Calculation method for gas compression and temperature rise in main steam line of nuclear reactor

Abstract

A computational method for investigating the gas compression and temperature rise phenomenon in the main steam pipe of a nuclear reactor is proposed. This method is applicable to simulating the abnormal temperature and pressure rise in the secondary steam pipe of the steam generator caused by the loss of secondary feedwater and the closure of the main steam isolation valve in a full-pressure small reactor. The method mainly includes the following parts: 1. Simplifying the geometry of the secondary steam main pipe and constructing a 3D model; 2. Using SpaceClaim software to perform 3D modeling of the simplified steam main pipe structure; 3. Using the mesh function of Fluent software to generate a computational mesh for the model; 4. Establishing suitable compressible flow and turbulence models; 5. Defining boundary conditions and initial conditions; 6. Setting the solver and solution method, and setting the time and spatial discretization formats; 7. Adjusting the sub-relaxation factor and setting the time step that conforms to the computational time scale; 8. Setting the convergence criterion and monitoring physical quantities such as cross-sectional temperature and inlet pressure until the pressurization process in the 3D model ends and the container reaches a steady state.

Description

Technical Field

[0001] This invention relates to the steam compression phenomenon in the secondary side main steam pipe of a small reactor steam generator, specifically to a numerical simulation calculation method for investigating the steam compression and temperature rise phenomenon in the main steam pipe of a nuclear reactor. Background Technology

[0002] Fully pressurized marine small pressurized water reactors typically employ a compact design, with their steam generators and secondary main steam systems exhibiting significant differences in structural dimensions and thermal parameters compared to conventional large pressurized water reactors. Under normal operating conditions, the secondary side of the steam generator continuously supplies cooling water to the tube bundle region via the main feedwater system to maintain the temperature and pressure within the main steam system within the design range. However, in the event of a secondary main feedwater loss accident, the main feedwater is interrupted and the main steam isolation valve closes. Until new cooling water is reintroduced, the secondary side system of the steam generator will be in a relatively closed, heated state, and the steam temperature and pressure within the secondary main steam pipes and steam bridges will continue to rise over time.

[0003] Regarding the aforementioned accident conditions, existing system-level calculations and analyses indicate that during the period of main feedwater loss accompanied by the closure of the main steam isolation valve, the overall temperature variation trend of the primary and secondary fluids in the steam generator tube bundle region still conforms to conventional thermal laws. However, in local sections such as the main steam pipeline and steam bridge, the steam temperature may exhibit significant anomalies, with values ​​significantly higher than the primary side inlet coolant temperature, and even exceeding the temperature resistance limits of relevant structural materials for a short period. As new cooling water is reinjected into the secondary side system, these abnormal temperature phenomena gradually disappear, and the system temperature returns to the normal operating range.

[0004] The occurrence of abnormally high steam temperature in the main steam pipeline has a significant impact on the structural integrity and operational safety of the secondary side system of a fully pressurized marine small modular reactor (SMR). On the one hand, short-term high temperatures may lead to degradation of the mechanical properties of the pipeline materials, increasing the risk of structural failure. On the other hand, if such temperature anomalies are not fully identified and assessed, they may adversely affect the safety analysis conclusions under accident conditions. Therefore, it is necessary to conduct in-depth research on the formation mechanism of abnormally high steam temperature in the main steam pipeline under the accident condition of loss of secondary side main feedwater.

[0005] From a physical process perspective, under conditions of closed main steam isolation valve, limited system volume, and continuous heating, the steam in the main steam pipeline may undergo a significant unsteady compression process. This compression process is accompanied by rapid changes in steam density, pressure, and temperature, and occurs over a short timescale, with flow and heat transfer behaviors significantly different from steady-state or quasi-steady-state conditions. Simultaneously, due to the large wall thickness and long heat conduction path of the main steam pipeline, the heat exchange between the steam and the external environment is relatively limited during the short-term compression process. Therefore, the steam temperature rise is primarily controlled by the internal compressibility effect of the fluid, resulting in a complex thermal response mechanism.

