In-pile natural circulation test method for liquid metal cooled fast reactor

Abstract

Embodiments of the present application relate to the field of residual heat removal during shutdown, and particularly to a method for testing in-pile natural circulation of a liquid metal cooled fast reactor. The method provided by the embodiments of the present application determines an initial test power by using a natural circulation design program, and makes the fast reactor run at the initial test power, so as to obtain the running data of the in-pile natural circulation of the fast reactor under the premise of ensuring the safety of the fast reactor, check the natural circulation design program, and improve the reliability and accuracy of the evaluation result of the natural circulation capability of the fast reactor obtained by using the natural circulation design program. Meanwhile, the running data of the fast reactor at a low-power test power and a high-power test power are used to check the natural circulation design program, so as to verify the influence of temperature difference on the natural circulation, and further improve the reliability and accuracy of the evaluation result of the natural circulation capability of the fast reactor obtained by using the natural circulation design program.

Description

Technical Field

[0001] The embodiments of this application relate to the field of removing residual heat from reactor shutdown, specifically to a real-world natural circulation test method for a liquid metal-cooled fast reactor. Background Technology

[0002] The statements herein are provided merely as background information in connection with this application and do not necessarily constitute prior art.

[0003] After an emergency shutdown, the removal of residual heat from the reactor core is crucial to ensuring reactor safety. Natural circulation, as a method of removing residual heat without relying on external power, has become a commonly used method for removing residual heat from the core in liquid metal cooled fast reactors due to its high reliability.

[0004] Due to the complexity of the sodium-cooled fast reactor core structure, it is difficult to accurately assess its natural circulation thermal-hydraulic phenomena using systematic programs, and it is also difficult to conduct detailed off-core natural circulation design verification simulation tests through large-scale sodium bench tests.

[0005] In the design and verification of natural circulation in a real reactor, the design program (currently generally a system program) needs reliable data that can represent the natural circulation phenomenon of a real reactor for calculation reliability verification, while the natural circulation test of a real reactor relies on reliable program calculations for prediction and support.

[0006] In this context, conducting safe full-scale reactor natural circulation tests is of significant engineering importance. Firstly, these tests reveal the reactor's natural circulation capability. Sufficient capability enhances the reactor's safety and economics. Secondly, the natural circulation data obtained from these tests can be used to validate design procedures and subsequently applied to the design of similar reactors, improving accuracy, reliability, and efficiency, thereby enhancing the overall safety and economics of the reactor.

[0007] However, there are still many challenges in conducting real-world natural circulation tests of liquid metal-cooled fast reactors. Summary of the Invention

[0008] A brief overview of this application is provided below to offer a basic understanding of certain aspects thereof. It should be understood that this overview is not an exhaustive summary of the application. It is not intended to identify key or essential parts of the application, nor is it intended to limit its scope. Its purpose is merely to present certain concepts in a simplified form as a prelude to the more detailed description that follows.

[0009] To address the aforementioned problems, embodiments of this application provide a method for conducting a real-world natural circulation test of a liquid metal-cooled fast reactor, comprising: S1, determining an initial test power using a natural circulation design program to enable the fast reactor to establish a stable natural circulation within a preset time after pump shutdown while operating at the test power. The preset time is longer than the pump coasting time, and the initial test power is less than the core power corresponding to the residual core heat when the fast reactor is shut down at rated power; S2, operating the fast reactor at the test power and continuing to operate at the test power after pump shutdown until a stable natural circulation is established, acquiring operating data of the fast reactor after pump coasting stops, and using the operating data to verify the natural circulation design program; S3, based on the initial test power in step S1, gradually increasing the test power of the fast reactor and repeating step S2 until the test power of the fast reactor is increased to the level corresponding to the residual core heat when the fast reactor is shut down at rated power. S4. Determine the low-power test power that can create a temperature difference between the hot and cold pools of the fast reactor using the natural circulation design program verified in step S3; S5. Shut down the fast reactor and pumps at the low-power test power and obtain the operating data of the fast reactor after the pumps stop coasting, and use the operating data to verify the natural circulation design program; S6. Determine the high-power test power using the natural circulation design program verified in step S5, where the high-power test power is higher than the low-power test power; S7. Shut down the fast reactor and pumps at the high-power test power and obtain the operating data of the fast reactor after the pumps stop coasting, and use the operating data to verify the natural circulation design program; S8. If the high-power test power is less than the rated power of the fast reactor, evaluate the natural circulation capability of the fast reactor when shutting down and stopping the pumps at the rated power using the natural circulation design program verified in step S7.

[0010] The testing method provided in the embodiments of this application determines the initial test power by utilizing a natural circulation design program, enabling the fast reactor to establish a stable natural circulation within a preset time after pump shutdown while operating at the test power. This allows the fast reactor to operate in the actual reactor at the initial test power. Since the initial test power is less than the core power corresponding to the residual core heat when the fast reactor is shut down at rated power, it is beneficial to obtain the first operational data of the actual reactor's natural circulation while ensuring the safety of the fast reactor. This data is then used to verify the natural circulation design program, thereby improving the reliability and accuracy of the natural circulation design program.

[0011] The test power of the fast reactor is then gradually increased and step S2 is repeated until the test power is increased to the core power corresponding to the residual core heat when the fast reactor is shut down at rated power, or until the test power can no longer be increased. Since the fast reactor is run at that test power each time a new test power is determined, and the operating data of the fast reactor after the pump stops coasting is obtained, the operating data is used to verify the natural circulation design program. This allows the natural circulation design program to be continuously verified, thereby gradually improving its design accuracy.

[0012] Subsequently, by determining a low-power test power that can create a temperature difference between the hot and cold pools, the fast reactor is operated at this test power, and operating data is acquired after the pumps stop coasting. This operating data is used to verify the natural circulation design program, preliminarily verifying the impact of temperature difference on natural circulation. Then, the test power is increased to a high-power test power, and the operating data of the fast reactor at this high-power test power is used to verify the natural circulation design program, further verifying the impact of temperature difference on natural circulation and improving the reliability and accuracy of the natural circulation design program. When it is not possible to directly conduct a rated power emergency shutdown natural circulation test on the actual reactor, this embodiment uses actual reactor data combined with the verified natural circulation design program to calculate and evaluate the natural circulation capability of the fast reactor under rated power shutdown and pump cessation. The embodiments of this application achieve this by designing experiments that cover key phenomena of natural circulation operation after a full-power emergency shutdown, thereby ensuring that the obtained actual reactor data can reliably verify the design program and guaranteeing the reliability of the program's calculations for full-power emergency shutdown natural circulation operation. Attached Figure Description

[0013] Other objects and advantages of this application will become apparent from the following description of embodiments of this application with reference to the accompanying drawings, and will help to provide a comprehensive understanding of this application.

