with a larger amount of fuel cladding.

[0034] A fuel cladding tube made of Zircaloy-2 was used in the experiment. This fuel cladding tube was attached to a heater pin and immersed in a hematite suspension. The outer diameter of the heater pin was in contact with the inner diameter of the fuel cladding tube. The hematite suspension was prepared by heating and aging a 3% aqueous solution of iron(III) nitrate notahydrate. After immersion of the fuel cladding tube, an electric current was applied to the heater pin to induce boiling on the surface of the Zircaloy-2, causing the hematite in the suspension to adhere to the surface of the Zircaloy-2. In this way, a hematite simulating fuel cladding was applied at a concentration of 100 μg / cm³. 2 From 3000 μg / cm³ 2 A test specimen with the attached material was prepared. This specimen is a fuel cladding tube made of Zircaloy-2 with hematite attached, simulating a fuel rod with fuel cladding attached to its surface.

[0035] Next, the heater pin with the fuel cladding (test specimen) attached was mounted in a high-temperature water loop test apparatus equipped with an autoclave capable of handling high-temperature water at 280°C and 7 MPa, which are reactor water conditions. The high-temperature water loop was started, and an aqueous solution containing Co-60 was injected into the 280°C circulating water flowing into the autoclave. By energizing the heater pin, boiling was induced, causing the Co-60 injected into the high-temperature circulating water to adhere to the hematite (simulated fuel cladding) on ​​the test specimen.

[0036] After continuing to operate the high-temperature water loop for approximately 100 hours in this state, the temperature was lowered and the test specimen was removed. The amount of Co-60 adhering to the test specimen was determined by gamma-ray spectroscopy using a Ge semiconductor detector. Subsequently, the test specimen with the quantified amount of Co-60 adhering was reattached to the heater pin, and this heater pin was placed in an autoclave. The high-temperature water loop was operated at 280°C, and after reaching 280°C, the heater pin was energized.

[0037] Subsequently, the reactor water was operated at a cooling rate of 20°C / h until the temperature of the high-temperature water loop reached 60°C. While the reactor water temperature was being lowered, the pressure-holding valve of the high-temperature water loop was adjusted to reach 100°C, which corresponds to the saturation vapor pressure of the high-temperature water loop temperature, thereby inducing depressurized boiling. After the temperature of the high-temperature water loop reached 60°C, the test specimen was removed, and the amount of Co-60 remaining in the specimen was measured. The weight of the specimen was also measured to determine the amount of hematite remaining in the specimen (amount of simulated fuel cladding).

[0038] Next, the same experiment was conducted, but with the reactor water temperature reduction rate changed to 40°C / h.

[0039] From the experimental results obtained in this way, the relationship between the amount of hematite remaining in the test specimen and the Co-60 exfoliation rate from the test specimen was determined. Hereafter, the Co-60 exfoliation rate from the test specimen will be referred to as the Co-60 exfoliation rate. The amount of Co-60 exfoliated from the test specimen will be referred to as the Co-60 exfoliation amount. In actual BWR plants, the Co-60 exfoliation rate is the rate at which Co-60 is exfoliated from the fuel rods to the reactor water, and the Co-60 exfoliation amount is the amount of Co-60 exfoliated from the fuel rods to the reactor water.

[0040] The amount of hematite remaining on the test specimen corresponds to the amount of fuel cladding, which is the amount of cladding attached to the surface of the fuel rod. The Co-60 detachment rate is determined by subtracting the amount of Co-60 remaining on the test specimen from the amount of Co-60 attached to the test specimen, and then dividing the determined amount of Co-60 detached from the test specimen by the amount of Co-60 attached to the test specimen. Changes in the Co-60 detachment rate correspond to changes in the radioactivity concentration of the reactor water.

[0041] Figure 1 shows an example of experimental results showing the relationship between fuel cladding amount and Co-60 delamination rate for two different reactor water cooling rates. Figure 1 shows an example of results obtained in an experiment simulating reactor shutdown, as described above. The two cooling rates are 20°C / h and 40°C / h. In Figure 1, the horizontal axis represents the fuel cladding amount per unit area, i.e., the amount of hematite remaining on the test specimen per unit area, and the vertical axis represents the Co-60 delamination rate.

[0042] In the following, the amount of fuel cladding per unit area may simply be referred to as the amount of fuel cladding.

[0043] Figure 1 shows that the fuel cladding amount is 1000 μg / cm³. 2 From the above, as disclosed in Non-Patent Documents 1 and 2, it can be seen that a low rate of reactor water cooling results in a low Co-60 separation rate. On the other hand, a fuel cladding amount of 1000 μg / cm³ is observed. 2 When the value is smaller, it can be seen that the Co-60 detachment rate does not depend on the reactor water cooling rate.

[0044] The fact that a lower rate of reactor water cooling results in a lower Co-60 delamination rate suggests that in older plants prior to the improvement and standardization, the feedwater iron concentration was high and the fuel cladding volume was large, thus reflecting the conventional trend that a lower rate of reactor water cooling leads to less fuel cladding delamination. Furthermore, the fact that the Co-60 delamination rate does not depend on the reactor water cooling rate when the feedwater iron concentration is low and the fuel cladding volume is small suggests that in plants after the improvement and standardization, where it is possible to operate with a low feedwater iron concentration, a high rate of cooling during reactor shutdown does not necessarily lead to increased fuel cladding delamination.

[0045] However, in the experiment, the fuel cladding formation time was shorter than in an actual BWR plant in order to obtain results in a short time, and the flow velocity around the test specimen was slower than in an actual plant. Therefore, the conditions in an actual plant could not be fully reproduced. For this reason, the Co-60 separation rate relative to the fuel cladding amount may differ between the results obtained in the experiment and those obtained in an actual plant.