[0006] Currently, research on the thermal behavior of secondary steam systems under nuclear reactor accident conditions mainly focuses on the overall performance of steam generators, system pressure response, and safety valve operating characteristics. There is relatively little research in the published literature on the localized temperature anomalies caused by unsteady steam compression processes within the main steam pipeline and steam bridge. Existing analytical methods are mostly based on system-level programs or simplified models, which insufficiently consider the spatial distribution characteristics, temporal evolution, and wall heat transfer effects of the steam compression and temperature rise process within local pipe sections, making it difficult to accurately reflect the true thermal response within the main steam pipeline.

[0007] In summary, under the accident condition of loss of main feedwater on the secondary side of a fully pressurized marine small modular reactor (SMR), the abnormal temperature rise caused by steam compression in the main steam pipeline has a clear engineering background and safety significance. However, its formation mechanism and influencing factors still lack systematic and in-depth research. Therefore, it is necessary to conduct analysis and research on the unsteady steam compression and temperature rise process in the pipeline to provide technical basis for the safety assessment and design optimization of the secondary side system under relevant accident conditions. Summary of the Invention

[0008] To address the problem that existing analytical methods cannot accurately characterize the abnormal temperature rise caused by steam compression in the main steam pipeline under the condition of loss of secondary main feedwater, this invention proposes a numerical simulation calculation method for investigating the steam compression and temperature rise phenomenon in the pipeline.

[0009] To achieve the above objectives, the present invention adopts the following technical solution: A calculation method for investigating the gas compression and temperature rise phenomenon in the main steam pipe of a nuclear reactor includes the following steps: Step 1: Design a simplified geometric model suitable for investigating the steam compression and temperature rise phenomenon in the main steam pipe of a nuclear reactor, including the inlet section and the main steam pipe; Step 2: For the simplified geometric model, the SpaceClaim software is used to sequentially establish three-dimensional geometric models of the inlet section fluid domain, the inlet section wall, the main steam pipe fluid domain, and the main steam pipe wall. Step 3: Use the mesh function of Fluent software to generate mesh nodes for the 3D geometric model, adjust the mesh size for the inlet section and all solid domains, add a boundary layer to the wall of the main steam pipe, and generate mesh nodes using the poly-hexcore method; Step 4: Establish compressible flow and turbulence models suitable for numerical simulation of steam compression and heating phenomena in the main steam pipes of nuclear reactors; Step 5: Set boundary conditions and initial conditions for the model established in Step 4, using a custom function for pressure inlet; Step 6: Select a split solver. Use a second-order upwind scheme for spatial discretization of density, momentum and turbulent kinetic energy gradient variables in the model solution process, and a first-order implicit discretization method for time discretization. Use the PISO method for pressure-velocity coupling calculation. Step 7: Adjust the sub-relaxation factor to control the computational convergence problem caused by updating the boundary conditions in each step, and adjust the Courant number to improve computational accuracy and efficiency; Step 8: Set convergence criteria, monitor cross-sectional temperature and inlet pressure until the pressurization process in the three-dimensional geometric model ends and the pressure and temperature inside the main steam pipe of the nuclear reactor reach a stable state.

[0010] Preferably, in step 1, a simplified geometric model is designed to investigate the steam compression phenomenon in the main steam pipe of a nuclear reactor. The simplified geometric model includes a cylindrical inlet section nozzle section model, a cylindrical steam main pipe test section model, and thermocouple temperature measuring points that may be arranged in different ways inside the cylindrical steam main pipe; the simplified geometric model is a combined structure composed of the nozzle section and the test section connected horizontally in parallel along the axial direction; the temperature and pressure changes caused by steam compression under different inlet conditions in the test section are studied.

[0011] Preferably, in step 3, different local dimensions are added to the cylindrical inlet section, the solid domain of the main steam pipe wall, and the possible thermocouple structure according to different geometric structure dimensions, based on the detailed dimensions of the geometric body. This makes the compressible flow calculation during the compression process more accurate, accelerates the calculation convergence, and satisfies the minimum number of nodes for heat transfer in the solid domain at different geometric local details. In step 3, a boundary layer mesh is added to the pipe wall of the fluid domain of the main steam pipe test section to make the mesh near the wall denser, thereby reducing the impact of the thermal boundary gradient on the calculated heat transfer and enhancing the accuracy of the fluid-solid coupled surface flow heat transfer calculation.

[0012] Preferably, in step 3, since the fluid undergoes large-scale velocity and density changes during compression flow, mesh quality is particularly important for computational convergence and accuracy. The poly-hexcore method is used in the mesh function of Fluent software to divide the computational domain into mesh nodes, forming a hexahedral mesh with small computational error in the central region of the fluid domain, while forming a hexagonal mesh at the boundary and interface. The minimum orthogonal quality is checked to ensure correct computation and convergence.