[0014] Figure 1 This is a schematic flowchart of the real-world natural circulation test method for a liquid metal-cooled fast reactor provided in an embodiment of this application.

[0015] Figure 2 The diagram shows the curves of the highest cladding temperature and the average core outlet temperature of a fast reactor after pump shutdown at 0.5% Pn (Pn is the rated power), determined using a natural circulation design program. It also shows the curve of the highest cladding temperature of the fast reactor under adiabatic conditions after pump stoppage at the given power value.

[0016] Figure 3 It is a calculated curve showing the change of the highest cladding temperature of a certain fast reactor over time under different pump coasting speeds, coolant flow rates, and different test powers.

[0017] It should be noted that the accompanying drawings are not necessarily drawn to scale, but are shown only in a schematic manner without affecting the reader's understanding. Detailed Implementation

[0018] Exemplary embodiments of this application will be described below with reference to the accompanying drawings. For clarity and brevity, not all features of actual implementations are described in the specification. However, it should be understood that many implementation-specific decisions must be made in the development of any such actual embodiment to achieve the developer's specific goals, such as complying with constraints related to the system and business, and these constraints may vary depending on the implementation. Furthermore, it should be understood that while development work can be very complex and time-consuming, such development work is merely a routine task for those skilled in the art who benefit from the content of this application.

[0019] It should also be noted that, in order to avoid obscuring this application with unnecessary details, only the equipment structure and / or processing steps closely related to the solution according to this application are shown in the accompanying drawings, while other details that are not closely related to this application are omitted.

[0020] The main challenges in conducting natural circulation tests in a live reactor are as follows: Fuel rod cladding temperature is the most critical parameter, but in a live reactor, the core outlet temperature can only be measured using temperature measuring devices (temperature measurement points) installed at the core outlet, not directly within the core. This is because the distance between the temperature measuring device at the core outlet and the hottest part of the cladding is typically more than one meter, and there is no one-to-one correspondence between the temperature measuring device and the fuel assembly. Furthermore, it is impossible to indirectly obtain the precise cladding temperature under natural circulation conditions through measurements of other parameters such as flow rate. In addition, the lack of representative data for validating live reactor natural circulation tests makes it difficult to ensure the reliability of the design program's predictions of the live reactor's natural circulation conditions, thus hindering the safety of the test implementation through program calculations. Moreover, the test must ensure that the obtained data reliably demonstrates the reactor's natural circulation capability; that is, the safety and effectiveness of the test implementation must be addressed.

[0021] Therefore, how to safely conduct real-world natural circulation tests of liquid metal-cooled fast reactors is an issue that needs to be addressed.

[0022] To address the aforementioned issues, embodiments of this application provide a method for conducting a real-world natural circulation test of a liquid metal-cooled fast reactor.

[0023] See Figure 1 , Figure 1 This is a schematic flowchart of a real-world natural circulation test method for a liquid metal-cooled fast reactor provided in an embodiment of this application, which may include steps S1 to S8.

[0024] S1. The initial test power is determined using the natural circulation design program so that the fast reactor can establish a stable natural circulation within a preset time after the pump is stopped while operating at the test power. The preset time is longer than the pump coasting time, and the initial test power is less than the core power corresponding to the residual core heat when the fast reactor is stopped at rated power.

[0025] S2. Run the fast reactor at the test power and then run it at the test power after the pump stops until a stable natural circulation is established. Obtain the operating data of the fast reactor after the pump stops coasting, and use the operating data to verify the natural circulation design program.

[0026] S3. Based on the initial test power in step S1, gradually increase the test power of the fast reactor and repeat step S2 until the test power of the fast reactor is increased to the core power corresponding to the residual core heat when the fast reactor is shut down at rated power, or until the test power can no longer be increased.

[0027] S4. Using the natural circulation design procedure verified in step S3, determine the low-power test power that can create a temperature difference between the hot and cold pools of the fast reactor.

[0028] S5. Shut down the fast reactor and pumps at a low power test power, and obtain the operating data of the fast reactor after the pumps stop coasting. Use the operating data to verify the natural circulation design program.

[0029] S6. Use the natural loop design program verified in step S5 to determine the high power test power, which is higher than the low power test power.

[0030] S7. Shut down the fast reactor and pumps at high power test power, and obtain the operating data of the fast reactor after the pumps stop coasting. Use the operating data to verify the natural circulation design program.

[0031] S8. If the high-power test power is less than the rated power of the fast reactor, the natural circulation design procedure verified in step S7 shall be used to evaluate the natural circulation capability of the fast reactor when shutting down and stopping the pumps at the rated power.

[0032] The testing method provided in the embodiments of this application is for existing fast reactors. By using a natural circulation design program to determine the initial test power, the fast reactor can establish a stable natural circulation within a preset time after pump shutdown while operating at the test power. The fast reactor is then put into actual operation at the initial test power. Since the initial test power is less than the core power corresponding to the residual core heat when the fast reactor is shut down at rated power, it is beneficial to obtain the actual natural circulation operation data of the fast reactor while ensuring the safety of the fast reactor. The natural circulation design program can be verified based on the actual natural circulation operation data of the fast reactor, thereby improving the reliability and accuracy of the natural circulation design program.

[0033] The test power of the fast reactor is then gradually increased and step S2 is repeated until the test power is increased to the core power corresponding to the residual core heat when the fast reactor is shut down at rated power, or until the test power can no longer be increased. Since the fast reactor is run at that test power each time a new test power is determined, and the operating data of the fast reactor after the pump stops coasting (i.e., natural circulation data) is obtained, the natural circulation design program is verified using the operating data. This allows the natural circulation design program to be continuously verified, thereby gradually improving the accuracy of its design.

