A method and apparatus for simulating a loss of heat sink accident for a heat pipe reactor

Abstract

This invention discloses a method and apparatus for simulating heat pipe reactor loss-of-heat-sink accidents, including a heat pipe reactor simulation device, a DC power supply, and a secondary loop auxiliary system. The DC power supply is independently located outside the heat pipe reactor simulation device and electrically connected to it. The secondary loop auxiliary system is connected to the piping of the heat pipe reactor simulation device. This solution, based on "scaled modeling + precise modeling + safe substitution," constructs a heat pipe reactor accident simulation system that combines realism, safety, and economy. It avoids nuclear risks while accurately reproducing key scenarios of coolant loss accidents, solving the pain points of real reactor testing. It provides an efficient technical means for the safe design, optimization, and accident response of heat pipe reactors, and is of great significance for improving the safe operation level of nuclear reactors.

Description

Technical Field

[0001] This invention relates to the field of heat pipe reactor technology, specifically to a method and apparatus for simulating heat pipe reactor loss of heat sink accidents. Background Technology

[0002] Loss of coolant is a typical accident condition in heat pipe reactors. Failure of air supply system piping valves, unexpected shutdown of power machinery, etc., can all lead to loss of coolant in heat pipe reactors.

[0003] In existing technologies, the reactor core's safety system would cut off the nuclear fuel chain reaction in the event of such an accident. However, the heat generated by the decay of nuclear fuel could not be dissipated, and the core temperature continued to rise. If the core temperature continued to rise, it would cause the core and protective facilities to melt, leading to a serious nuclear accident.

[0004] Therefore, a solution is needed. Summary of the Invention

[0005] (a) Technical problems to be solved To address the shortcomings of existing technologies, this invention provides a method and apparatus for simulating heat sink loss accidents in heat pipe reactors, thereby resolving the problems mentioned in the background section.

[0006] (II) Technical Solution To achieve the above objectives, the present invention provides the following technical solution: A method and apparatus for simulating heat pipe reactor loss of heat sink accident includes a heat pipe reactor simulation device, a DC power supply and a secondary loop auxiliary system. The DC power supply is independently installed outside the heat pipe reactor simulation device and electrically connected to the heat pipe reactor simulation device. The secondary loop auxiliary system is connected to the pipeline of the heat pipe reactor simulation device. The heat pipe reactor simulation device includes a heat exchanger, an electric heating tube, a waste heat removal device, insulation cotton, and temperature measuring points. The electric heating tube is located inside the heat exchanger, the waste heat removal device is located at the end of the heat exchanger, the insulation cotton is located on the outside of the heat exchanger and at the connection between the heat exchanger and the waste heat removal device, and the temperature measuring points are located on the inner wall of the heat exchanger, the inner wall of the waste heat removal device, and the surface of the electric heating tube. The DC power supply is electrically connected to the electric heating tube and is used to supply power. The waste heat removal device is a metal cylindrical structure and is wrapped with a removable insulation layer.

[0007] Preferably, the secondary loop auxiliary system includes a compressor, an electric heater, a valve group, and measuring instruments. The compressor, electric heater, and valve group are connected in series via pipelines to form the core fluid channel of the secondary loop. The outlet of the compressor is connected to the inlet of the electric heater via a pipeline, the outlet of the electric heater is connected to the inlet of the valve group via a pipeline, and the outlet of the valve group is connected to the heat pipe reactor simulation device via a pipeline, ultimately forming a closed loop. The compressor is used for pressurizing the air in the secondary loop, the electric heater is used for regulating the air temperature in the secondary loop, the valve group is used for regulating the air pressure and flow rate in the secondary loop, and the measuring instruments are used for monitoring the operating parameters of the secondary loop.

[0008] Preferably, the heat pipe reactor simulation device is designed in scale according to the original reactor heat exchanger structure. The metal materials used in the heat exchanger, electric heating tube, and waste heat removal device, as well as the insulation materials used in the insulation cotton, are all consistent with the materials of the components corresponding to the real heat pipe reactor.

[0009] Preferably, the method for simulating a heat pipe reactor loss of heat sink accident includes the following steps: a. Determine the event-time series of a heat pipe reactor coolant loss accident, wherein the event-time series includes T0 (steady state before the accident), T1 (coolant flow rate begins to decrease), T2 (coolant flow rate drops to zero), T3 (core high temperature signal trigger), T4 (start of core power change according to the decay heat curve), T5 (time when the residual heat removal system is put into operation), T6 (start of the stage where the heat-carrying capacity of the residual heat removal system is continuously higher than the core decay heat power), or T7 (start of the stage where the heat-carrying capacity of the residual heat removal system is continuously lower than the core decay heat power). b. Modeling event-time test parameters for coolant loss accidents in heat pipe reactors; c. Perform test parameter loading and result verification to determine the safety of the heat pipe reactor under accident conditions.