[0046] However, as shown in Figure 1, the behavior in which the Co-60 separation rate initially decreases as the amount of fuel cladding increases, then remains almost constant, and then increases again as the amount of fuel cladding increases further, is considered to be the same in experiments and in actual reactors. Furthermore, in the fuel cladding range where the Co-60 separation rate decreases as the amount of fuel cladding increases, and in the fuel cladding range where the Co-60 separation rate does not change even when the amount of fuel cladding increases, the reactor water cooling rate is considered to have no effect on the Co-60 separation rate. Conversely, in the fuel cladding range where the Co-60 separation rate increases as the amount of fuel cladding increases, the reactor water cooling rate is considered to have an effect on the Co-60 separation rate.

[0047] From the above, the reactor water cooling rate during reactor shutdown can be determined by first estimating the amount of fuel cladding, and then determining whether the estimated amount of fuel cladding is a value at which the Co-60 separation rate changes depending on the cooling rate, based on the relationship between the amount of fuel cladding and the Co-60 separation rate for multiple reactor water cooling rates (for example, the experimental results shown in Figure 1). Alternatively, whether the estimated amount of fuel cladding is a value at which the Co-60 separation rate changes depending on the cooling rate can be determined not by experimental results, but by using data obtained from past reactor shutdowns at the actual plant (for example, reactor water cooling rate, Co-60 separation rate, and fuel cladding amount). In other words, the reactor water cooling rate during reactor shutdown can be determined based on the relationship between the amount of fuel cladding and the Co-60 separation rate (amount of separation) obtained in advance from experimental results or results from the actual plant for multiple cooling rates.

[0048] To estimate the amount of fuel cladding, the amount of iron in the feedwater (the amount of iron brought into the reactor), which can be calculated from the feedwater iron concentration (the concentration of iron in the feedwater) and feedwater flow rate, and the fuel assembly replacement history are used. The fuel assembly replacement history refers to, for example, the number of fuel assemblies loaded and the number of assemblies replaced in each operating cycle.

[0049] This section outlines the method for estimating fuel cladding volume. Specific examples of this method are shown in the examples described below.

[0050] In the operation of a BWR plant, the concentration of metal impurities in the feedwater and reactor water is measured to control the quality of the reactor water. Of these concentrations, the iron concentration in the feedwater is used to estimate the amount of fuel cladding. This is because the concentration of iron in the feedwater is higher than that of other metal elements.

[0051] Because the feedwater has a neutral pH, the iron in it is dissolved at a low concentration and exists as insoluble iron cladding. Of the iron cladding introduced into the reactor water along with the feedwater, some is removed by the reactor water purification system, some adheres to the surfaces of in-reactor equipment such as pressure vessels and piping, but the majority of the iron cladding precipitates as fuel cladding at the points where bubbles are generated due to boiling on the surface of the fuel rods.

[0052] Therefore, the amount of fuel cladding can be roughly estimated by multiplying the feedwater iron concentration by the feedwater flow rate and operating time, and then dividing the result by the total surface area of ​​the fuel rods. Furthermore, each time a fuel assembly is inspected, a portion of it is replaced with a new one, and the remainder is reloaded. For this reason, the fuel assembly replacement history must also be considered when estimating the amount of fuel cladding.

[0053] In the method for determining the reactor water cooling rate according to this embodiment, the cooling rate can be set high within a range that does not increase the Co-60 peeling rate from the fuel rod surface when the reactor is shut down. Therefore, even with a high cooling rate, the increase in the radioactivity concentration of the reactor water can be suppressed. For this reason, for example, the period until the reactor is shut down can be shortened while suppressing the increase in the dose rate of primary system equipment, and the period of regular inspections can be shortened while suppressing the increase in the radiation dose to workers during regular inspection work.

[0054] The following describes, with reference to the drawings, a method for determining the rate of temperature reduction of reactor water during reactor shutdown, according to embodiments of the present invention. As an example, the application of the present invention to a boiling water reactor (BWR) plant will be described. [Examples]

[0055] This invention describes a method for determining the rate of temperature reduction of reactor water during reactor shutdown, according to Embodiment 1 of the present invention.

[0056] First, we will explain the schematic configuration of a BWR plant to which the method for determining the reactor water temperature reduction rate according to this embodiment is applied, using Figure 3.

[0057] Figure 3 is a diagram showing the configuration of a BWR plant, as already explained. A BWR plant comprises a reactor 1, a turbine 7, a condenser 8, a recirculation system, a feedwater system, a reactor water purification system, and a residual heat removal system. Furthermore, a BWR plant is equipped with a reactor shutdown temperature and pressure control device 40.

[0058] The reactor shutdown temperature and pressure control device 40 executes the method for determining the rate of reactor water cooling during reactor shutdown according to this embodiment, and also controls, for example, the turbine bypass valve 17, the residual heat removal system pump 14, the main steam isolation valve 27, and the reactor water purification system pump 24, and switches the residual heat removal system.

[0059] The reactor 1, installed inside the reactor containment vessel, has an RPV2 containing a reactor core 3. A jet pump 5 is installed inside the RPV2. Multiple fuel assemblies (not shown) are loaded into the reactor core 3. Each fuel assembly comprises multiple fuel rods filled with multiple fuel pellets made of nuclear fuel material. The recirculation system includes a recirculation pump 4 and stainless steel recirculation system piping 10, with the recirculation pump 4 installed in the recirculation system piping 10.