[0034] Subsequently, by determining a low-power test power that can create a temperature difference between the hot and cold pools, the fast reactor was operated at this test power, and operating data was acquired after the pumps stopped coasting. This operating data was used to validate the natural circulation design program, preliminarily verifying the impact of temperature difference on natural circulation. Then, the test power was increased to a high-power test power, and the operating data from the fast reactor at this high-power test power was used to validate the natural circulation design program, further verifying the impact of temperature difference on natural circulation and improving the reliability and accuracy of the natural circulation design program. When it is not possible to directly conduct a rated power emergency shutdown natural circulation test on the actual reactor, this embodiment uses actual reactor data combined with the validated natural circulation design program to calculate and evaluate the natural circulation capability of the fast reactor under rated power shutdown and pump cessation.

[0035] The embodiments of this application achieve this by designing experiments to cover key phenomena under natural circulation conditions following a full-power emergency shutdown. This ensures that the obtained real-world reactor data can reliably verify the design program and guarantee the reliability of the program's calculations for the full-power emergency shutdown natural circulation condition. This provides important guidance and reference for developing implementation plans for fast reactor real-world natural circulation tests, improving the economy, safety, and reliability of subsequent reactors.

[0036] In step S2, because the test power is low, the core can continue to operate at the test power after the pump stops, thereby simulating the residual heat generated by the core to successfully establish natural circulation.

[0037] Liquid metal cooled fast reactors include, for example, liquid sodium cooled fast reactors (sodium-cooled fast reactors for short).

[0038] Unless otherwise specified, the pumps mentioned in this article refer to the primary loop pumps of the fast reactor.

[0039] In some embodiments, step S1 may include: S11, for any given power value, using a natural circulation design procedure to determine the changes in the average core outlet temperature and the maximum cladding temperature over time after the fast reactor stops pumping at the given power value, wherein the given power value is less than the core power corresponding to the residual core heat when the fast reactor stops at rated power; S12, determining the changes in the maximum cladding temperature over time under adiabatic conditions after the fast reactor stops coasting at the given power value; S13, based on the changes in the average core outlet temperature and the maximum cladding temperature over time determined in step S11 and the changes in the maximum cladding temperature over time determined in step S12, determining whether to use the given power value as the initial test power. In such embodiments, determining whether the given power value can be used as the initial test power through the above steps helps to make the determined initial test power more conservative, thereby ensuring the safety during actual reactor testing.

[0040] The embodiments of this application do not rely on measurement points and system program calculations as safety guarantees, but instead acquire the first batch of natural circulation phenomena and data. Simultaneously, they also consider ensuring the reliability of existing actual reactor data in verifying program calculations and measurement points when program calculation results and measurement point data are needed to ensure experimental safety, thereby guaranteeing the safety of the experimental implementation. Furthermore, when it is not possible to directly conduct rated power emergency shutdown natural circulation tests on an actual reactor, the embodiments of this application propose a scheme using actual reactor data combined with verified program calculations to illustrate the reactor's natural circulation capability. This scheme achieves this through experimental design that covers key phenomena of natural circulation conditions after a full-power emergency shutdown, thereby ensuring that the obtained actual reactor data can reliably verify the design program and guarantee the reliability of the program's calculations for full-power emergency shutdown natural circulation conditions.

[0041] In some embodiments, step S13 may include: S131, determining the relationship between the average core outlet temperature and the maximum cladding temperature under a given power value based on the changes in the average core outlet temperature over time and the changes in the maximum cladding temperature over time as determined in step S11, and the first duration t1 for the fast reactor to establish natural circulation under the given power value; S132, determining the second duration t2 corresponding to when the maximum cladding temperature reaches the conservative limit of the normal operation limit based on step S12, and if the second duration t2 is greater than the first duration t1, then the given power value is used as the initial test power; otherwise, the given power value is reselected.

[0042] In this embodiment, the relationship between the average core outlet temperature and the maximum cladding temperature allows the average core outlet temperature to reflect the maximum cladding temperature. This enables the determination of the maximum cladding temperature using the measured average core outlet temperature in subsequent steps, thus ensuring the safety of the actual reactor test. Furthermore, when the second duration t2 is determined to be greater than the first duration t1, it indicates that the fast reactor operating with pumps shut down at a given power value has a longer time to establish natural circulation. This allows for the measurement of corresponding natural circulation data, and therefore, this given power value can be used as the initial test power.

[0043] In some embodiments, the preset duration can be a second duration t2. The initial test power is less than or equal to 1% of the fast reactor's rated power, and low-power stable operation is used to simulate core residual heat.

[0044] See Figure 2 , Figure 2 The diagram shows the curves of the highest cladding temperature and the average core outlet temperature of a fast reactor after pump shutdown at 0.5% Pn, determined using a natural circulation design program. It also shows the curve of the highest cladding temperature of the fast reactor under adiabatic conditions after pump stoppage at a given power value. Curve A is the curve of the highest cladding temperature of the fast reactor after pump shutdown at 0.5% Pn, determined using a natural circulation design program. Curve B is the curve of the average core outlet temperature of the fast reactor after pump shutdown at 0.5% Pn, determined using a natural circulation design program.

[0045] As can be seen from curves A and B, when the fast reactor is operating stably at a given power value, after the pumps are stopped, both the average core outlet temperature and the maximum cladding temperature first enter a rapid rise phase (corresponding to A1 and B1 respectively), and then enter a stable phase (corresponding to A2 and B2 respectively). The time it takes for the maximum cladding temperature to transition from the rapid rise phase A1 to the stable phase A2 is shorter than the time it takes for the average core outlet temperature to transition from the rapid rise phase B1 to the stable phase B2. Furthermore, the maximum cladding temperature remains essentially constant after entering the stable phase A2, and the average core outlet temperature also remains essentially constant after entering the stable phase B2. Thus, after the fast reactor establishes natural circulation, there is essentially a stable difference between the maximum cladding temperature and the average core outlet temperature, allowing the maximum cladding temperature to be determined using the core measurement point temperatures.

[0046] The highest cladding temperature enters a stable phase A2 after approximately 580s. Therefore, the first duration t1 for establishing natural circulation in the fast reactor, determined using the natural circulation design program, is 580s.

