136 results about "Reactor safety" patented technology

The invention discloses a nuclear reactor core channel analysis method based on two fluid models, which comprises the following steps of: 1, establishing a fluid domain computational grid according to the structure of a nuclear reactor core or an assembly; 2, respectively establishing a control equation set and a heat conduction model of the fuel rod for the continuous gas-liquid phase and the discrete liquid drop phase; 3, considering physical phenomena such as flow heat exchange, resistance pressure drop and turbulence mixing in a reactor core, and implanting a corresponding constitutive model; 4, discretizing the control equation based on a second-order discrete format to obtain a discrete equation set; 5, dividing parallel tasks and parallel regions through an MPI parallel strategy to form a large coefficient matrix of a pressure equation; and 6, solving a discrete equation set and a large coefficient matrix through a pressure coupling semi-implicit algorithm and a stable biconjugate gradient method. According to the method, the key physical parameter distribution of the pressurized water reactor can be more accurately and reasonably calculated, so that technical guidance is provided for safety analysis and engineering design of the reactor.

The embodiment of the invention provides a safety injection system and method, and relates to the field of reactor safety systems. The problem that the water injection flow of a safety injection system is not matched with the flow needed in the reactor water loss accident stage is solved. The safety injection device comprises a pressure-bearing container and a damping device, wherein a safety injection outlet pipe is arranged at the bottom of the pressure-bearing container; the damping device is arranged in the pressure-bearing container, the damping device is provided with a water inlet and a water outlet, and safety injection water in the pressure-bearing container flows through the damping device from the water inlet and then flows to the safety injection outlet pipe from the water outlet. The drainage flow of the safety injection water in the pressure-bearing container is gradually reduced along with the process of the water loss accident, and compared with a pressure-bearing container without the damping, the flow in the initial stage is lower, so that the water quantity of the safety injection bypass can be reduced; the flow in the middle stage can quickly realize refilling of water in the lower cavity of the reactor core and resubmerging of the reactor core; the flow submerged in the reactor core is maintained to be lower in the later period of the accident, the safety injection time is fully prolonged, and safety injection water is effectively utilized.

A rapid inference method for a neutron flux distribution change of a reactor core of a nuclear reactor. In an offline training stage, a transient neutron transport equation is solved by means of a full-order solver, so as to obtain spatial distribution data of neutron flux and a delayed neutron precursor density changing over time, a reduced-order basis and a reduced-order coefficient data set of a training data set are obtained by means of proper orthogonal decomposition, and a reduced-order model is obtained by means of least squares fitting into the reduced-order coefficient data set; and in an online prediction stage, after the reduced-order model is solved on the basis of parameters of a predicted working state, the reduced-order basis is applied to a solved reduced-order coefficient, so as to reconstruct a physical field, which includes a neutron flux distribution of a reactor core, a delayed neutron precursor density distribution, etc. changing over time. In the present invention, data of a high-fidelity full-order model is used to construct a rapid prediction model, and a high-resolution change in the neutron flux of a reactor core over time is finely and efficiently described in a neutronics simulation, which plays an important role in reactor safety analysis, reactor data assimilation and inverse problem solving.

The invention is suitable for the technical field of nuclear reactor safety of a nuclear power plant, and provides an efficient nuclear reactor safety margin evaluation method and system, terminal equipment and a computer program product, and the method comprises the following steps: constructing a training data set based on a sample data set of uncertainty parameters and time sequence data of key safety parameters; training the initial agent model based on the training data set to obtain a target agent model; the sequence overrun probability of the key safety parameters is efficiently calculated through the target agent model; and calculating the shutdown risk of the nuclear reactor based on the sequence overrun probability and taking the shutdown risk as a safety margin assessment result of the nuclear reactor. According to the method provided by the invention, the sequence overrun probability of the key safety parameters is efficiently calculated, and the shutdown risk of the nuclear reactor is calculated based on the sequence overrun probability and is used as the safety margin evaluation result of the nuclear reactor, so that the safety margin of the nuclear reactor is efficiently, quickly and accurately evaluated.