[0010] Preferably, the method for determining each time point and corresponding stage in step a is as follows: T0: Steady state before the accident The average wall temperature and average surface temperature of the heat pipes before an accident are obtained by using actual reactor operating parameters or design parameters. T1: Coolant flow rate begins to decrease Let the assumed start time of the accident be T1, and the time length between T1 and T0 is unlimited; T2: Coolant flow rate drops to zero During the T1-T2 time period, the coolant flow rate gradually decreases to zero according to the flow rate-time curve. The flow rate-time curve is obtained by recording actual reactor operating parameters or by calculating pipeline valve failure accidents or measuring unexpected shutdown accidents of power machinery. T3: Core high temperature signal triggered During the T2-T3 time period, the reactor core continues to operate at the power at time T0. The duration of this time period is obtained from the actual reactor operating parameters or by solving the heat conduction formula using the finite element method, until the highest temperature of the calculation unit reaches the core high temperature signal trigger temperature. T4: Core power based on decay heat curve During the T3-T4 time period, the reactor core continues to operate at the power at time T0. The duration of this time period is the delay time from when the high temperature signal of the reactor core is triggered to when the control system receives the shutdown signal. Starting from time T4, the reactor core power changes according to the decay heat curve. T5: Waste heat removal system put into operation During the T4-T5 time period, the core power changes according to the decay heat curve. The duration of this time period is the delay between the control system receiving the signal to activate the residual heat removal system and the activation of the residual heat removal system. The residual heat removal system is put into operation starting from time T5. T6: The residual heat removal system's heat-carrying capacity is consistently higher than the core decay heat power. During the T4-T6 time period, the core power varied according to the decay heat curve. The residual heat removal system remained operational. The heat-carrying capacity of the residual heat removal system consistently exceeded the core decay heat power, indicating that the residual heat removal device could promptly remove the decay heat released by the core nuclear fuel after a loss-of-coolant accident in the heat pipe reactor, and that the heat pipe reactor was safe under this accident. The parameters of the heat pipe reactor during this time period were verified by experiments. T7: The residual heat removal system's heat-carrying capacity is consistently higher than the core decay heat power. During the T4-T7 time period, the core power varied according to the decay heat curve. The residual heat removal system remained operational. The residual heat removal system's heat-carrying capacity consistently fell below the core decay heat power, indicating that after a coolant loss accident in a heat pipe reactor, the residual heat removal device was unable to promptly remove the decay heat released from the core nuclear fuel, rendering the heat pipe reactor unsafe under this accident condition. The parameters of the heat pipe reactor during this time period were verified experimentally.

[0011] Preferably, the method for calculating the flow-time curve of the pipeline valve failure in step T2 of step a is as follows: measure the valve opening-time curve under failure conditions; measure the relationship between valve opening and Cv value; convert the opening-time curve into a Cv value-time curve; and use the formula... (Where —— pressure difference across the valve, —— working fluid density) Calculate the relationship curve between pipeline flow rate and Cv value; convert the Cv value-time curve into a flow rate-time curve; unexpected shutdown of power machinery: measure the flow rate-time curve during the coasting process of the unexpected shutdown of power machinery.

[0012] Preferably, the heat conduction formula in step a, T3, is: In the formula, — calculate the unit density; Cp — Specific heat capacity of the calculation unit; k — thermal conductivity of the calculation unit; T — Temperature of the calculation unit; Q – Heat generated within the calculation unit.

[0013] Preferably, the modeling rule in step b is: To reduce testing costs, the simulation of coolant loss accidents in heat pipe reactors requires a scaled-down design of the original reactor heat exchanger, and the test parameters also need to be modeled. The test parameters are modeled using the following method.

[0014] f. Determine the scale n of the heat pipe reactor simulation device, i.e. ; i. The modulus ratio of decay heat power is n, i.e. ; j. The time scaling ratio is 1 / n, that is... ; k. The temperature scaling ratio is 1, that is... ; l. The metal materials and insulation materials are consistent with the actual pile.