[0047] The first duration t1 typically includes the sum of the pump coasting duration and the duration from the end of pump coasting until the highest cladding temperature reaches a stable phase A2. During pump coasting, the coolant in the fast reactor can still be forced to circulate by the pump. After pump coasting ends, the flow of coolant in the reactor depends entirely on natural circulation.

[0048] In some embodiments, in step S132, the second duration t2 can be determined by the following method: the pump shutdown duration during which the cladding temperature can be maintained without exceeding the normal operating limit, assuming all boundaries of the active section of the hottest component in the reactor are under adiabatic conditions after the pump stops coasting. In such embodiments, the second duration t2 can also be called the extremely conservative pump shutdown safety duration. During the second duration t2 of the fast reactor pump shutdown and operation at test power, the cladding temperature can be maintained without exceeding the normal operating limit. Therefore, by determining the initial test power based on the second duration determined in the above steps, it can be ensured that the cladding temperature of the fast reactor operating at the initial test power will also not exceed the normal operating limit, thereby ensuring the safety of the fast reactor operating at the initial test power.

[0049] The second duration t2 can be considered as the sum of the pump coasting duration and the duration during which all boundaries of the hottest active section of the reactor core can maintain adiabatic conditions after the pump coasting ends. The pump coasting duration is known. Since determining the duration during which all boundaries of the hottest active section of the reactor core can maintain adiabatic conditions after the pump coasting ends only involves the active section of a single fuel rod, a small-scale process, a detailed 3D CFD simulation is employed, combined with some manual calculations for verification (during the adiabatic condition stage). This ensures the reliability of the calculated second duration t2 and guarantees the safety of the experiment.

[0050] A suitable initial test power can be determined by the following approach: At this power, the second duration t2 is relatively long, with the duration maintained after coasting ending at least 100 seconds. Simultaneously, within this duration, the core is expected to establish stable natural circulation, meaning the measured core outlet temperature can move beyond the rapid rise phase and enter a stable phase (e.g., ...). Figure 2 (As shown). Generally, at low power, the time for natural circulation to establish up and the safe time for pump shutdown are both in the hundreds of seconds. When selecting an appropriate power, the former will be shorter than the latter.

[0051] Figure 2 Curve C in the figure shows the change of the highest cladding temperature over time under adiabatic conditions after the pump stops coasting at 0.5% Pn. Curve C shows that when the fast reactor is running stably at 0.5% Pn, the rate of increase in the highest cladding temperature is initially slow after the pump stops, but after a certain period, the rate of increase significantly increases, and the temperature rises rapidly.

[0052] It's easy to understand that when designing a reactor, a maximum cladding temperature limit is typically set based on the cladding material. For safety reasons, when determining the initial test power, a certain value can be further reduced from the maximum cladding temperature limit, thus using the normal operating limit of the cladding to determine the initial test power. For example, when the maximum cladding temperature limit is 800°C, the normal operating limit of the cladding can be 700°C. For curve C, when the maximum cladding temperature reaches 700°C, the corresponding second duration t2 is 680s. Since the second duration t2 of 680s is greater than the first duration t1 of 580s, this given power value of 0.5%Pn can be used as the initial test power.

[0053] exist Figure 2 In the process, after the highest cladding temperature reaches a stable stage A2 following the shutdown of the 0.5% Pn pump, the relationship between the average core outlet temperature and the highest cladding temperature is that the highest cladding temperature is always 80°C higher than the average core outlet temperature. When the normal operating limit for the cladding is 700°C, in order to ensure the safety of the fast reactor, the average core outlet temperature cannot exceed 620°C (i.e., 700°C - 80°C). A conservative limit of 590°C for the average core outlet temperature can be selected. That is, when the actual measured average core outlet temperature is less than 590°C, it indicates that the fast reactor is safe.

[0054] In some embodiments, step S2 may include: S21, running the fast reactor at an initial test power, and then running the fast reactor at the initial test power for a preset time after the pump is stopped, obtaining the operating data of the fast reactor within the preset time, using the operating data to determine whether the fast reactor has established a stable natural cycle, and using the operating data to verify the natural cycle design program; S22, if it is determined that the fast reactor has established a stable natural cycle, then stopping the reactor and starting the pump, and executing step S3. In such an embodiment, after performing a real-world test at the initial test power, it is possible to obtain the natural cycle data of the fast reactor after the pump is stopped, so as to use the natural cycle data to determine whether the fast reactor has established a stable natural cycle, and to verify the natural cycle design program, making the prediction of the natural cycle design program more accurate.

[0055] In some embodiments, in step S21, the obtained operating data of the fast reactor can be the actual measured change in the average core outlet temperature of the fast reactor over time. In step S22, it can be determined whether the fast reactor has established a stable natural cycle based on the rate of change of the actual measured average core outlet temperature of the fast reactor per unit time. Specifically, if the rate of change of the actual measured average core outlet temperature of the fast reactor per unit time is less than a threshold, it indicates that the fast reactor has established a stable natural cycle; if the rate of change of the actual measured average core outlet temperature of the fast reactor per unit time is greater than the threshold, it indicates that the fast reactor has not established a stable natural cycle. For example, when the measured average core outlet temperature changes by less than 1°C within 300 seconds, it is determined that the fast reactor has established a stable natural cycle.

[0056] In some embodiments, step S2 may further include: S23, if it is determined that the fast reactor has not established stable natural circulation, the reactor is immediately shut down and the pump is restarted after the preset pump shutdown time ends, and the changes in the average core outlet temperature and the maximum cladding temperature of the fast reactor over time after the preset time are determined according to the natural circulation design program verified in step S21; S24, based on the changes in the average core outlet temperature and the maximum cladding temperature over time determined in S23, the relationship between the average core outlet temperature and the maximum cladding temperature is determined; S25, based on the relationship between the average core outlet temperature and the maximum cladding temperature, the safe limit for the core outlet temperature is determined; S26, the fast reactor is operated at the initial test power, and the operating data of the fast reactor is obtained after the preset pump shutdown time. If the measured core outlet temperature is lower than the safe limit for the core outlet temperature, the pump is kept shut down until stable natural circulation is established, and then the reactor is shut down and the pump is restarted; S27, the operating data after the preset pump shutdown time obtained in step S26 is used to verify the natural circulation design program.