To suppress horizontal vibration of a nuclear pressure vessel during an earthquake, and make a dynamic vibration absorber compact to suppress an installation space of the dynamic vibration absorber, in a nuclear power plant.SOLUTION: An earthquake resistance support device for a nuclear power plant is arranged in a reactor containment vessel, and suppresses horizontal vibration of a reactor pressure vessel supported by a pedestal when an earthquake occurs. The earthquake resistance support device comprises a horizontal beam that is arranged between the reactor containment vessel and the reactor pressure vessel, and the pedestal, and extends in the radial direction of the reactor pressure vessel and the pedestal; and a dynamic vibration absorber installed on the horizontal beam. The dynamic vibration absorber comprises a hollow weight extrapolated to the horizontal beam and movable in an axial direction of the horizontal beam, and a coil spring device that elastically supports the hollow weight in the axial direction of the horizontal beam, to allow the hollow weight to vibrate in the axial direction of the horizontal beam.SELECTED DRAWING: Figure 3

The invention relates to the field of nuclear reactor safety instrument control, in particular to a module-level testing device and method for nuclear reactor safety instrument control equipment. The device comprises a box body which is provided with a front panel, a rear panel and a test module; the test module is arranged in the box body and comprises a main controller module, and a storage circuit, an analog electric signal acquisition circuit, an analog electric signal output circuit, an analog resistance output circuit, an impedance measurement circuit, an insulation detection circuit and a power supply circuit which are connected with the main controller module; and the master controller module is also connected with the display screen and the data export interface. Test items can be freely added and modified according to the to-be-tested equipment and test requirements, all modules of the nuclear reactor safety instrument control equipment are comprehensively tested, high-integration and modular design can be achieved, the portability of the equipment is improved, and the test efficiency and accuracy are improved.

The embodiment of the invention relates to the field of processes specially used for manufacturing a reactor or parts thereof, in particular to an assembling and charging testing method of a reactor using bulk nuclear fuel and the reactor comprising the bulk nuclear fuel. According to the method, loading test is carried out on the reactor by using the test facility after assembly, so that the backup reactivity of the reactor core can be determined, safe assembly and safe criticality of the reactor are realized, and the safety of experimenters is ensured; by gradually adding the partitioned side reflecting layers outside the reactor core formed by the bulk nuclear fuel, the reactor gradually reaches the critical value, the operation risk is reduced, and meanwhile, the critical point can be accurately obtained; after the critical loading of the reactor core is determined, the partitioned side reflecting layers are replaced by the complete side reflecting layers with the same value, so that the influence of gaps existing between the partitioned side reflecting layers can be avoided, and the reaction efficiency of nuclear fuel is improved. The reactor assembled according to the method has high neutron economy and low critical risk.

The invention relates to the technical field of lead-bismuth reactors, and particularly discloses a liquid lead-bismuth alloy thermal stripe oscillation experiment device and method. The device comprises an experiment loop, a gas protection loop, a data acquisition device and an integrated control module. According to the experimental loop, a magnetic pump is used for driving lead-bismuth fluid and an air cooler to realize cooling through a cold and hot double-loop structure, lead-bismuth jet flow with temperature difference is formed, and the lead-bismuth jet flow is mixed in a three-inlet experimental section to simulate a hot stripe oscillation phenomenon. The method comprises the steps of gas replacement, preheating, loop filling, temperature and flow rate regulation and control, data acquisition and the like. The device can truly simulate a lead bismuth alloy thermal stripe oscillation process, effectively prevents high-temperature oxidation, has a modular experimental section structure, can research thermal stress responses under different structures, and provides reliable experimental data for reactor safety design.