[0015] Preferably, the steps of loading experimental parameters and verifying results in step c are as follows: o. Initial state: Start the compressor, adjust the power of the electric heater and the valve group opening ratio of the secondary loop auxiliary system to make the temperature measuring points of the heat pipe reactor simulator reach the average temperature of the reactor heat pipe heat exchanger wall and the average temperature of the heat pipe surface at time T3, and then quickly cut off the hot air introduced. p. Phase 1: The time period is (T3—T4) / n, and the DC power supply is adjusted to the actual reactor power at time T3—T4 / n; q. Stage 2: Adjust the DC power supply according to... The change allows for the rapid removal of the removable insulation from the waste heat exhaust device; r. Phase 3: Monitor temperature changes through temperature measuring points. If the temperature measurement value continues to decrease, it indicates that the heat pipe reactor is safe under a coolant loss accident; if the temperature measurement value continues to rise, it indicates that the heat pipe reactor is unsafe under the accident.

[0016] Preferably, in step a, T6, the parameters are verified by experiment to confirm that the residual heat removal device can timely remove the decay heat released by the nuclear fuel in the reactor core; in step a, T7, the parameters are verified by experiment to confirm that the residual heat removal device cannot timely remove the decay heat released by the nuclear fuel in the reactor core.

[0017] (III) Beneficial Effects This invention provides a method and apparatus for simulating heat sink loss accidents in heat pipe reactors. It has the following beneficial effects: 1. This scheme accurately reproduces the accident process by matching materials, scaling down to the correct scale, and defining modeling rules, combined with temperature measurement points at the core location. The simulated data closely matches the actual operating conditions, providing a reliable basis for reactor safety assessment.

[0018] 2. The device uses electric heating tubes to simulate the core decay heat, without involving nuclear fuel throughout the process, achieving a ground simulation test with zero nuclear risk and avoiding the safety hazards of real reactor tests.

[0019] 3. This solution covers the core causes of coolant loss accidents, supports multi-scenario simulation, takes into account two key operating conditions of waste heat removal systems, and meets the needs of accident mechanism research and safety verification.

[0020] 4. The proportionally scaled design reduces the size of the device and the consumption of materials, eliminating the need for the complex system and high cost of a real reactor, significantly reducing the overall testing cost and making it suitable for large-scale iterative testing.

[0021] 5. The waste heat removal device supports the replacement of heat dissipation methods, and the secondary loop is flexibly adjustable to adapt to the diverse test requirements of different reactors.

[0022] 6. By monitoring temperature measurement points and establishing safety judgment standards, the effectiveness of the residual heat removal system can be directly verified, the safety boundaries of accidents can be clearly defined, and data support can be provided for reactor safety design optimization. Attached Figure Description

[0023] Figure 1 This is a schematic diagram of the heat pipe reactor simulation device of the present invention; Figure 2 This is an event-time series diagram of a coolant loss accident in a heat pipe reactor according to the present invention; Figure 3 This is a diagram showing the loading of test parameters for this invention.

[0024] In the figure, 1-heat pipe reactor simulation device; 11-heat exchanger; 12-electric heating tube; 13-waste heat removal device; 131-removable insulation layer; 14-insulation cotton; 15-temperature measuring point; 2-DC power supply; 3-secondary circuit auxiliary system; 31-compressor; 32-electric heater; 33-valve group; 34-measuring instrument. Detailed Implementation

[0025] The technical solutions of the embodiments of the present invention will be clearly and completely described below with reference to the accompanying drawings. Obviously, the described embodiments are only some embodiments of the present invention, and not all embodiments. Based on the embodiments of the present invention, all other embodiments obtained by those skilled in the art without creative effort are within the scope of protection of the present invention.

[0026] Please see Figure 1-3 The present invention provides a technical solution to achieve this: including a heat pipe reactor simulation device 1, a DC power supply 2, and a secondary loop auxiliary system 3. The DC power supply 2 is independently set outside the heat pipe reactor simulation device 1 and electrically connected to the heat pipe reactor simulation device 1. The secondary loop auxiliary system 3 is connected to the heat pipe reactor simulation device 1 through a pipeline.

[0027] The heat pipe reactor simulation device 1 includes a heat exchanger 11, an electric heating tube 12, a waste heat removal device 13, insulation cotton 14, and temperature measuring points 15. The electric heating tube 12 is located inside the heat exchanger 11, the waste heat removal device 13 is located at the end of the heat exchanger 11, the insulation cotton 14 is located on the outside of the heat exchanger 11 and at the connection between the heat exchanger 11 and the waste heat removal device 13, and the temperature measuring points 15 are located on the inner wall of the heat exchanger 11, the inner wall of the waste heat removal device 13, and the surface of the electric heating tube 12. A DC power supply 2 is electrically connected to the electric heating tube 12 and is used to supply power. The waste heat removal device 13 is a metal cylindrical structure and is wrapped with a removable insulation layer 131.