[0057] In this embodiment, if it is determined that the fast reactor has not established stable natural circulation after a preset pump shutdown time, the reactor must be shut down and the pump restarted to ensure reactor safety. Then, using a natural circulation design program that has been validated with natural circulation data, the changes in the average core outlet temperature and the maximum cladding temperature over time after the preset time at the test power are re-determined. The relationship between the average core outlet temperature and the maximum cladding temperature is then obtained, and the safe limit for the core outlet temperature is re-determined. The fast reactor is then operated again at the initial test power. Since the fast reactor operating data within the preset pump shutdown time has been obtained in step S21, the operating data can be measured again after the preset pump shutdown time in step S26. The operating data of the fast reactor includes the core outlet temperature measurement. In this actual reactor test, the core outlet temperature measurement can be used to ensure the safe operation of the fast reactor until a stable natural circulation is established, after which the reactor is shut down and the pumps are started. The operating data after the preset pump shutdown time is used to verify the natural circulation design program. Thus, at this initial test power, the natural circulation design program can be verified relatively completely, which is beneficial to improving the accuracy of the natural circulation design program in subsequent steps.

[0058] In some embodiments, if the measured core outlet temperature has exceeded the rapid rise phase B1 in curve B after a preset pump shutdown time, the test can be terminated by restarting the pump and shutting down the reactor once the measured core outlet temperature reaches a stable state (the rate of change of the average core outlet temperature per unit time is less than a threshold). If the measured core outlet temperature is still in the rapid rise phase, the test is immediately terminated by shutting down the reactor and restarting the pump. The operating data obtained from this test is then used to verify the natural circulation design program, and the test progress after extrapolating the preset time is calculated based on the natural circulation design program. Afterwards, to ensure the safety of the test after the preset time by establishing a defined relationship between the cladding and the core outlet temperature, the test is conducted again, and the test is terminated by shutting down the reactor and restarting the pump once the measured core outlet temperature reaches a stable state.

[0059] In some embodiments, step S3 may include: S31, determining that the fast reactor can remove the core heat by establishing a stable natural circulation at the initial test power, and determining a second test power based on the heat; S32, operating the fast reactor at the second test power, and operating the fast reactor at the second test power after pump shutdown until a stable natural circulation is established; acquiring the operating data of the fast reactor after the pump stops coasting, and using the operating data to verify the natural circulation design program; S33, determining that the natural circulation established by the fast reactor at the second test power can remove the core heat, and determining a third test power based on the heat; increasing the operating power of the fast reactor to the core power corresponding to the core residual heat when the fast reactor is shut down at rated power or until the test power can no longer be increased; using the operating data of the fast reactor after each pump stops coasting to verify the natural circulation design program.

[0060] In this embodiment, the subsequent test power is determined by using the heat removed from the reactor core by natural circulation at the previous test power. This allows for a gradual increase in test power while ensuring the safety of the fast reactor, and ensures the establishment of natural circulation for each test power until the test power reaches the residual heat at the time of fast reactor shutdown or the test power can no longer be increased using this method, at which point step S3 is completed. Because the natural circulation design program is validated using operating data after each pump stop coasting, its reliability and accuracy can be continuously improved. This facilitates accurate and reliable testing of the fast reactor's natural circulation capability, leading to accurate and reliable evaluation results.

[0061] In some embodiments, in steps S31 and S33, the core coolant flow rate is determined based on the fast reactor's operating power, the core outlet temperature measured after establishing stable natural circulation, and the core inlet temperature measured. The heat that the coolant can carry away is determined using the core coolant flow rate, and the corresponding core power is determined based on the core power corresponding to that heat. In such embodiments, determining the test power for each step through the above steps ensures the establishment of stable natural circulation, thereby guaranteeing the safety of the actual reactor test.

[0062] In some embodiments, in step S33, when the test power can no longer be increased, it means that the determined next test power is basically the same as the current test power. At this time, there is no need to repeat the test, and the actual stack test in step S3 can be stopped.

[0063] In some embodiments, a conservative core coolant flow rate is determined based on the operating power of the fast reactor, the core outlet measurement temperature after establishing stable natural circulation, and the core inlet measurement temperature. After determining the conservative core coolant flow rate, the amount of heat that the coolant can carry away at that flow rate is determined. Then, based on the amount of heat that the coolant can carry away at that flow rate, a conservative upper limit of the power that can be discharged at that flow rate is determined. This conservative upper limit of the power is the power for the next test.

[0064] In some embodiments, the average flow rate per fuel assembly monitored at each core outlet temperature measurement point can be determined using the following relationship: G = Q / [n(T1 + ΔT - T0) × cp], In the formula, G is the average flow rate per fuel assembly in n fuel assemblies monitored by the temperature measuring point at the core outlet, typically n is 3; Q is the total thermal power of the n fuel assemblies monitored by the temperature measuring point at the core outlet; T1 is the core outlet temperature measured by the temperature measuring point at the core outlet; T0 is the core inlet temperature measured by the temperature measuring point at the core inlet; n is the number of fuel assemblies monitored by the temperature measuring point at the in-core outlet, n can be 3; ΔT is the difference between the core outlet temperature measured by the temperature measuring point at the core outlet and the average temperature of the coolant at the outlet of the heated section of the monitored fuel assembly; cp is the specific heat, which is quickly calculated based on (T1+ΔT+T0) / 2.

[0065] ΔT can be obtained by calculating and conservatively taking the value based on the natural circulation design program (depending on the program model, it can also be conservatively replaced by the difference between the measured value and the average outlet temperature of the core). This difference is the data calculated solely by the natural circulation design program in step S3. At this point, the natural circulation program has been verified by real-world natural circulation data of similar phenomena in step S2, and the calculation of natural circulation phenomena at lower power levels is reliable.

[0066] Due to the central symmetry of the arrangement of the reactor core and the upper central measuring column, the lateral heat transfer caused by the flow between fuel assemblies has a relatively small impact on the temperature measurement points located at the center. Therefore, the in-core assemblies monitored by the temperature measurement points at the core outlet can be the assemblies located at the center, and T1 is the measured value of the outlet temperature of each assembly located at the center.