Device and Method for Maintaining a Pressure Reactor Containment The present invention relates to a maintenance device (10) for a pressure reactor containment, comprising a connecting fitting (120) on a wall (20) of the containment, said connecting fitting (120) extending in a longitudinal direction (X) between an internal base (55) and an external base (60) and comprising: a) an insertion portion (65) intended to be positioned in a receiving recess (90) machined in the wall (20) of the containment, b) an external portion (70) intended to be positioned outside (30) of the containment and comprising a support element (105) intended to be placed opposite the wall (20) of the containment on a periphery (P) of the receiving recess (90), a fluid conduit (75) extending through the connecting fitting (120), a groove (88) of reception of a compression seal (100) being provided in the insertion portion,the maintenance device (10) comprising means for fixing said connecting branch (120) to the wall (20). Figure for the abbreviation: 4,

The present application relates to the technical field of reactors, and particularly relates to a calculation method of corrosion product deposition rate of a pressurized water reactor. The calculation method of corrosion product deposition rate of the pressurized water reactor comprises the following steps: obtaining the concentration distribution and Zeta potential distribution of metal ions related to corrosion products; obtaining the electric field force on the particle state corrosion products based on the Zeta potential distribution; tracking the transport trajectory of the particle state corrosion products by using a discrete phase model, wherein the force on the particles in the transport process comprises the electric field force; calculating the particle mass flow rate according to the discrete phase model; and calculating the deposition rate and erosion rate of the particle state corrosion products on the wall surface according to the particle mass flow rate. The present application can be used to solve the problem that the current calculation model cannot accurately predict the CRUD deposition rate, and can improve the authenticity of CRUD growth prediction and the reliability of reactor safety analysis.

To cool a molten corium flowing out of a reactor vessel at an early stage.SOLUTION: A molten corium cooling device, in which a structure installed on a foundation portion of a reactor containment vessel supports a reactor and which cools a molten corium falling from a lower portion of a reactor vessel, includes: a cavity provided in a hollow shape in the foundation portion; a water storage tank provided in the cavity and storing cooling water; a heat transfer member that receives the molten corium and transfers heat of the molten corium to the cooling water in the water storage tank; and a jet nozzle that jets the cooling water in the water storage tank above the heat transfer member.SELECTED DRAWING: Figure 2

The invention relates to the technical field of reactor safety analysis, in particular to a test device for simulating steam-water two-phase jet flow impact under a cold loss accident, which comprises a blow-off container provided with a test cavity, the top of the test cavity is provided with a blow-off nozzle, and the blow-off nozzle is arranged downwards and is used for connecting an upstream stagnation system and a simulation crevasse; the mounting assembly is mounted in the test cavity, is positioned below the spraying nozzle and is used for mounting a test piece; and the trash holding assembly is mounted in the test cavity, is positioned below the mounting assembly, and can collect fragments impacted by the steam-water two-phase jet flow from the test piece. The testing device for simulating steam-water two-phase jet flow impact under the cold loss accident can carry out crevasse impact tests on materials such as heat preservation, plugging, fire prevention, painting and cables, and provides test data support in the aspect of disintegrating slag source items for upstream effect analysis of a pit filter.

A nuclear reactor protection system includes a plurality of functionally independent modules, each of the modules configured to receive a plurality of inputs from a nuclear reactor safety system, and logically determine a safety action based at least in part on the plurality of inputs, each of the functionally independent modules comprising a digital module or a combination digital and analog module, an analog module electrically coupled to one or more of the functionally independent modules, and one or more nuclear reactor safety actuators communicably coupled to the plurality of functionally independent modules to receive the safety action determination based at least in part on the plurality of inputs.

The present disclosure provides a system for implementing a logic configuration for a bursting disk burst alarm in a primary-loop compartment of a high-temperature gas-cooled reactor, including: a trigger condition module for setting a first trigger signal and an automatic cycle trigger signal based on the current time-series pressure measurement point value; a pressure measurement point value assignment module for assigning the current time-series pressure measurement point value to a temporary pressure value based on the first trigger signal and the automatic cycle trigger signal; multiple pressure measurement point value retention modules, each of which is used to assign and retain temporary pressure values ​​of different time sequences based on the first trigger signal and the automatic cycle trigger signal, thereby temporarily storing multiple temporary pressure values ​​of different time sequences; and a bursting disk pressure alarm module for determining whether to issue a bursting disk pressure alarm based on the temporary pressure values ​​of different time sequences. The logic configuration implementation system of the present disclosure implements the DCS's cyclic calculation of the bursting disk burst alarm, enabling main control room personnel to conveniently and promptly monitor whether the bursting disk pressure exceeds the limit, thereby enhancing reactor safety.