[0013] Preferably, in step 4, compressible flow is the key to calculating the vapor compression phenomenon. Real gas parameters and variable density methods are used, and a property table is established to determine the density, thermal conductivity and specific heat of the vapor through interpolation methods. In step 4, a Realizable k-epsilon turbulence model suitable for pipe flow calculations was adopted, and standard wall functions were used.

[0014] Preferably, in step 5, boundary conditions are necessary conditions for numerical simulation calculations, while the compression process is an unsteady flow and therefore requires initial conditions; there are countless calculations that obey the same set of differential equations, and boundary conditions and initial conditions are single-valued conditions for the flow and also conditions for flow similarity. In the phenomenon of steam compression in the main steam pipe of a nuclear reactor, the Fourier number is small and the compression process is short. Therefore, the steam pipe structure heats up slowly and the outer wall temperature is low. During this process, the heat exchange with the external environment is minimal, and the outer wall of the solid domain is considered to be an adiabatic boundary. However, the heat exchange between the steam and the wall on the inner surface is relatively intense, and the heat exchange on the inner surface has a significant impact on the steam temperature. Therefore, the inner wall is a fluid-solid coupled heat transfer boundary. The main steam pipeline maintains normal operating pressure and temperature before the compression process, which serves as the initial conditions for the calculation.

[0015] Preferably, in step 5, the inlet boundary conditions conform to the pressure change characteristics over time during the process of secondary side pressure rise caused by continuous heating of the steam generator; a custom function is used to construct the inlet pressure change function over time, and the inlet boundary conditions are defined.

[0016] Preferably, in step 7, since the inlet pressure is updated in real time with the time step and the pressure rise rate is large, it is easy to diverge during the calculation process; in order to ensure the convergence of the calculation while taking into account the calculation efficiency, the sub-relaxation factor is adjusted to 0.3 for pressure, 1 for density, 1 for volume force, 0.7 for momentum, 0.8 for turbulent kinetic energy, 0.8 for turbulent diffusivity, and 1 for turbulent viscosity.

[0017] Preferably, in step 7, the magnitude of the Courant number indicates the convergence speed of the differential equation. When the Courant number is close to 1, the convergence speed of the differential equation is the fastest. Based on the Courant number, the time step is changed according to the change of gas phase velocity, so the calculation time is different under different working conditions. In the calculation, the Courant number range of 0.2 to 0.4 is usually used to balance the calculation accuracy and calculation efficiency.

[0018] Preferably, in step 8, the inlet pressure is set to rise to a fixed pressure and then the pressure increase stops. After this time, the steam temperature and pressure in the main steam pipe of the nuclear reactor gradually stabilize over time. The internal cross-section, measuring point temperature, and inlet pressure of the main steam pipe of the nuclear reactor are detected and judged to determine that the gas temperature inside the main steam pipe of the nuclear reactor eventually reaches a stable state.

[0019] The beneficial effects of this invention are: 1. This invention can study the steam compression and temperature rise phenomenon in the main steam pipeline through numerical simulation without relying on experimental conditions, providing an effective means for the safety analysis of the main steam system under the condition of loss of secondary main feedwater.

[0020] 2. By reasonably simplifying the geometry of the main steam pipe of the nuclear reactor, this invention significantly reduces the difficulty of modeling and calculation while ensuring calculation accuracy, and improves the feasibility and calculation efficiency of numerical simulation.

[0021] 3. The numerical simulation method proposed in this invention is not limited to a specific main steam pipeline structure. It can be used to analyze other pipeline systems with unsteady steam compression processes after changing the geometric model and boundary conditions, and has good applicability. Attached Figure Description

[0022] Figure 1 This is a flowchart of the method of the present invention.