[0067] Since core assembly cooling is a combined effect of in-cell and inter-cell cooling, under stable natural circulation conditions, the in-cell and inter-cell states are relatively stable. The combined mass flow rate can be calculated using the relationship that heat output equals the product of specific heat, mass flow rate, and temperature rise. This conservative approach is used for higher power applications with similar conditions. Furthermore, regardless of whether the temperature measurement points cover the hottest assembly in the reactor core, the average flow rate G per assembly calculated using the above expression will be lower than the flow rate of the hottest assembly. Therefore, the average flow rate per assembly is conservative for assessing the cladding temperature of the hottest assembly. The calculated average flow rate per assembly is the aforementioned conservative core coolant flow rate.

[0068] See Figure 3 , Figure 3 These are the calculated curves showing the maximum cladding temperature versus time for a certain fast reactor under different pump coasting coolant flow rates and different test powers. Curves D1, D2, D3, D4, and D5 represent the maximum cladding temperature versus time under different coolant flow rates after pump coasting at a test power of 2.8%Pn, respectively. Curves E1, E2, E3, and E4 represent the maximum cladding temperature versus time under different coolant flow rates after pump coasting at a test power of 2.80%Pn, respectively. Curves F1, F2, F3, and F4 represent the cladding temperature versus time under different coolant flow rates after pump coasting at a test power of 0.8%Pn, respectively.

[0069] The coolant flow rates corresponding to curves D1, D2, D3, D4, and D5 are 0%Gn (Gn is the rated coolant flow rate of the reactor core), 1.4%Gn, 1.6%Gn, 1.8%Gn, and 2%Gn, respectively. The coolant flow rates corresponding to curves E1, E2, E3, and E4 are 0%Gn, 0.4%Gn, 0.42%Gn, and 0.5%Gn of the reactor core coolant flow rate, respectively. The coolant flow rates corresponding to curves F1, F2, F3, and F4 are 0%Gn, 0.25%Gn, 0.26%Gn, and 0.27%Gn of the reactor core coolant flow rate, respectively.

[0070] As mentioned earlier, the normal operating limit for the cladding is 700°C. When determining the conservative power limit, the maximum cladding temperature limit can be set at 650°C to allow for the temperature rise space at the core inlet and the conservative margin for the cladding temperature, ensuring the safety of the fast reactor during testing.

[0071] See Figure 3When the coolant flow rates are 0.26%Gn, 0.42%Gn, and 1.6%Gn, the cladding temperature of the fast reactor is near the limit of the maximum cladding temperature. The corresponding conservative power limits are 0.5%Pn, 0.8%Pn, and 2.8%Pn, respectively. That is, the conservative power limit of the fast reactor at a conservative coolant flow rate of 0.26%Gn is 0.5%Pn, the conservative power limit of the fast reactor at a conservative coolant flow rate of 0.42%Gn is 0.8%Pn, and the conservative power limit of the fast reactor at a conservative coolant flow rate of 1.6%Gn is 2.8%Pn.

[0072] Also see Figure 3 The ratio of the conservative coolant flow rate to the conservative power limit, which satisfies the limit that the cladding temperature does not exceed the maximum cladding temperature, gradually decreases as the conservative power limit increases. Therefore, when determining the conservative power limit, a critical ratio can also be selected. For example, when the conservative coolant flow rate is not higher than 1.6%Gn, the corresponding conservative power limit is taken as 1.750 times the flow rate (as a percentage).

[0073] In some embodiments, in step S3, after determining the next test power, the changes in the average core outlet temperature and the maximum cladding temperature over time of the fast reactor are determined according to the currently validated natural circulation design procedure; based on the determined changes in the average core outlet temperature and the maximum cladding temperature over time, the relationship between the average core outlet temperature and the maximum cladding temperature is determined; based on the relationship between the average core outlet temperature and the maximum cladding temperature, a safe limit for the core outlet temperature is determined; if the measured core outlet temperature is lower than the safe limit for the core outlet temperature, the pumps are kept off until a stable natural circulation is established, after which the reactor is shut down and the pumps are restarted. Since the cladding temperature within a fast reactor is difficult to obtain through measurement, in the embodiments of this application, after verifying the natural circulation design program, the changes in the average core outlet temperature over time and the changes in the maximum cladding temperature over time are obtained again. The relationship between the average core outlet temperature and the maximum cladding temperature is determined, so that the measured average core outlet temperature can more accurately reflect the maximum cladding temperature. Thus, based on the limit of the maximum cladding temperature and the relationship between the average core outlet temperature and the maximum cladding temperature, the safe limit of the core outlet temperature can be determined. At the same time, in the embodiments of this application, if the measured core outlet temperature is lower than the safe limit of the core outlet temperature, the pumps are kept off until a stable natural circulation is established, after which the reactor is shut down and the pumps are restarted to conduct a safe reactor test to obtain operational data.

[0074] In some embodiments, the establishment of a stable natural circulation can be determined based on the rate of change of the core outlet temperature measured per unit time. In such embodiments, since the relationship between the average core outlet temperature and the highest cladding temperature is determined, the rate of change of the core outlet temperature measured per unit time can reflect the rate of change of the highest cladding temperature per unit time, thereby reflecting the heat changes within the core and facilitating accurate determination of whether a stable natural circulation has been established.

[0075] In some embodiments, step S8, which uses the natural circulation design program verified in step S7 to evaluate the natural circulation capability of a fast reactor during shutdown and pump shutdown at rated power, includes: using the natural circulation design program verified in step S7 to determine whether the cladding temperature is within the cladding temperature safety limit before establishing natural circulation after shutdown and pump shutdown at rated power. If yes, the natural circulation capability of the fast reactor is reliable; if no, the natural circulation capability of the fast reactor is unreliable. A cladding temperature within the cladding temperature safety limit indicates that the time to establish natural circulation is less than the time for the cladding temperature to rise to the maximum cladding temperature, thus indicating that the natural circulation capability of the fast reactor is reliable; a cladding temperature exceeding the cladding temperature safety limit indicates that the time to establish natural circulation is greater than the time for the cladding temperature to rise to the maximum cladding temperature, thus indicating that the natural circulation capability of the fast reactor is unreliable. Therefore, the embodiments of this application, by determining whether the cladding temperature is within the cladding temperature safety limit before establishing natural circulation, facilitate an accurate and objective evaluation of the natural circulation capability of a fast reactor during shutdown and pump shutdown at rated power.