[0060] The water supply system includes a water supply pipe 18 connecting the condenser 8 and the RPV2, a condensate pump 19, a condensate purification device 9, a low-pressure feedwater heater 20, a feedwater pump 21, a high-pressure feedwater heater 22, and a bypass pipe 23. The condensate pump 19, condensate purification device 9, low-pressure feedwater heater 20, feedwater pump 21, and high-pressure feedwater heater 22 are installed in this order on the water supply pipe 18 from the condenser 8 towards the RPV2. The bypass pipe 23 is connected to the water supply pipe 18 and bypasses the condensate purification device 9.

[0061] The reactor water purification system comprises reactor water purification system piping 29 connecting the recirculation system piping 10 and the feedwater piping 18, a reactor water purification system pump 24, a regenerative heat exchanger 25, a non-regenerative heat exchanger 26, and a reactor water purification device 11. The reactor water purification system pump 24, the regenerative heat exchanger 25, the non-regenerative heat exchanger 26, and the reactor water purification device 11 are installed in the reactor water purification system piping 29. The reactor water purification system piping 29 is connected to the recirculation system piping 10 upstream of the recirculation pump 4.

[0062] The cooling water inside the RPV2 is pressurized by the recirculation pump 4 and ejected through the recirculation system piping 10 into the bell mouth (not shown) of the jet pump 5 from its nozzle (not shown). The reactor water present around this nozzle is also drawn into the bell mouth by the action of the ejected flow from the nozzle.

[0063] The reactor water discharged from the jet pump 5 is supplied to the reactor core 3 and heated by the heat generated by the nuclear fission of the nuclear fuel material in the fuel rods. Some of the heated reactor water turns into steam. This steam is led from the RPV 2 through the main steam pipe 6 and the main steam isolation valve 27 to the turbine 7, which rotates the turbine 7. When the turbine 7 rotates, a generator (not shown) connected to the turbine 7 rotates, generating electricity. The steam discharged from the turbine 7 is condensed into water in the condenser 8.

[0064] This water is supplied as feedwater to the RPV2 through the feedwater pipe 18. The feedwater flowing through the feedwater pipe 18 is pressurized by the condensate pump 19, impurities are removed by the condensate purification device 9, it is further pressurized by the feedwater pump 21, and heated by the low-pressure feedwater heater 20 and the high-pressure feedwater heater 22. The extracted steam extracted from the main steam pipe 6 and the turbine 7 through the extraction pipe 28 is supplied to the low-pressure feedwater heater 20 and the high-pressure feedwater heater 22, respectively, and serves as a heat source for the feedwater.

[0065] The residual heat removal system (RHR system) is a system that connects the recirculation system and RPV2, and removes residual heat from the reactor core when the reactor is shut down. The residual heat removal system comprises residual heat removal system piping 13 made of carbon steel, residual heat removal system heat exchanger 15, residual heat removal system pump 14, and bypass piping 16.

[0066] As shown in Figure 3, one end of the residual heat removal system piping 13 is connected to the recirculation system piping 10 upstream of the recirculation pump 4. The other end of the residual heat removal system piping 13 branches downstream of the recirculation pump 4 into a pipe connected to the recirculation system piping 10 and a pipe connected to the RPV2. The residual heat removal system piping 13 connected to the RPV2 is connected to a core spray sparger (not shown) within the core shroud installed inside the RPV2.

[0067] The core spray sparger is located above the core and at the upper end of the core shroud, and is equipped with multiple core spray nozzles (not shown). The core spray sparger and core spray nozzles are part of the components that make up the high-pressure spray system.

[0068] During the shutdown of a BWR plant, control rods are inserted between the fuel assemblies to stop the nuclear fission chain reaction. In addition, while the reactor water continues to generate heat from the decay of the fuel during the initial stages of the shutdown, it evaporates on the surface of the fuel rods as it does during operation, and the generated steam is sent to the turbine 7 via the main steam piping 6 and returned to water in the condenser 8. When the amount of heat generated from decay decreases and the amount of steam generated decreases, making it difficult to maintain the rated reactor pressure, the turbine 7 is disconnected and power generation is stopped.

[0069] Subsequently, the steam flow rate is adjusted by adjusting the opening of the turbine bypass valve 17. For example, as the reactor water temperature decreases, the opening of the turbine bypass valve 17 is gradually increased. As the reactor water temperature decreases, the decrease in reactor pressure also gradually decreases. Therefore, in order to maintain a constant rate of decrease in reactor water temperature, it is necessary to gradually increase the rate of change in the opening of the turbine bypass valve 17. Such operation of the turbine bypass valve 17 requires the skill of the operator.

[0070] When the reactor water temperature reaches 130°C or below, the residual heat removal system pump 14 is activated, and water is passed through the residual heat removal system piping 13 to the residual heat removal system heat exchanger 15 to cool the reactor water.

[0071] Here, we will explain the method for determining the rate of temperature reduction of the furnace water according to this embodiment.

[0072] Figure 2 is a flowchart showing the procedure for determining the reactor water temperature reduction rate according to this embodiment.

[0073] To determine the rate of cooling of reactor water during reactor shutdown, it is necessary to estimate the amount of cladding attached to the surface of the fuel rods in the reactor core (fuel cladding amount). To estimate the fuel cladding amount, information on the feedwater iron concentration (iron concentration in feedwater) during operation of the BWR plant and the fuel assembly replacement history is obtained.

[0074] Furthermore, the relationship between fuel cladding amount and Co-60 separation rate (for example, Figure 1) is determined in advance for multiple cooling rates. Based on this relationship, it is determined whether the estimated fuel cladding amount is a value in which the Co-60 separation rate changes depending on the cooling rate. In this embodiment, the relationship between fuel cladding amount and Co-60 separation rate is determined in advance for two cooling rates from data obtained in experiments simulating reactor shutdown, as described above.