To provide a concrete structure capable of preventing foreign matter from entering a steam exhaust pipe during concrete placing as well as after completion, and reactor containment vessel using the concrete structure.SOLUTION: A concrete structure comprises a concrete skeleton 2, a steel plate (an inner steel plate) 3 stretched on a surface of the concrete skeleton 2 via studs 8, and a steam exhaust pipe 10 buried inside the concrete skeleton 2 for exhausting steam generated from the concrete skeleton 2 when the concrete skeleton 2 is exposed to high temperatures. The steam exhaust pipe 10 includes one end disposed at a boundary part between the concrete skeleton 2 and the steel plate 3, and the other end disposed outside the concrete skeleton 2, and the steam exhaust pipe 10 includes a steam introduction mechanism 11 at least at one end for introducing steam into the steam exhaust pipe 10.SELECTED DRAWING: Figure 1

The invention provides a method, system, equipment and medium for identifying the number and size of reactor fuel cladding damage, and the method comprises the steps: S1, obtaining the unit crevasse release amount reference data of each nuclide based on transient release modeling experiments of different groove sizes; s2, obtaining the total release amount of nuclides based on the obtained unit crevasse release amount reference data of each nuclide; and S3, constructing a matrix reversibility constraint, and realizing the unique solution inversion of the damage parameter. The method can improve the accuracy and efficiency of fuel cladding damage detection, provides reliable method support for safe operation of a reactor, and can be widely applied to monitoring and evaluation of the safety of the fuel assembly of the pressurized water reactor nuclear power station.

The present disclosure relates to the technical field of nuclear reactor safe operation assessment, proposes a method and system for hydrogen risk analysis in a steel containment vessel with water film cooling, constructs a water film cooling simulation model for simulating the heat and mass transfer process of the external liquid drop model of the containment vessel, transmits the simulation results to a three-dimensional hydrogen risk analysis CFD program for coupling, realizes the supplement of the CFD program, and can be applied to the three-dimensional hydrogen risk analysis of passive nuclear power plants, effectively improving the accuracy of hydrogen risk analysis after nuclear power plant accidents.

The invention belongs to the technical field of nuclear power, and particularly relates to a compensation voltage setting method and device for a compensation ionization chamber, a computer readable storage medium and electronic equipment. The method comprises the following steps: determining a current attenuation multiple of a compensation ionization chamber; based on a preset mapping relation model, determining a compensation voltage increment corresponding to the current attenuation multiple; wherein the mapping relation model is used for representing a mapping relation between the compensation voltage increment and the current attenuation multiple; and setting the compensation voltage according to the compensation voltage increment. According to the invention, the mapping relation model used for representing the mapping relation between the compensation voltage increment and the current attenuation multiple can be constructed in advance, so that the compensation voltage increment corresponding to the current attenuation multiple can be determined based on the mapping relation model, and the compensation voltage is set according to the compensation voltage increment. The measurement precision of the compensation ionization chamber can be effectively improved, and a more reliable decision basis is provided for safety control of a reactor.

The invention discloses a natural circulation system based on in-pipe condensation and an operation method, and the system comprises a cooling water tank which is arranged outside a reactor containment and is filled with cooling water. And the heat exchanger is arranged in the cooling water tank. The inlet end of the air inlet pipeline is connected inside the containment, and the outlet end is connected with the heat exchanger. The inlet end of the drainage pipeline is connected with the heat exchanger, and the outlet end extends into the in-shell water tank. The inlet end of the exhaust pipeline is communicated with the heat exchanger, and the outlet end extends into the suppression pool. And the suppression pool is internally filled with water, is arranged outside the containment, has a pressure difference with the containment, and is communicated with the containment through a low-pressure-difference communication device and a pipeline. And the in-shell water tank is arranged in the safety shell. The inlet end of the gravity injection pipeline is communicated with the in-shell water tank, and the outlet end is connected with the reactor. The problem that an existing pipe external condensation type natural circulation system is not suitable for the low level difference condition is solved, the operation stability of the natural circulation system under the condition is improved, and the starting time of the natural circulation system is shortened.