[0076] In some embodiments, when the fast reactor is shut down at rated power, the core power corresponding to the residual core heat is 3-4% of the fast reactor's rated power; the low-power test power is 15-25% of the fast reactor's rated power; and the high-power test power is more than 50% of the fast reactor's rated power. In such embodiments, setting the low-power test power to 15-25% of the fast reactor's rated power and the high-power test power to more than 50% of the fast reactor's rated power ensures that the required temperature difference can be formed between the cold pool and the hot pool of the fast reactor, while also ensuring that the fast reactor operates under safe conditions.

[0077] When a fast reactor is shut down at its rated power, the core power corresponding to the residual heat in the core is, for example, 3% of the rated power of the fast reactor.

[0078] In some embodiments, after determining the test power, the safety of shutting down the reactor and pumps or shutting down the reactor and pumps without stopping the reactor is verified using a natural circulation design procedure before each actual reactor test at that test power. This is to confirm the relationship between the highest cladding temperature after pump shutdown and the average core outlet temperature. Then, based on the average core outlet temperature after pump shutdown, it is determined that the highest cladding temperature after pump shutdown does not exceed the limit, ensuring the safety of the fast reactor after pump shutdown and ensuring that operational data can be obtained through safe actual reactor tests.

[0079] The low-power test power is, for example, 20% of the rated power of the fast reactor. At this test power, a low temperature difference can be formed between the cold pool and the hot pool of the fast reactor, and the three loops can be maintained by bypassing the system without involving the operation of the power grid and turbine. The highest cladding temperature at this test power, determined by the natural circulation design program, does not exceed 700°C, and the core outlet temperature measured in the actual reactor test does not exceed 590°C, thus meeting the safety requirements of the fast reactor.

[0080] In some embodiments, step S5 may include: stabilizing the fast reactor at 20% of its rated power, manually shutting down the reactor and pumps, simultaneously disconnecting the main heat transfer system and activating the emergency residual heat removal system, at which point the reactor core has residual heat; after the pumps stop coasting, acquiring the fast reactor's operating data until the core outlet temperature drops to a stable level, restarting the reactor and pumps, stopping the acquisition of the fast reactor's operating data, and ending the test.

[0081] For example, the high-power test power is 60% of the rated power of the fast reactor. At this test power, the highest temperature obtained by the natural circulation design program is lower than the highest cladding temperature of 650°C at the rated power of the fast reactor obtained by the natural circulation design program. Therefore, conducting actual reactor testing at this test power can create a high temperature difference of 170°C-180°C between the cold pool and the hot pool of the fast reactor, while ensuring the safety of the fast reactor.

[0082] In step S6, the high-power test power can be as high as possible. If the natural circulation design program calculates that the maximum cladding temperature is less than 700°C when the power is 100% Pn, then the high-power test power can be directly set to full power.

[0083] In this application, steps S1 and S2 aim to demonstrate the existence of natural circulation in the reactor under absolutely safe thermal conditions. They also aim to obtain, for the first time, the basic phenomena and data of natural circulation in a real reactor, without relying on measuring points or natural circulation design program calculations to ensure experimental safety, and to verify the natural circulation design program. Step S3 aims to obtain the fast reactor's steady-state natural circulation heat dissipation capacity and the drag characteristics at low flow rates during the critical stage of natural circulation after a full-power emergency shutdown. The real-reactor data obtained from steps S3 to S7 cover the key thermal-hydraulic phenomena and flow regime range of natural circulation after a full-power emergency shutdown, reliably verifying the relevant calculation models of the natural circulation design program, including the drag characteristics of the natural circulation flow regime and the ability to simulate the temperature field with large temperature differences in the hot and cold sodium pools. The verified natural circulation design program is used to calculate the natural circulation condition under a full-power emergency shutdown. By comparing the cladding temperature with the operating condition acceptance criteria, a conclusion is drawn regarding the adequacy of the fast reactor's natural circulation capacity.

[0084] Steps S3 through S7, combined with the validated natural circulation design program, calculate the natural circulation condition after a full-power emergency shutdown. This reliably demonstrates the reactor's natural circulation capability and ensures the effectiveness of the test implementation. These tests already encompass the key phenomena in the full-power emergency shutdown natural circulation condition, reliably validating the design program. Furthermore, the tests begin with an initial test under adiabatic conditions ensuring absolute thermal safety, subsequently progressively validating the program and advancing the tests based on actual reactor data and validated program calculations, thus ensuring the safety of the test implementation.

[0085] Regarding the embodiments of this application, it should also be noted that, without conflict, the embodiments of this application and the features in the embodiments can be combined with each other to obtain new embodiments.

[0086] The above description is merely a specific embodiment of this application, but the scope of protection of this application is not limited thereto. The scope of protection of this application shall be determined by the scope of the claims.

Claims

1. A method for conducting a real-world natural circulation test of a liquid metal-cooled fast reactor, characterized in that, include: S1. The initial test power is determined using a natural circulation design program so that the fast reactor can establish a stable natural circulation within a preset time after the pump is stopped while operating at the test power. The preset time is longer than the pump coasting time, and the initial test power is less than the core power corresponding to the residual core heat when the fast reactor is stopped at rated power. S2. Run the fast reactor at the test power, and run the fast reactor at the test power after the pump stops until a stable natural circulation is established. Obtain the operating data of the fast reactor after the pump stops coasting, and use the operating data to verify the natural circulation design program. S3. Based on the initial test power in step S1, gradually increase the test power of the fast reactor and repeat step S2 until the test power of the fast reactor is increased to the core power corresponding to the residual core heat when the fast reactor is shut down at rated power or until the test power can no longer be increased. S4. Using the natural circulation design procedure verified in step S3, determine the low-power test power that can create a temperature difference between the hot pool and the cold pool of the fast reactor. S5. Stop the fast reactor and pumps at the low power test power, and obtain the operating data of the fast reactor after the pump stops coasting, and use the operating data to verify the natural circulation design program. S6. Determine the high-power test power using the natural loop design program verified in step S5, wherein the high-power test power is higher than the low-power test power; S7. Stop the fast reactor and pumps at the high power test power, and obtain the operating data of the fast reactor after the pump stops coasting, and use the operating data to verify the natural circulation design program. S8. If the high-power test power is less than the rated power of the fast reactor, the natural circulation design program verified in step S7 is used to evaluate the natural circulation capability of the fast reactor when shutting down and stopping the pumps at the rated power.