[0075] In S21, the feedwater iron concentration is obtained during operation of the BWR plant. The feedwater iron concentration is measured during the operation of the BWR plant.

[0076] In S22, information on the fuel assembly replacement history is acquired. When fuel assemblies are replaced during routine inspections, the arrangement of the fuel assemblies to be removed, the fuel assemblies to be reused, and the arrangement of the new fuel assemblies are determined based on the burnup of the fuel assemblies and the neutron distribution in the reactor. This data is acquired retrospectively as information on the fuel assembly replacement history.

[0077] In S23, the amount of fuel cladding is estimated based on the feedwater iron concentration and the fuel assembly replacement history.

[0078] In S24, it is determined whether the estimated fuel cladding amount is such that the Co-60 separation rate changes depending on the cooling rate, based on the relationship between the fuel cladding amount and the Co-60 separation rate (for example, Figure 1). For example, it is determined whether the Co-60 separation rate changes when the cooling rate is increased for the estimated fuel cladding amount.

[0079] If the change in the Co-60 separation rate when the cooling rate is increased is less than a predetermined threshold for the estimated fuel cladding amount, it is determined that the Co-60 separation rate does not change with respect to the cooling rate, and the process proceeds to S25. If the change in the Co-60 separation rate when the cooling rate is increased for the estimated fuel cladding amount is greater than or equal to this threshold, it is determined that the Co-60 separation rate does change with respect to the cooling rate, and the process proceeds to S26.

[0080] In S25, the highest of several cooling rates used to determine the relationship between fuel cladding volume and Co-60 delamination rate (e.g., Figure 1) is determined as the reactor water cooling rate during reactor shutdown. For example, if the estimated fuel cladding volume is 1150 μg / cm³ 2 In this case, since the Co-60 peeling rate can be considered to not change with the cooling rate, the higher cooling rate of 40°C / h is determined.

[0081] In S26, the lowest of the multiple cooling rates used to determine the relationship between fuel cladding amount and Co-60 delamination rate (e.g., Figure 1) is determined as the reactor water cooling rate during reactor shutdown. For example, if the estimated fuel cladding amount is 2000 μg / cm³ 2 In this case, the Co-60 peeling rate changes depending on the rate of cooling (as the rate of cooling increases, the rate of Co-60 peeling increases), so the lower rate of cooling is determined to be 20°C / h.

[0082] Here, the method for estimating the fuel cladding amount in S23 of Figure 2 will be explained using Figures 4A to 4C. In Figures 4A to 4C, each cell in the table corresponds to the others.

[0083] Figure 4A is a diagram illustrating an example of fuel assembly replacement history, showing an example of the number of fuel assemblies loaded into the core during each operating cycle of a BWR plant. In the example shown in Figure 4A, 800 fuel assemblies are loaded into the core for one operating cycle. Figure 4A shows 10 operating cycles, from the 1st to the 10th cycle.

[0084] In Figure 4A, the vertical rows represent the operating cycles, and the horizontal columns indicate which cycle a fuel assembly originated from. In other words, Figure 4A shows how many fuel assemblies were loaded from which operating cycle for each of the 10 operating cycles, from the 1st to the 10th cycle.

[0085] For example, in the first cycle, there are 800 fuel assemblies loaded from the first cycle. In the second cycle, there are 500 fuel assemblies loaded from the first cycle and 300 fuel assemblies loaded from the second cycle. In the third cycle, there are 300 fuel assemblies loaded from the first cycle, 300 fuel assemblies loaded from the second cycle, and 200 fuel assemblies loaded from the third cycle.

[0086] Figure 4B shows examples of the amount of iron introduced into the reactor by feedwater during each operating cycle, and the amount of iron adhering to the fuel assemblies at the end of each operating cycle, i.e., the amount of fuel cladding on the fuel assemblies at the end of each operating cycle. The units of the values ​​shown in Figure 4B are kg.

[0087] Figure 4C shows an example of the fuel cladding amount of a fuel assembly at the start of each operating cycle. The units of the values ​​shown in Figure 4C are kg.

[0088] As shown in Figure 4A, let's assume that 800 fuel assemblies are loaded into the reactor core during the first cycle at the start of operation. All 800 fuel assemblies are brand new and have no fuel cladding attached. Therefore, the amount of fuel cladding in the fuel assemblies at the start of the first cycle is 0.0 kg, as shown in Figure 4C.

[0089] As shown in Figure 4B, it is estimated that 30 kg of iron was introduced into the reactor by feedwater during the first cycle. This amount of iron can be calculated by multiplying the feedwater iron concentration by the feedwater flow rate and the operating time of one cycle. It is assumed that this 30 kg of fuel cladding was evenly distributed among 800 fuel assemblies. That is, at the end of the first cycle, as shown in Figure 4B, the amount of fuel cladding is 30 kg.

[0090] Next, we consider that 300 of the 800 fuel assemblies that have completed the first cycle have been removed, and 300 new fuel assemblies have been loaded, and the system is ready to operate the second cycle. That is, as shown in Figure 4A, in the second cycle there are 500 fuel assemblies loaded from the first cycle and 300 new fuel assemblies loaded from the second cycle.

[0091] The 500 fuel assemblies that underwent the first cycle had 18.8 kg of fuel cladding attached to them (500 / 800 of the 30 kg total fuel cladding). Therefore, the amount of fuel cladding in the fuel assemblies at the start of the second cycle is 18.8 kg, as shown in Figure 4C.

[0092] In this state, the second cycle is operated, and the amount of iron brought into the furnace by feedwater during the second cycle is assumed to be 26 kg, as shown in Figure 4B.