[0028] In detail, the secondary loop auxiliary system 3 includes a compressor 31, an electric heater 32, a valve group 33, and a measuring instrument 34. The compressor 31, electric heater 32, and valve group 33 are connected in series through pipelines to form the core fluid channel of the secondary loop. The outlet of the compressor 31 is connected to the inlet of the electric heater 32 through a pipeline. The outlet of the electric heater 32 is connected to the inlet of the valve group 33 through a pipeline. The outlet of the valve group 33 is connected to the heat pipe reactor simulation device 1 through a pipeline, ultimately forming a closed loop. The compressor 31 is used for pressurizing the air in the secondary loop, the electric heater 32 is used for adjusting the temperature of the air in the secondary loop, the valve group 33 is used for regulating the air pressure and flow rate in the secondary loop, and the measuring instrument 34 is used to monitor the operating parameters of the secondary loop.

[0029] The heat pipe reactor simulation device 1 is designed in scale according to the original reactor heat exchanger structure. The metal materials used in the heat exchanger 11, electric heating tube 12, waste heat removal device 13 and the insulation materials used in the insulation cotton 14 are all the same as the corresponding components of the real heat pipe reactor.

[0030] The simulation method for heat pipe reactor loss of heat sink accident includes the following steps: a. Determine the event-time series of a coolant loss accident in a heat pipe reactor. The event-time series includes T0 (steady state before the accident), T1 (coolant flow rate begins to decrease), T2 (coolant flow rate drops to zero), T3 (core high temperature signal is triggered), T4 (start of core power change according to the decay heat curve), T5 (time when the residual heat removal system is put into operation), T6 (start of the stage where the heat-carrying capacity of the residual heat removal system is continuously higher than the core decay heat power) or T7 (start of the stage where the heat-carrying capacity of the residual heat removal system is continuously lower than the core decay heat power). b. Modeling event-time test parameters for coolant loss accidents in heat pipe reactors; c. Perform test parameter loading and result verification to determine the safety of the heat pipe reactor under accident conditions. The methods for determining each moment and corresponding stage in step a are as follows: T0: Steady state before the accident The average wall temperature and average surface temperature of the heat pipes before an accident are obtained by using actual reactor operating parameters or design parameters. T1: Coolant flow rate begins to decrease Let the assumed start time of the accident be T1, and the time length between T1 and T0 is unlimited; T2: Coolant flow rate drops to zero During the T1-T2 time period, the coolant flow rate gradually decreases to zero according to the flow rate-time curve. The flow rate-time curve is obtained by recording actual reactor operating parameters or by calculating pipeline valve failure accidents or measuring unexpected shutdown accidents of power machinery. T3: Core high temperature signal triggered During the T2-T3 time period, the reactor core continues to operate at the power at time T0. The duration of this time period is obtained from the actual reactor operating parameters or by solving the heat conduction formula using the finite element method, until the highest temperature of the calculation unit reaches the core high temperature signal trigger temperature. T4: Core power based on decay heat curve During the T3-T4 time period, the reactor core continues to operate at the power at time T0. The duration of this time period is the delay time from when the high temperature signal of the reactor core is triggered to when the control system receives the shutdown signal. Starting from time T4, the reactor core power changes according to the decay heat curve. T5: Waste heat removal system put into operation During the T4-T5 time period, the core power changes according to the decay heat curve. The duration of this time period is the delay between the control system receiving the signal to activate the residual heat removal system and the activation of the residual heat removal system. The residual heat removal system is put into operation starting from time T5. T6: The residual heat removal system's heat-carrying capacity is consistently higher than the core decay heat power. During the T4-T6 time period, the core power varied according to the decay heat curve. The residual heat removal system remained operational. The heat-carrying capacity of the residual heat removal system consistently exceeded the core decay heat power, indicating that the residual heat removal device could promptly remove the decay heat released by the core nuclear fuel after a loss-of-coolant accident in the heat pipe reactor, and that the heat pipe reactor was safe under this accident. The parameters of the heat pipe reactor during this time period were verified by experiments. T7: The residual heat removal system's heat-carrying capacity is consistently higher than the core decay heat power. During the T4-T7 time period, the core power varied according to the decay heat curve. The residual heat removal system remained operational. The residual heat removal system's heat-carrying capacity consistently fell below the core decay heat power, indicating that after a coolant loss accident in a heat pipe reactor, the residual heat removal device was unable to promptly remove the decay heat released from the core nuclear fuel, rendering the heat pipe reactor unsafe under this accident condition. The parameters of the heat pipe reactor during this time period were verified experimentally.