[0023] Figure 2 This is a geometric model diagram of the test section. Detailed Implementation

[0024] This invention provides a numerical simulation method for investigating the phenomenon of steam compression and temperature rise in pipelines. It is applicable to investigating temperature anomalies caused by steam compression in main steam pipelines and steam bridges, and provides an effective numerical simulation scheme for the phenomenon of continuous temperature and pressure increases in the secondary steam pipelines of a steam generator after the loss of secondary feedwater and the closure of the main steam isolation valve in a small-scale steam generator. Figure 1 As shown, this embodiment takes the main steam pipeline as the research object, and specifically includes the following steps: Step 1: Design a simplified geometric model suitable for investigating the steam compression and heating phenomenon in the main steam pipe of a nuclear reactor, including an inlet section and a main steam pipe. The inlet section is used to simulate the secondary side pressure rise caused by continuous heating of the steam generator, and the test section of the main steam pipe is used to study the process of steam compression and heating within a confined volume. To highlight the dominant physical process of steam compression and heating and reduce computational complexity, the structure of the main steam pipe is reasonably simplified so that the model can reflect the temperature and pressure changes of steam in the main steam pipe of the nuclear reactor under different inlet conditions.

[0025] The simplified geometric model includes a cylindrical inlet section pipe section model, a cylindrical steam main pipeline test section model, and thermocouple temperature measuring points that may be arranged differently inside the cylindrical steam main pipeline; the simplified geometric model is a combined structure composed of the pipe section and the test section connected horizontally in parallel along the axis; the study investigates the temperature and pressure changes caused by steam compression under different inlet conditions in the test section.

[0026] Step 2: For the simplified geometric model, use SpaceClaim software to sequentially establish three-dimensional geometric models of the inlet section fluid domain, inlet section wall, main steam pipe fluid domain, and main steam pipe wall; the three-dimensional geometric model is as follows: Figure 2 As shown, the solid region includes the inlet section wall, the main steam pipe wall, and the thermocouple structure; the fluid region includes the inlet section fluid domain and the main steam pipe fluid domain; the geometric model has only one inlet and no outlet; providing a geometric basis for subsequent fluid-solid coupled heat transfer calculations.

[0027] Step 3: Use the mesh function of Fluent software to generate mesh nodes for the 3D geometric model. According to different geometric structure dimensions, set different local mesh sizes for the fluid domain of the cylindrical inlet section, the fluid domain of the main steam pipe, and their corresponding solid walls. Add a boundary layer mesh near the inner wall of the main steam pipe to enhance the accuracy of near-wall flow and heat transfer calculations. At the same time, use the poly-hexcore method to generate meshes for the computational domain, forming hexahedral meshes with smaller calculation errors in the central region of the fluid domain, and hexagonal polyhedral meshes in the boundary and interface regions. Check the minimum orthogonal mass and other indicators to ensure that the mesh quality meets the numerical calculation requirements.

[0028] Step 4: Establish compressible flow and turbulence models suitable for numerical simulation of the main steam compression and temperature rise phenomenon in nuclear reactors. Steam compression is an unsteady compressible flow. To accurately describe the changes in density, pressure, and temperature during steam compression, real gas parameters and the variable density method are used. Thermophysical properties such as steam density, specific heat, and thermal conductivity are determined by establishing a property table and using interpolation. Simultaneously, a Realizable k-ε turbulence model suitable for pipe flow calculations is selected, and standard wall functions are used to handle near-wall flow.

[0029] Step 5: Set boundary and initial conditions for the model established in Step 4, using a custom pressure inlet function. Due to the small Fourier number and short duration of steam compression in the main steam pipeline of the nuclear reactor, and the slow temperature rise of the pipeline structure, the heat exchange between the outer wall of the main steam pipeline and the external environment is relatively small; therefore, the outer wall of the pipeline is set as an adiabatic boundary. The heat exchange between the steam and the inner wall of the pipeline is relatively intense, significantly affecting the steam temperature; therefore, the inner wall is set as a fluid-solid coupled heat transfer boundary. The main steam pipeline maintains normal operating pressure and temperature before the compression process, serving as the initial conditions for the calculation. A pressure inlet is used as the inlet boundary condition, and a custom function is used to construct a function of inlet pressure changing with time to simulate the condition where continuous heating by the steam generator leads to a gradual increase in secondary side pressure.

[0030] Step 6: Select a split solver. Use a second-order upwind scheme for spatial discretization of density, momentum and turbulent kinetic energy gradient variables in the model solution process, and a first-order implicit discretization method for time discretization. Use the PISO method for pressure-velocity coupled calculation. This solution method can improve the stability of unsteady compressible flow calculation while ensuring calculation accuracy.