[0093] As a result, of the 500 fuel assemblies that experienced both the first and second cycles, 16.2 kg (500 / 800) of the 26 kg of iron was deposited during the second cycle, and when combined with the 18.8 kg deposited during the first cycle, a total of 35.0 kg (= 18.8 kg + 16.2 kg) of iron was deposited (Figure 4B). Of the 300 fuel assemblies loaded from the second cycle onward, the remaining 9.8 kg (300 / 800) of the 26 kg of iron was deposited during the second cycle (Figure 4B).

[0094] Next, we consider that 200 of the 500 fuel assemblies that have experienced the first and second cycles are removed, and 200 new fuel assemblies are loaded, and the system is ready to operate the third cycle. The 300 fuel assemblies loaded from the second cycle are reloaded as is. In other words, as shown in Figure 4A, in the third cycle, there are 300 fuel assemblies loaded from the first cycle, 300 fuel assemblies loaded from the second cycle, and 200 new fuel assemblies loaded from the third cycle.

[0095] The 300 fuel assemblies loaded from the first cycle had 21.0 kg of fuel cladding attached to them (300 / 500 of the 35 kg total). The 300 fuel assemblies loaded from the second cycle had 9.8 kg of fuel cladding attached to them. Therefore, as shown in Figure 4C, the amount of fuel cladding in the fuel assemblies at the start of the third cycle was 21.0 kg for the fuel assemblies loaded from the first cycle and 9.8 kg for the fuel assemblies loaded from the second cycle.

[0096] In this state, the third cycle is operated, and the amount of iron brought into the furnace by feedwater during the third cycle is assumed to be 33 kg, as shown in Figure 4B.

[0097] As a result, 12.4 kg (300 / 800) of the 33 kg of iron attached to the 300 fuel assemblies loaded from the first cycle during the third cycle, bringing the total amount of iron attached to 33.4 kg (= 21.0 kg + 12.4 kg) (Figure 4B). 12.4 kg (= 33 kg * 300 / 800) of iron attached to the 300 fuel assemblies loaded from the second cycle during the third cycle, bringing the total amount of iron attached to 22.2 kg (= 9.8 kg + 12.4 kg) (Figure 4B). 200 fuel assemblies loaded from the third cycle during the third cycle, with 8.3 kg (= 33 kg * 200 / 800) of iron attached (Figure 4B).

[0098] In this embodiment, as described above, the amount of iron brought into the reactor by feed water is determined from the feed water iron concentration. Assuming that this amount of iron adheres to the surface of the fuel rods, by considering the replacement history of the fuel assemblies (for example, the number of loaded fuel assemblies and the number of replaced fuel assemblies in each operating cycle), the amount of fuel cladding in the fuel assemblies at the end of each operating cycle can be estimated.

[0099] The amount of fuel cladding per unit area shown in FIG. 1 can be obtained by dividing the amount of fuel cladding in the fuel assembly by the surface area per fuel assembly (the product of the total surface area of the fuel rods per fuel assembly and the number of fuel assemblies). For example, assume that the fuel cladding tube is made of zircaloy-2, with an outer diameter of 1 cm and a length of 4 m. Also, assume that the fuel assembly has a configuration of 9×9 fuel rods, and about 5% (4 out of them) are replaced by water rods, and it is composed of 77 fuel rods. Then, the surface area per fuel assembly is about 96800 cm 2 Therefore, for example, the amount of fuel cladding per unit area of the fuel assemblies loaded from the first cycle at the end of the third cycle is 33.4 kg of iron adhering to 300 fuel assemblies (FIGS. 4A, 4B), so 33.4 kg / (96800 cm 2 / fuel assembly × 300 fuel assemblies) ≈ 1150 μg / cm 2 and can be estimated.

[0100] As described above, the amount of fuel cladding can be estimated in the process of S23 in FIG. 2.

[0101] In the processes from S24 to S26 in FIG. 2, the cooling rate of the reactor water is determined using the amount of fuel cladding (the amount of fuel cladding per unit area) estimated in the process of S23 and the relationship between the amount of fuel cladding and the Co-6o peeling rate for a plurality of pre-determined cooling rates (for example, FIG. 1).

[0102] The estimated amount of fuel cladding is 1150 μg / cm 2Therefore, if experimental results are obtained as shown in Figure 1 regarding the relationship between fuel cladding amount and Co-60 delamination rate for multiple cooling rates, the reactor water cooling rate is determined as follows. From Figure 1, the fuel cladding amount is 1150 μg / cm³. 2 Therefore, it can be concluded that there is almost no difference in the Co-60 detachment rate between a cooling rate of 20°C / h and 40°C / h, meaning that the Co-60 detachment rate does not change with respect to the cooling rate (S24 in Figure 2). In this case, the higher cooling rate, 40°C / h, is determined to be the cooling rate of the reactor water when the reactor is shut down (S25 in Figure 2). The cooling rate can be adjusted to 40°C / h by adjusting the opening of the turbine bypass valve 17.

[0103] In this embodiment, the rate of cooling of the reactor water during reactor shutdown can be determined as described above. Even if the rate of cooling of the reactor water during reactor shutdown is high, the increase in the radioactivity concentration of the reactor water can be suppressed, thus shortening the time until low-temperature shutdown and reducing the period of regular maintenance. [Examples]

[0104] This section describes a method for determining the reactor water temperature reduction rate during reactor shutdown according to Embodiment 2 of the present invention. In this embodiment, the relationship between the fuel cladding amount and the Co-60 detachment amount for multiple temperature reduction rates is determined in advance from data obtained from an actual plant (a real BWR plant), and this determined relationship is used to determine the reactor water temperature reduction rate during reactor shutdown.