[0031] The method for calculating the flow-time curve of the pipeline valve failure in step T2 of step a is as follows: Measure the valve opening-time curve under failure conditions; measure the relationship between valve opening and Cv value; convert the opening-time curve into a Cv value-time curve; and then use the formula... (Where —— pressure difference across the valve, —— working fluid density) Calculate the relationship curve between pipeline flow rate and Cv value; convert the Cv value-time curve into a flow rate-time curve; unexpected shutdown of power machinery: measure the flow rate-time curve during the coasting process of the unexpected shutdown of power machinery.

[0032] The heat conduction formula in step a, T3 is: In the formula, — calculates the unit density; Cp — Specific heat capacity of the calculation unit; k — thermal conductivity of the calculation unit; T — Temperature of the calculation unit; Q – Heat generated within the calculation unit.

[0033] The modeling rules in step b are as follows: To reduce testing costs, the simulation of heat pipe reactor loss-of-heat-sink accidents requires a scaled-down design of the original reactor heat exchanger, and the test parameters also need to be modeled. The test parameters are modeled using the following method.

[0034] f. Determine the scale n of the heat pipe reactor simulation device, i.e. ; i. The modulus ratio of decay heat power is n, i.e. ; j. The time scaling ratio is 1 / n, that is... ; k. The temperature scaling ratio is 1, that is... ; l. The metal materials and insulation materials are consistent with the actual pile.

[0035] The steps for loading experimental parameters and verifying results in step c are as follows: o. Initial state: Start the compressor, adjust the power of the electric heater and the valve group opening ratio of the secondary loop auxiliary system to make the temperature measuring points of the heat pipe reactor simulator reach the average temperature of the reactor heat pipe heat exchanger wall and the average temperature of the heat pipe surface at time T3, and then quickly cut off the hot air introduced. p. Phase 1: The time period is (T3—T4) / n, and the DC power supply is adjusted to the actual reactor power at time T3—T4 / n; q. Stage 2: Adjust the DC power supply according to... The change allows for the rapid removal of the removable insulation from the waste heat exhaust device; r. Phase 3: Monitor temperature changes through temperature measuring points. If the temperature measurement value continues to decrease, it indicates that the heat pipe reactor is safe under a coolant loss accident; if the temperature measurement value continues to rise, it indicates that the heat pipe reactor is unsafe under the accident.

[0036] In step a, T6, the parameters are verified through experiments to confirm that the residual heat removal device can promptly remove the decay heat released by the nuclear fuel in the reactor core; in step a, T7, the parameters are verified through experiments to confirm that the residual heat removal device cannot promptly remove the decay heat released by the nuclear fuel in the reactor core.

[0037] Solution Analysis: 1. High simulation realism and accuracy, with strong data reference value: This scheme adopts a design with materials consistent with the corresponding components of a real heat pipe reactor, and performs scaled-down modeling according to the original reactor heat exchanger structure. Combined with clear modeling rules (decay heat power modeling ratio n, time modeling ratio 1 / n, temperature modeling ratio 1), it accurately reproduces the key physical processes of accident evolution. Simultaneously, by deploying temperature measuring points on the heat exchanger wall and heat pipe surface, core temperature parameters are captured in real time, ensuring that the simulation data closely matches real accident conditions, providing a reliable basis for reactor safety assessment.

[0038] 2. The test is safe and controllable, with no actual nuclear risk: The device uses electric heating tube 12 to simulate the core decay heat, without involving real nuclear fuel throughout the process, thus fundamentally avoiding serious safety hazards such as nuclear leakage and core meltdown during the test. Compared to the high risk of real reactor accident tests, this scheme achieves a "zero nuclear risk" test through ground simulation, completely avoiding casualties and environmental damage caused by the expansion of an accident.

[0039] 3. Comprehensive applicability, covering typical accident types: The solution specifically covers the core causes of coolant loss accidents in heat pipe reactors. It supports obtaining flow-time curves through real operating parameters and can also simulate typical scenarios such as pipeline valve failures and unexpected shutdowns of power machinery through calculation methods. At the same time, it takes into account two key operating conditions: the heat-carrying capacity of the residual heat removal system is "higher" and "lower" than the core decay heat power, thus meeting multiple needs such as accident mechanism research and safety boundary verification.

[0040] 4. Excellent economic efficiency, reducing R&D and testing costs: The scaled-down design significantly reduces the size of the device and material consumption, eliminating the need to build the complex system of a real reactor; the testing process can be repeated without incurring high costs related to nuclear fuel use and nuclear safety protection. Compared to full-scale reactor testing, this approach significantly reduces the overall cost of equipment development, operating condition simulation, and data acquisition, making it suitable for large-scale iterative testing.