2. The method according to claim 1, characterized in that, Step S1 includes: S11. For any given power value, the natural circulation design program is used to determine the change of the average core outlet temperature and the change of the highest cladding temperature of the fast reactor over time after the pump is stopped at the given power value. The given power value is less than the core power corresponding to the core residual heat when the fast reactor is stopped at the rated power. S12. Determine the change of the highest cladding temperature of the fast reactor under adiabatic conditions over time after the pump stops coasting at the given power value; S13. Based on the changes in the average core outlet temperature and the highest cladding temperature over time determined in step S11 and the changes in the highest cladding temperature over time determined in step S12, determine whether to use the given power value as the initial test power.

3. The method according to claim 2, characterized in that, Step S13 includes: S131. Based on the changes in the average core exit temperature and the maximum cladding temperature over time determined in step S11, determine the relationship between the average core exit temperature and the maximum cladding temperature under the given power value, and the first duration for the fast reactor to establish natural circulation under the given power value. S132. Determine the second duration corresponding to when the highest cladding temperature reaches the conservative limit of normal operation according to step S12. If the second duration is longer than the first duration, then use the given power value as the initial test power; otherwise, reselect the given power value.

4. The method according to claim 3, characterized in that, In step S132, the second duration is determined by the following method: the pump stop duration that can be maintained when the cladding temperature does not exceed the normal operating limit after the pump stops coasting, assuming that all boundaries of the active section of the hottest component in the reactor are in adiabatic conditions.

5. The method according to claim 1, characterized in that, Step S2 includes: S21. Run the fast reactor at the initial test power, and run the fast reactor at the initial test power for the preset duration after the pump is stopped, obtain the operating data of the fast reactor within the preset duration, use the operating data to determine whether the fast reactor has established a stable natural cycle, and use the operating data to verify the natural cycle design program. S22. If it is determined that the fast reactor has established a stable natural circulation, then stop the reactor and start the pump, and execute step S3.

6. The method according to claim 2, characterized in that, Step S2 also includes: S23. If it is determined that the fast reactor has not established a stable natural circulation, the reactor shall be stopped and the pump shall be started immediately after the preset time period of pump stoppage ends. The changes in the average core outlet temperature and the maximum cladding temperature of the fast reactor over time after the preset time period shall be determined according to the natural circulation design program verified in step S21. S24. Based on the changes in the average core exit temperature and the maximum cladding temperature over time determined in S23, determine the relationship between the average core exit temperature and the maximum cladding temperature. S25. Determine the safe limit for core exit temperature based on the relationship between the average core exit temperature and the highest cladding temperature. S26. Run the fast reactor at the initial test power, obtain the fast reactor's operating data after the preset time of pump shutdown, and keep the pump shut down if the measured core outlet temperature is lower than the core outlet temperature safety limit, until a stable natural circulation is established and then the reactor is shut down and the pump is restarted. S27. Verify the natural circulation design program using the operating data obtained in step S26 after the preset pump stop time.

7. The method according to claim 1, characterized in that, Step S3 includes: S31. Determine that the fast reactor can remove the heat from the reactor core by establishing a stable natural circulation at the initial test power, and determine the second test power based on this heat. S32. Run the fast reactor at the second test power, and run the fast reactor at the second test power after the pump stops until a stable natural circulation is established; obtain the operating data of the fast reactor after the pump stops coasting, and use the operating data to verify the natural circulation design program; S33. Determine that the natural circulation established by the fast reactor at the second test power can remove the heat from the reactor core, and determine the third test power based on the heat; until the operating power of the fast reactor is increased to the core power corresponding to the residual heat of the core when the fast reactor is shut down at rated power or until the test power can no longer be increased; use the operating data of the fast reactor after each pump stops coasting to verify the natural circulation design program.

8. The method according to claim 7, characterized in that, In steps S31 and S33 The core coolant flow rate is determined based on the operating power of the fast reactor, the core outlet temperature after establishing stable natural circulation, and the core inlet temperature. The amount of heat that the coolant can carry away is determined by the core coolant flow rate, and the corresponding core power is determined based on the core power corresponding to that amount of heat.

9. The method according to claim 7, characterized in that, In step S3, after determining the next test power, the changes in the core average exit temperature and the maximum cladding temperature of the fast reactor over time are determined according to the currently verified natural circulation design procedure. The relationship between the average core exit temperature and the maximum cladding temperature is determined based on the time-dependent changes in the average core exit temperature and the time-dependent changes in the maximum cladding temperature. The safe limit for core exit temperature is determined based on the relationship between the average core exit temperature and the highest cladding temperature. If the measured temperature at the core outlet is lower than the safe limit for the core outlet temperature, the pumps will remain shut down until a stable natural circulation is established, at which point the reactor will be shut down and the pumps will be restarted.

10. The method according to claim 1, characterized in that, Whether a stable natural circulation should be established is determined based on the rate of change of the core outlet temperature per unit time.

11. The method according to claim 1, characterized in that, In step S8, the steps for evaluating the natural circulation capability of the fast reactor under rated power shutdown and pump shutdown using the natural circulation design procedure verified in step S7 include: Using the natural circulation design procedure verified in step S7, determine whether the cladding temperature is within the safe limits for cladding temperature after the fast reactor is shut down and pumps are stopped at rated power and before natural circulation is established. If yes, then the natural cycling capability of the fast heap is reliable; if no, then the natural cycling capability of the fast heap is unreliable.

12. The method according to claim 1, characterized in that, The initial test power is less than or equal to 1% of the fast reactor's rated power; When a fast reactor is shut down at its rated power, the core power corresponding to the residual heat in the core is 3-4% of the rated power of the fast reactor. The low-power test power is 15-25% of the rated power of the fast reactor; The high-power test power is more than 50% of the rated power of the fast reactor.

Images