[0105] In Example 1, we described an example (Figure 1) in which the relationship between fuel cladding amount and Co-60 detachment rate for multiple cooling rates was obtained experimentally. In this example, we will describe an example in which the relationship between fuel cladding amount and Co-60 detachment rate for multiple cooling rates can be obtained in an actual reactor. Specifically, we will use data on reactor water cooling rate, Co-60 detachment rate, and fuel cladding amount obtained during past reactor shutdowns in an actual reactor to determine the relationship between fuel cladding amount and Co-60 detachment rate for multiple cooling rates.

[0106] The amount of Co-60 detachment during reactor shutdown can be determined, for example, as follows: First, the change in the concentration of Co-60 in the reactor water during reactor shutdown is measured over time, for example, every hour. The amount of Co-60 removed is determined by assuming that the average concentration of Co-60 at consecutive measurement points is removed by the flow rate of the purification system. Co-60 may detach from the fuel and its concentration may increase. If the concentration of Co-60 increases, the amount of Co-60 that has detached from the fuel and remained in the reactor water is calculated from the increase in concentration and the amount of reactor water. Then, the amount of Co-60 removed and the amount of Co-60 that has detached from the fuel and remained in the reactor water are added together, and the resulting amount of Co-60 is taken as the amount of Co-60 detachment between consecutive measurement points. This amount of Co-60 detachment is accumulated over the reactor shutdown period, and the accumulated value is taken as the amount of Co-60 detachment during reactor shutdown.

[0107] In this embodiment, it is assumed that the fuel assembly replacement history, the amount of fuel cladding in the fuel assembly at the end of each operating cycle, and the amount of fuel cladding in the fuel assembly at the start of each operating cycle are as shown in Figures 4A, 4B, and 4C.

[0108] Figure 5 shows an example of the relationship between fuel cladding and Co-60 detachment for multiple reactor water cooling rates obtained during past reactor shutdowns in an actual plant. In Figure 5, the horizontal axis represents the fuel cladding per unit area, and the vertical axis represents the Co-60 detachment rate. The reactor water cooling rate is the average cooling rate at the end of each operating cycle. In the example shown in Figure 5, multiple cooling rates are divided into two categories: cooling rates of 20°C / h or less and cooling rates of 30°C / h or more.

[0109] As shown in Figure 5, the amount of Co-60 peeling is approximately 1250 μg / cm³ of fuel cladding. 2 The above results vary depending on the rate of cooling of the reactor water, and the fuel cladding amount is approximately 1000 μg / cm³. 2 The following shows that it does not depend on the rate of cooling. That is, the fuel cladding amount is approximately 1250 μg / cm³. 2 Based on the above, a higher cooling rate results in a larger amount of Co-60 delamination and a fuel cladding amount of approximately 1000 μg / cm³.2 In the following cases, the amount of Co-60 peeling does not increase even with a high rate of temperature reduction.

[0110] In a BWR plant where the relationship between fuel cladding and Co-60 detachment for multiple cooling rates, as shown in Figure 5, has been obtained, the amount of Co-60 detachment during reactor shutdown can be determined, for example, as follows. The amount of fuel cladding when the reactor is shut down can be estimated from the feedwater iron concentration obtained from the feedwater quality and the fuel assembly replacement history, etc., by calculations as shown in Figures 4A to 4C.

[0111] The estimated fuel cladding amount is, for example, 1250 μg / cm³. 2 In the above case, Figure 5 indicates that increasing the reactor water temperature reduction rate to 30°C / h or higher is likely to increase the amount of Co-60 detachment during reactor shutdown. Therefore, the lowest temperature reduction rate among the multiple rates used to determine the relationship between fuel cladding amount and Co-60 detachment rate (for example, Figure 5) is determined as the reactor water temperature reduction rate during reactor shutdown (S26 in Figure 2).

[0112] The estimated fuel cladding amount is, for example, 1000 μg / cm³. 2 In the following cases, as shown in Figure 5, the amount of Co-60 detachment differs little between cases where the reactor water temperature reduction rate is 30°C / h or higher and cases where it is 20°C / h or lower. Therefore, the highest temperature reduction rate among the multiple temperature reduction rates used to determine the relationship between fuel cladding amount and Co-60 detachment amount (for example, Figure 5) is determined as the reactor water temperature reduction rate at reactor shutdown (S25 in Figure 2).

[0113] In this embodiment, as described above, the rate of cooling of reactor water during reactor shutdown can be determined by utilizing data obtained during past reactor shutdowns at an actual plant. Therefore, even if the rate of cooling of reactor water is high during reactor shutdown, the increase in the radioactivity concentration of the reactor water can be suppressed, shortening the time until low-temperature shutdown and thus shortening the period of regular maintenance.

[0114] Furthermore, the relationship between fuel cladding volume and Co-60 detachment volume for multiple cooling rates, as shown in Figure 5, can be obtained from a single BWR plant. However, in actual plants, the difference in reactor water cooling rates between each reactor shutdown may be small, or the measurement interval for Co-60 concentration during reactor shutdown may be long, making it difficult to discern clear trends like those shown in Figure 5. For this reason, it may be advisable to collect data from multiple BWR plants of the same type and output, and then use this data to determine the relationship between fuel cladding volume and Co-60 detachment volume for multiple cooling rates. [Examples]

[0115] In Examples 1 and 2, the amount of fuel cladding at the end of the operating cycle was estimated based on the feedwater iron concentration and the fuel assembly replacement history. The amount of fuel cladding can also be estimated by other methods, for example, using a mass balance model of reactor water quality, including reactor water radioactivity.