[0041] 5. Flexible and convenient operation, and strong test scalability: The waste heat removal device supports the replacement of natural convection and forced convection heat dissipation methods. The secondary loop can flexibly adjust the initial test conditions through the coordinated control of the compressor 31, electric heater 32, and valve group 33. The test steps are carried out in an orderly manner according to the event-time sequence, and the parameter loading and result verification logic is clear. It is not only easy for operators to get started quickly, but also allows the scale ratio and modeling parameters to be adjusted according to different reactor models to adapt to diverse test needs.

[0042] 6. Direct Data Support for Safety Design Optimization: Continuous monitoring at 15 temperature measurement points and the application of safety assessment criteria (continuous temperature decrease indicates safety, continuous temperature increase indicates unsafety) directly verify the heat dissipation effectiveness of the residual heat removal system and clarify the reactor's safety boundaries under coolant loss accidents. The test results can provide targeted feedback on design flaws, offering direct data support for core structure optimization, residual heat removal system improvement, and the refinement of safety control strategies.

[0043] Working principle: Using a scaled-down heat pipe reactor simulation device with materials identical to those of a real heat pipe reactor as its core, the device is powered by a DC power supply 2 to supply electric heating tubes 12 to simulate core decay heat. First, the event-time sequence of a coolant loss accident is determined according to real reactor parameters or calculation methods. Then, the test parameters are modeled according to the scale n and the rules of "decay heat power modeling ratio n, time modeling ratio 1 / n, and temperature modeling ratio 1". The initial test environment is coordinated and controlled by the compressor 31, electric heater 32, and valve group 33 of the secondary loop auxiliary system 3. The device is operated in stages according to the parameter loading process after modeling. Temperature changes are monitored in real time using temperature measuring points 15 located at key positions. Finally, the device verifies whether the residual heat removal device can timely remove decay heat based on the response results of continuous temperature decrease or increase, thereby judging the safety of the heat pipe reactor under a coolant loss accident.

[0044] Technical effects of implementing this solution: This scheme, through its core design of "scaled modeling + precise modeling + safe substitution," constructs a heat pipe reactor accident simulation system that combines realism, safety, and economy. Its core value lies in achieving high-precision reproduction of key coolant loss-of-coolant accident scenarios while completely avoiding nuclear risks. This addresses the pain points of "high risk, high cost, and difficulty in implementation" in real reactor accident testing. Furthermore, through comprehensive scenario coverage and direct result verification, it provides efficient and feasible technical means for the safety design, performance optimization, and accident response strategy formulation of heat pipe reactors. It balances experimental accuracy, operational flexibility, and engineering practicality, and has significant practical implications for improving the safe operation level of nuclear reactors.

[0045] The present invention comprises: 1-heat pipe reactor simulation device; 11-heat exchanger; 12-electric heating tube; 13-residual heat removal device; 131-removable insulation layer; 14-insulation cotton; 15-temperature measuring point; 2-DC power supply; 3-secondary loop auxiliary system; 31-compressor; 32-electric heater; 33-valve group; 34-measuring instrument. These components are all general standard parts or components known to those skilled in the art. Their structure and principle can be learned by those skilled in the art through technical manuals or conventional experimental methods. The problem solved by the present invention is that failure of the air supply system pipeline valves, unexpected shutdown of power machinery, etc., may cause heat pipe reactor coolant loss accidents. At this time, although the safety system cuts off the nuclear fuel chain reaction, the nuclear fuel decay heat cannot be discharged, which will lead to the melting of the reactor core and protective facilities, and thus cause a serious nuclear accident. This invention, with "scaled modeling + precise modeling + safe substitution" as its core, constructs a heat pipe reactor accident simulation system that combines realism, safety and economy. It avoids nuclear risks while accurately reproducing key scenarios of coolant loss accidents, solving the pain points of real reactor testing. It provides efficient technical means for the safe design, optimization and accident response of heat pipe reactors, and is of great significance to improving the level of safe operation of nuclear reactors.

[0046] The foregoing has shown and described the basic principles, main features, and advantages of the present invention. It will be apparent to those skilled in the art that the present invention is not limited to the details of the exemplary embodiments described above, and that the invention can be implemented in other specific forms without departing from its spirit or essential characteristics. Therefore, the embodiments should be considered illustrative and non-limiting in all respects, and the scope of the invention is defined by the appended claims rather than the foregoing description. Thus, all variations falling within the meaning and scope of equivalents of the claims are intended to be included within the present invention. No reference numerals in the claims should be construed as limiting the scope of the claims.

[0047] Furthermore, it should be understood that although this specification describes embodiments, not every embodiment contains only one independent technical solution. This narrative style is merely for clarity. Those skilled in the art should consider the specification as a whole, and the technical solutions in each embodiment can also be appropriately combined to form other embodiments that can be understood by those skilled in the art.