[0031] Step 7: Adjust the sub-relaxation factor to control the computational convergence issues caused by updating the boundary conditions at each step, and adjust the Courant number to improve computational accuracy and efficiency. During the calculation, the sub-relaxation factor is adjusted to control the computational convergence issues caused by updating the inlet boundary conditions at each step. To ensure stable convergence while maintaining computational efficiency, sub-relaxation factors are set for pressure, density, body force, momentum, and turbulent kinetic energy. The Courant number is used as the time step control criterion; when the Courant number is close to 1, the differential equation converges faster. Since the steam velocity varies under different operating conditions, the time step is dynamically adjusted according to the steam velocity changes while keeping the number of grids constant, ensuring stable convergence of the calculation results.

[0032] Step 8: Set convergence criteria and monitor the cross-sectional temperature and inlet pressure until the pressurization process within the 3D geometric model ends and the main steam pipe model of the nuclear reactor reaches a stable state. The calculation is considered converged when the steam compression pressurization process within the 3D geometric model ends (i.e., the inlet pressure is set to rise to a fixed pressure and pressurization stops), and the steam temperature and pressure within the pipe tend to stabilize over time. This completes the numerical simulation calculation of the steam compression and heating process under this operating condition.

[0033] The above description is merely a preferred embodiment of the present invention, used to illustrate the technical solution of the present invention, and is not intended to limit the scope of protection of the present invention. For those skilled in the art, equivalent substitutions or simple modifications made to the above embodiments without departing from the concept of the present invention should be considered to fall within the scope of protection defined by the claims of the present invention.

Claims

1. A calculation method for investigating the gas compression and temperature rise phenomenon in the main steam pipe of a nuclear reactor, characterized in that, Includes the following steps: Step 1: Design a simplified geometric model suitable for investigating the steam compression and temperature rise phenomenon in the main steam pipe of a nuclear reactor, including the inlet section and the main steam pipe; Step 2: For the simplified geometric model, the SpaceClaim software is used to sequentially establish three-dimensional geometric models of the inlet section fluid domain, the inlet section wall, the main steam pipe fluid domain, and the main steam pipe wall. Step 3: Use the mesh function of Fluent software to generate mesh nodes for the 3D geometric model, adjust the mesh size for the inlet section and all solid domains, add a boundary layer to the wall of the main steam pipe, and generate mesh nodes using the poly-hexcore method; Step 4: Establish compressible flow and turbulence models suitable for numerical simulation of steam compression and heating phenomena in the main steam pipes of nuclear reactors; Step 5: Set boundary conditions and initial conditions for the model established in Step 4, using a custom function for pressure inlet; Step 6: Select a split solver. Use a second-order upwind scheme for spatial discretization of density, momentum and turbulent kinetic energy gradient variables in the model solution process, and a first-order implicit discretization method for time discretization. Use the PISO method for pressure-velocity coupling calculation. Step 7: Adjust the sub-relaxation factor to control the computational convergence problem caused by updating the boundary conditions in each step, and adjust the Courant number to improve computational accuracy and efficiency; Step 8: Set convergence criteria, monitor cross-sectional temperature and inlet pressure until the pressurization process in the three-dimensional geometric model ends and the pressure and temperature inside the main steam pipe of the nuclear reactor reach a stable state.

2. The calculation method for investigating the gas compression and temperature rise phenomenon in the main steam pipe of a nuclear reactor according to claim 1, characterized in that: In step 1, a simplified geometric model is designed to investigate the steam compression phenomenon in the main steam pipe of a nuclear reactor. The simplified geometric model includes a cylindrical inlet section nozzle section model, a cylindrical steam main pipe test section model, and thermocouple temperature measuring points that may be arranged differently inside the cylindrical steam main pipe. The simplified geometric model is a combined structure composed of the nozzle section and the test section connected horizontally in parallel along the axis. The temperature and pressure changes caused by the steam compression phenomenon under different inlet conditions in the test section are studied.