[0116] This embodiment describes an example of estimating the fuel cladding amount using a mass balance model of reactor water quality. The mass balance model is described, for example, in Patent Document 2 (Japanese Patent Publication No. 01-063894) and Patent Document 3 (Japanese Patent Publication No. 06-289179). The outline of the mass balance model is described below.

[0117] Figure 6 illustrates a mass balance model that describes the behavior of metal corrosion products transferring to reactor water. This mass balance model is a physical model that describes the dynamic behavior of metal corrosion products contained in feedwater 61 and metal corrosion products generated as a result of corrosion of structural materials 63 inside and outside the reactor that are in contact with reactor water 62, by means of reactor water 62, such as adhering to the surface of fuel rods 64, re-adhering to the surface of structural materials 63 inside and outside the reactor, or being removed from the system by the reactor water purification system 65, using the macroscopic law of conservation of mass. The solid and dashed arrows in Figure 6 indicate transfer by ions and cladding, respectively.

[0118] The mass balance model of metal corrosion products in reactor water is described by the following system of differential equations, from Equation 1 to Equation 8.

[0119]

number

[0120]

number

[0121]

number

[0122]

number

[0123]

number

[0124]

number

[0125]

number

[0126]

number

[0127] In equations 1 through 8, the meanings of the variables and parameters are as follows: C: Concentration of metal corrosion products in furnace water (concentration of metal corrosion products in furnace water, e.g., concentration of iron, nickel, cobalt, etc.) t: time V: Amount of reactor water held F f :Water supply flow rate C f:Concentration of metal corrosion products in feedwater (concentration of metal corrosion products in feedwater) X: Occurrence rate of metal corrosion products generated by corrosion of internal furnace structural materials ζ: Dissolution or detachment constant of deposits on fuel rods ζ p 1 : Leaching or detachment constant of deposits attached to the internal structural materials of the furnace ζ p 2 : Leaching or detachment constant of deposits attached to the outer structural material M: Amount of metal corrosion products adhering to the fuel rod m1: Amount of non-radioactive metal corrosion products adhering to the surface of structural materials inside the furnace m2: Amount of non-radioactive metal corrosion products adhering to the surface of structural materials outside the furnace. δ: Adhesion constant to the fuel rod β: Removal rate in reactor water purification system δ p 1 : Adhesion constant to the internal structural material of the furnace δ p 2 : Adhesion constant to the outer structural material S1: Surface area of ​​structural materials inside the furnace S2: Surface area of ​​structural materials outside the furnace R: Concentration of radioactive metal corrosion products in reactor water (concentration of radioactive metal corrosion products in reactor water, e.g., concentrations of cobalt-60, cobalt-58, manganese-54, etc.) Y: Occurrence rate of radioactive metal corrosion products generated by corrosion of internal reactor structural materials. A: Amount of radioactive metal corrosion products adhering to fuel rods Γ1: Amount of radioactive metal corrosion products adhering to the surface of structural materials inside the reactor. Γ2: Amount of radioactive metal corrosion products adhering to the surface of structural materials outside the reactor. λ: Decay constant of radioactive metal corrosion products G: Production rate of radionuclides on fuel rods G1: Production rate of radionuclides on structural materials inside the reactor.

[0128] Of the above variables, C, C f F fR and Γ2 are state variables that can be measured during operation. M, A, and Γ1 are state variables that can be measured when the fuel rods are removed from the reactor during reactor shutdown, such as for periodic inspections. V, S1, and S2 are plant parameters specific to BWR plants. λ, G, and G1 are physical constants determined by the radioactive nuclides of the metal corrosion products. X, Y, ζ, ζ p 1 ζ p 2 δ, δ p 1 , δ p 2 β is, in principle, a model parameter. Note that m1 and m2 are state variables that are usually difficult to measure because it is impossible to distinguish between substances adhering from water (furnace water, feedwater) and those generated by the corrosion of structural materials.

[0129] During operation of a BWR plant, the concentration of corrosion products in the feedwater is C f It can be expressed as and the concentration of corrosion products in the furnace water can be expressed as C. In this example, the iron concentration in the feedwater is the concentration of corrosion products in the feedwater C f It is represented as follows.

[0130] Furthermore, the radioactivity concentration can be expressed as R, the fuel cladding amount (the amount of cladding attached to the surface of the fuel rods) as M, the amount of material attached to the equipment surface as Γ1, Γ2, m1, and m2, and the plant operating conditions as V, Ff, β, S1, and S2.

[0131] The amount of fuel cladding is determined by adjusting the model parameters so that the calculated values ​​of state variables such as C, M, and R on the left side of equations 1 to 8 match the measured values, and then using the adjusted model parameters, C f The predicted values ​​can be used as input and the state variables on the left side can be calculated to estimate the values. Therefore, in this embodiment, the feedwater iron concentration (the corrosion product concentration in the feedwater C) f ) contains C predicted by any method f We will obtain and use the predicted value.

[0132] In the mass balance model, the amount of metallic elements (metallic corrosion products) and radioactive nuclides adhering to the surface of the fuel rods is determined for each operating cycle (burnup period of the fuel assembly). Then, using the fuel assembly replacement history, the number of fuel assemblies replaced for each operating cycle is determined. Finally, the amount of metallic elements and radioactive nuclides adhering to the removed fuel is subtracted from the determined amount of metallic elements and radioactive nuclides to calculate the amount of metallic elements and radioactive nuclides adhering to the surface of the fuel rods in the next operating cycle.

[0133] When using this mass balance model to estimate the amount of fuel cladding at the end of the operating cycle, the concentration of corrosion products in the feedwater up to the end of the operating cycle C f Using the predicted values ​​as input, we determine the amount of metal corrosion products (M) for each element that adhere to the fuel rods at the end of the operating cycle. Then, we use the calculated M (amount of metal corrosion products adhering to the fuel rods) as the fuel cladding amount at the end of the operating cycle.