Claims

1. A heat pipe reactor loss of heat sink accident simulation apparatus, characterized by: It includes a heat pipe reactor simulation device (1), a DC power supply (2) and a secondary loop auxiliary system (3). The DC power supply (2) is independently located outside the heat pipe reactor simulation device (1) and electrically connected to the heat pipe reactor simulation device (1). The secondary loop auxiliary system (3) is connected to the heat pipe reactor simulation device (1) via pipeline. The heat pipe reactor simulation device (1) includes a heat exchanger (11), an electric heating tube (12), a waste heat removal device (13), insulation cotton (14), and temperature measuring points (15). The electric heating tube (12) is located inside the heat exchanger (11), the waste heat removal device (13) is located at the end of the heat exchanger (11), the insulation cotton (14) is located on the outside of the heat exchanger (11) and at the connection between the heat exchanger (11) and the waste heat removal device (13), and the temperature measuring points (15) are located on the inner wall of the heat exchanger (11), the inner wall of the waste heat removal device (13), and the surface of the electric heating tube (12). The DC power supply (2) is electrically connected to the electric heating tube (12) and is used to supply power. The waste heat removal device (13) is a metal cylindrical structure and is wrapped with a removable insulation layer (131).

2. The thermal pipe reactor loss of heat sink accident simulation apparatus according to claim 1, characterized by: The secondary loop auxiliary system (3) includes a compressor (31), an electric heater (32), a valve group (33), and a measuring instrument (34). The compressor (31), the electric heater (32), and the valve group (33) are connected in series through pipelines to form the core fluid channel of the secondary loop. The outlet of the compressor (31) is connected to the inlet of the electric heater (32) through a pipeline. The outlet of the electric heater (32) is connected to the inlet of the valve group (33) through a pipeline. The outlet of the valve group (33) is connected to the heat pipe reactor simulation device (1) through a pipeline, thus forming a closed loop. The compressor (31) is used for pressurizing the air in the secondary loop, the electric heater (32) is used for adjusting the air temperature in the secondary loop, the valve group (33) is used for regulating the air pressure and flow rate in the secondary loop, and the measuring instrument (34) is used to monitor the operating parameters of the secondary loop.

3. The thermal pipe reactor loss of heat sink accident simulation apparatus of claim 1, wherein: The heat pipe reactor simulation device (1) is designed in scale according to the original reactor heat exchanger structure. The metal materials used in the heat exchanger (11), electric heating tube (12), waste heat discharge device (13) and the insulation material used in the insulation cotton (14) are consistent with the materials of the components corresponding to the real heat pipe reactor.

4. A method for simulating loss of heat sink accident of a heat pipe reactor based on the device according to claims 1 to 3, characterized in that: Includes the following steps: a. Determine the event-time series of a heat pipe reactor coolant loss accident, wherein the event-time series includes T0 (steady state before the accident), T1 (coolant flow rate begins to decrease), T2 (coolant flow rate drops to zero), T3 (core high temperature signal trigger), T4 (start of core power change according to the decay heat curve), T5 (time when the residual heat removal system is put into operation), T6 (start of the stage where the heat-carrying capacity of the residual heat removal system is continuously higher than the core decay heat power), or T7 (start of the stage where the heat-carrying capacity of the residual heat removal system is continuously lower than the core decay heat power). b. Modeling event-time test parameters for coolant loss accidents in heat pipe reactors; c. Perform test parameter loading and result verification to determine the safety of the heat pipe reactor under accident conditions.