3. The calculation method for investigating the gas compression and temperature rise phenomenon in the main steam pipe of a nuclear reactor according to claim 1, characterized in that: In step 3, based on different geometric dimensions, different local dimensions are added to the cylindrical inlet section, the solid domain of the main steam pipe wall, and the possible thermocouple structure according to the detailed dimensions of the geometry. This makes the compressible flow in the compression process more accurate, speeds up the calculation convergence, and satisfies the minimum number of nodes for heat transfer in the solid domain at different local geometric details. In step 3, a boundary layer mesh is added to the pipe wall of the fluid domain of the main steam pipe test section to make the mesh near the wall denser, thereby reducing the impact of the thermal boundary gradient on the calculated heat transfer and enhancing the accuracy of the fluid-solid coupled surface flow heat transfer calculation.

4. The calculation method for investigating the gas compression and temperature rise phenomenon in the main steam pipe of a nuclear reactor according to claim 1, characterized in that: In step 3, the poly-hexcore method is used in the mesh function of the Fluent software to divide the computational domain into mesh nodes. A hexahedral mesh with small computational error is formed in the central region of the fluid domain, while a hexagonal mesh is formed at the boundary and interface. The minimum orthogonal mass is checked to ensure that the computation can be calculated correctly and converge.

5. The calculation method for investigating the gas compression and temperature rise phenomenon in the main steam pipe of a nuclear reactor according to claim 1, characterized in that: In step 4, compressible flow is the key to calculating the vapor compression phenomenon. Real gas parameters and variable density methods are used, and a property table is established to determine the density, thermal conductivity and specific heat of the vapor through interpolation. In step 4, a Realizable k-epsilon turbulence model suitable for pipe flow calculations was adopted, and standard wall functions were used.

6. The calculation method for investigating the gas compression and temperature rise phenomenon in the main steam pipe of a nuclear reactor according to claim 1, characterized in that: In step 5, boundary conditions are necessary for numerical simulation calculations, and the compression process is an unsteady flow, so initial conditions are required; there are countless calculations that obey the same set of differential equations, and boundary conditions and initial conditions are single-valued conditions for the flow, as well as conditions for flow similarity. In the phenomenon of steam compression in the main steam pipe of a nuclear reactor, the Fourier number is small and the compression process is short. Therefore, the steam pipe structure heats up slowly and the outer wall temperature is low. During this process, the heat exchange with the external environment is minimal, and the outer wall of the solid domain is considered to be an adiabatic boundary. However, the heat exchange between the steam and the wall on the inner surface is relatively intense, and the heat exchange on the inner surface has a significant impact on the steam temperature. Therefore, the inner wall is a fluid-solid coupled heat transfer boundary. The main steam pipeline maintains normal operating pressure and temperature before the compression process, which serves as the initial conditions for the calculation.

7. The calculation method for investigating the gas compression and temperature rise phenomenon in the main steam pipe of a nuclear reactor according to claim 1, characterized in that: In step 5, the inlet boundary conditions conform to the pressure change characteristics over time during the process of secondary side pressure rise caused by continuous heating of the steam generator; a custom function is used to construct the inlet pressure change function over time, and the inlet boundary conditions are defined.

8. The calculation method for investigating the gas compression and temperature rise phenomenon in the main steam pipe of a nuclear reactor according to claim 1, characterized in that: In step 7, to ensure computational convergence while also considering computational efficiency, the sub-relaxation factor is adjusted as follows: pressure is 0.3, density is 1, volume force is 1, momentum is 0.7, turbulent kinetic energy is 0.8, turbulent diffusivity is 0.8, and turbulent viscosity is 1.

9. The calculation method for investigating the gas compression and temperature rise phenomenon in the main steam pipe of a nuclear reactor according to claim 1, characterized in that: In step 7, the magnitude of the Courant number indicates the convergence speed of the differential equation. When the Courant number is close to 1, the convergence speed of the differential equation is the fastest. Based on the Courant number, the time step is adjusted according to the change in gas phase velocity, so the calculation time is different under different operating conditions. The Courant number is used in the calculation in the range of 0.2 to 0.4 to balance the calculation accuracy and calculation efficiency.

10. The calculation method for investigating the gas compression and temperature rise phenomenon in the main steam pipe of a nuclear reactor according to claim 1, characterized in that: In step 8, the inlet pressure is set to rise to a fixed pressure and then the pressure increase is stopped. After this time, the steam temperature and pressure in the main steam pipe of the nuclear reactor gradually stabilize over time. The internal cross-section, measuring point temperature, and inlet pressure of the main steam pipe of the nuclear reactor are detected and judged to determine that the gas temperature inside the main steam pipe of the nuclear reactor finally reaches a stable state.

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