[0134] The rate at which reactor water cools during reactor shutdown can be determined based on this M (fuel cladding amount). For example, using the relationship between fuel cladding amount and Co-60 detachment rate shown in Figure 1 in Example 1, or the relationship between fuel cladding amount and Co-60 detachment amount shown in Figure 5 in Example 2, the rate at which reactor water cools down to suppress the increase in radioactivity concentration can be determined based on this M (fuel cladding amount).

[0135] As described above, in the embodiments of the present invention, the relationship between the amount of fuel cladding and the Co-60 detachment rate (or amount of detachment) for multiple cooling rates of reactor water during reactor shutdown (for example, Figures 1 and 5) is determined based on experimental results and data from actual reactors. The amount of fuel cladding during reactor shutdown is estimated using feedwater iron concentration and fuel assembly replacement history, or using a mass balance model of reactor water quality from reactor water quality and fuel assembly replacement history. Then, using the estimated amount of fuel cladding and the relationship between the amount of fuel cladding and the Co-60 detachment rate (or amount of detachment) for multiple cooling rates, the cooling rate of reactor water can be determined so as not to increase the Co-60 detachment rate (or amount of detachment). For this reason, in the embodiments of the present invention, it is possible to achieve both suppression of the rise in radioactivity concentration of reactor water and shortening of the shutdown period during reactor shutdown.

[0136] Examples 1 and 2 show how the relationship between fuel cladding and Co-60 detachment rate (or detachment amount) for two different cooling rates was used to determine the reactor water cooling rate (Figures 1 and 5). To determine the reactor water cooling rate, the relationship between fuel cladding and Co-60 detachment rate (or detachment amount) for three or more different cooling rates may also be used.

[0137] It should be noted that the present invention is not limited to the embodiments described above, and various modifications are possible. For example, the embodiments described above are explained in detail to make the present invention easier to understand, and the present invention is not necessarily limited to embodiments having all the configurations described. Furthermore, it is possible to replace parts of the configuration of one embodiment with the configuration of another embodiment. It is also possible to add configurations from other embodiments to the configuration of one embodiment. Furthermore, it is possible to delete parts of the configuration of each embodiment, or to add or replace other configurations. [Explanation of symbols]

[0138] 1...Reactor, 2...Reactor pressure vessel (RPV), 3...Core, 4...Recirculation pump, 5...Jet pump, 6...Main steam piping, 7...Turbine, 8...Condenser, 9...Condensate purification system, 10...Recirculation system piping, 11...Reactor water purification system, 12...Turbine bypass piping, 13...Residual heat removal system piping, 14...Residual heat removal system pump, 15...Residual heat removal system heat exchanger, 16...Bypass piping, 17...Turbine bypass valve, 18...Feed 19...Water piping, 20...Condensate pump, 21...Low-pressure feedwater heater, 22...Feedwater pump, 23...High-pressure feedwater heater, 24...Bypass piping, 25...Reactor water purification system pump, 26...Regenerative heat exchanger, 27...Non-regenerative heat exchanger, 28...Main steam isolation valve, 29...Extraction piping, 40...Reactor shutdown temperature and pressure control device, 61...Feedwater, 62...Reactor water, 63...In-core and out-of-core structural materials, 64...Fuel rods, 65...Reactor water purification system.

Claims

1. The first step in obtaining the iron concentration in the feedwater during operation of a nuclear power plant, The second step is to obtain information on the replacement history of fuel assemblies, A third step involves estimating the amount of fuel cladding, which is the amount of cladding attached to the surface of the fuel rod, using the feedwater iron concentration and the replacement history. A fourth step is to determine the rate of cooling of reactor water during reactor shutdown based on the estimated amount of fuel cladding, A method for determining the rate of temperature reduction of reactor water during reactor shutdown, characterized by comprising the following:

2. For multiple of the aforementioned cooling rates, the relationship between the amount of fuel cladding and the amount of Co-60 detachment from the fuel rods is determined in advance from data obtained during past reactor shutdowns at the nuclear power plant. In the fourth step, the rate of temperature reduction during reactor shutdown is determined based on the estimated amount of fuel cladding and the previously determined relationship. A method for determining the rate of temperature reduction of reactor water during reactor shutdown, as described in claim 1.

3. For multiple of the aforementioned cooling rates, the relationship between the amount of fuel cladding and the rate of Co-60 separation from the fuel rods is determined in advance from data obtained in experiments simulating reactor shutdown. In the fourth step, the rate of temperature reduction during reactor shutdown is determined based on the estimated amount of fuel cladding and the previously determined relationship. A method for determining the rate of temperature reduction of reactor water during reactor shutdown, as described in claim 1.

4. In the fourth step, if the estimated amount of fuel cladding is such that the amount of delamination does not change with respect to the cooling rate, the highest cooling rate among the plurality of cooling rates is determined to be the cooling rate at the time of reactor shutdown. The method for determining the rate of temperature reduction of reactor water during reactor shutdown, as described in claim 2.

5. In the fourth step, if the estimated fuel cladding amount is such that the delamination rate does not change with respect to the cooling rate, the highest cooling rate among the plurality of cooling rates is determined to be the cooling rate at reactor shutdown. The method for determining the rate of temperature reduction of reactor water during reactor shutdown, as described in claim 3.

6. In the third step, the amount of fuel cladding is estimated using the feedwater iron concentration, the exchange history, and a mass balance model that represents the behavior of the transfer of metal corrosion products into the reactor water. A method for determining the rate of temperature reduction of reactor water during reactor shutdown, as described in claim 1.

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