5. The method of claim 4, wherein: The methods for determining each time point and corresponding stage in step a are as follows: T0: Steady state before the accident The average wall temperature and average surface temperature of the heat pipes before an accident are obtained by using actual reactor operating parameters or design parameters. T1: Coolant flow rate begins to decrease Let the assumed start time of the accident be T1, and the time length between T1 and T0 is unlimited; T2: Coolant flow rate drops to zero During the T1-T2 time period, the coolant flow rate gradually decreases to zero according to the flow rate-time curve. The flow rate-time curve is obtained by recording actual reactor operating parameters or by calculating pipeline valve failure accidents or measuring unexpected shutdown accidents of power machinery. T3: Core high temperature signal triggered During the T2-T3 time period, the reactor core continues to operate at the power at time T0. The duration of this time period is obtained from the actual reactor operating parameters or by solving the heat conduction formula using the finite element method, until the highest temperature of the calculation unit reaches the core high temperature signal trigger temperature. T4: Core power based on decay heat curve During the T3-T4 time period, the reactor core continues to operate at the power at time T0. The duration of this time period is the delay time from when the high temperature signal of the reactor core is triggered to when the control system receives the shutdown signal. Starting from time T4, the reactor core power changes according to the decay heat curve. T5: Waste heat removal system put into operation During the T4-T5 time period, the core power changes according to the decay heat curve. The duration of this time period is the delay between the control system receiving the signal to activate the residual heat removal system and the activation of the residual heat removal system. The residual heat removal system is put into operation starting from time T5. T6: The residual heat removal system's heat-carrying capacity is consistently higher than the core decay heat power. During the T4-T6 time period, the core power varied according to the decay heat curve. The residual heat removal system remained operational. The heat-carrying capacity of the residual heat removal system consistently exceeded the core decay heat power, indicating that the residual heat removal device could promptly remove the decay heat released by the core nuclear fuel after a loss-of-coolant accident in the heat pipe reactor, and that the heat pipe reactor was safe under this accident. The parameters of the heat pipe reactor during this time period were verified by experiments. T7: The residual heat removal system's heat-carrying capacity is consistently higher than the core decay heat power. During the T4-T7 time period, the core power varied according to the decay heat curve. The residual heat removal system remained operational. The residual heat removal system's heat-carrying capacity consistently fell below the core decay heat power, indicating that after a coolant loss accident in a heat pipe reactor, the residual heat removal device was unable to promptly remove the decay heat released from the core nuclear fuel, rendering the heat pipe reactor unsafe under this accident condition. The parameters of the heat pipe reactor during this time period were verified experimentally.

6. The method for simulating a heat pipe reactor heat sink loss accident according to claim 5, characterized in that: The flow-time curve calculation method of the pipeline valve failure accident in T2 of step a is: measuring the opening-time curve of the valve under the failure condition; measuring the relationship between the valve opening and Cv value; converting the opening-time curve into Cv value-time value curve; calculating the pipeline flow-Cv value relationship curve through formula (wherein ——pressure difference before and after the valve, ——working medium density) converting the Cv value-time curve into flow-time curve; power mechanical unexpected shutdown accident: measuring the flow-time curve during the mechanical inertia rotation process of the power mechanical unexpected shutdown.

7. The method for simulating a heat pipe reactor heat sink loss accident according to claim 5, characterized in that: The heat conduction formula in step a, T3, is: In the formula, — calculates the unit density; Cp — Specific heat capacity of the calculation unit; k — thermal conductivity of the calculation unit; T — Temperature of the calculation unit; Q – Heat generated within the calculation unit.

8. The method for simulating a heat pipe reactor heat sink loss accident according to claim 4, characterized in that: The modularization rule in step b is: To reduce testing costs, the simulation of coolant loss accidents in heat pipe reactors requires a scaled-down design of the original reactor heat exchanger, and the test parameters also need to be modeled. The test parameters are modeled using the following method. f. Determine the scale n of the heat pipe reactor simulation device, i.e. ; i. The modulus ratio of decay heat power is n, i.e. ; j. The time scaling ratio is 1 / n, that is... ; k. The temperature scaling ratio is 1, that is... ; l. The metal materials and insulation materials are consistent with the actual pile.

9. The method for simulating a heat pipe reactor heat sink loss accident according to claim 4, characterized in that: The steps for loading experimental parameters and verifying results in step c are as follows: o. Initial state: Start the compressor, adjust the power of the electric heater and the valve group opening ratio of the secondary loop auxiliary system to make the temperature measuring points of the heat pipe reactor simulator reach the average temperature of the reactor heat pipe heat exchanger wall and the average temperature of the heat pipe surface at time T3, and then quickly cut off the hot air introduced. p. Phase 1: The time period is (T3—T4) / n, and the DC power supply is adjusted to the actual reactor power at time T3—T4 / n; q. Stage 2: Adjust the DC power supply according to... The change allows for the rapid removal of the removable insulation from the waste heat exhaust device; r. Phase 3: Monitor temperature changes through temperature measuring points. If the temperature measurement value continues to decrease, it indicates that the heat pipe reactor is safe under a coolant loss accident; if the temperature measurement value continues to rise, it indicates that the heat pipe reactor is unsafe under the accident.

10. A method for simulating a heat pipe reactor heat sink loss accident according to claim 5, characterized in that: In step a, T6, the parameters are verified through experiments to confirm that the residual heat removal device can promptly remove the decay heat released by the nuclear fuel in the reactor core; in step a, T7, the parameters are verified through experiments to confirm that the residual heat removal device cannot promptly remove the decay heat released by the nuclear fuel in the reactor core.

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