The invention discloses an advanced safety injection system and a pressurized water reactor and belongs to the field of nuclear power. The advanced safety injection system comprises an integrated pressure water tank, the integrated pressure water tank comprises a CMT cavity and an ACC cavity which are divided by a conical surface wall with the axis vertically extending, the ACC cavity surrounds the outer side of the CMT cavity, the sectional area of the ACC cavity is gradually reduced from top to bottom, and an upper through hole and a lower through hole are formed in the conical surface wall so that the CMT cavity can be communicated with the ACC cavity. Pressure gas is stored at the top of the ACC cavity; a top inlet communicated with a reactor primary loop cooling system is formed in the top end of the CMT cavity, and an injection pipeline for injecting a coolant into the reactor core is arranged at the bottom end of the CMT cavity. The advanced safety injection system is simple in structure, the CMT and ACC functions can be achieved through the single water tank under different accident working conditions, the manufacturing and maintenance cost is effectively reduced, and the reactor safety is improved.

To provide a construction method of a reactor building capable of shortening a construction time, and a roof structure.SOLUTION: A construction method for a reactor building 1 comprises an outside skeleton 3 that houses a reactor containment vessel 2 and a crane 50 installed at an inside upper end part of the skeleton 3. The construction method includes an assembly process in which a skeleton wall 32a, which is an upper end part of a wall of the skeleton 3, the crane 50, and a support structure for the crane 50 are integrated into a crane structure 7 in an assembly area, and an installation process in which the crane structure 7 integrated in the assembly area is transported to a construction position and installed. In the assembly process, the crane structure 7 and a skeleton roof (skeleton roof part 33), which is the roof of the skeleton 3, are integrated in the assembly area. In the installation process, a roof structure 6 consisting of the crane structure 7 and the skeleton roof (skeleton roof portion 33) are transported to a construction position and installed.SELECTED DRAWING: Figure 1

The present application discloses a reactor safety system combining active and passive functions and a control method thereof. The reactor safety system includes a safety injection subsystem combining active and passive functions, a containment energy management subsystem, a secondary side heat removal subsystem, and a reactor cavity water injection subsystem; wherein the safety injection subsystem includes two DVI safety injection pipelines and two independently arranged safety injection sub-trains, each DVI safety injection pipeline includes a DVI pipe, which receives the safety injection flow from a safety injection pump in each of the two safety injection sub-trains, and injects the safety injection flow into the reactor through the DVI pipe after being cooled by a series-connected safety injection heat exchanger; the control method includes: for a large breach accident, under a single fault criterion, providing the reactor with the safety injection flow of at least two safety injection boxes and two safety injection pumps, with the safety injection boxes providing a large short-term safety injection flow in the early stage, and the safety injection pumps providing a continuous safety injection flow thereafter; for other accidents, at least one DVI safety injection pipeline is effective.

The invention relates to the technical field of reactor safety analysis, and provides a pressurized water reactor pressure vessel lower chamber flow distribution analysis method. The method comprises the following steps: establishing a fluid domain three-dimensional geometric model; performing unstructured grid division on the fluid domain three-dimensional geometric model to generate a three-dimensional calculation grid; selecting a turbulence model; setting inlet boundary conditions and physical property parameters of the coolant; the reactor core area is equivalent to a porous medium area, and flow resistance parameters of the porous medium area are set; based on the three-dimensional calculation grid, adopting a selected turbulence model, loading flow resistance parameters of a porous medium area and inlet boundary conditions and physical property parameters of a coolant, and carrying out three-dimensional flow simulation to obtain coolant flow distribution at the section of a reactor core inlet; and evaluating the flow distribution uniformity of the reactor core inlet according to a preset flow distribution evaluation criterion. The method is used for simulating three-dimensional flow of a coolant in a lower cavity of a reactor pressure vessel, acquiring a flow field structure and flow distribution data of the coolant, and performing quantitative evaluation.

The invention relates to the technical field of reactors, in particular to a method for calculating the deposition rate of a pressurized water reactor corrosion product. The method for calculating the deposition rate of the corrosion product of the pressurized water reactor comprises the following steps: acquiring concentration distribution and Zeta potential distribution of metal ions related to the corrosion product; acquiring an electric field force applied to the particulate corrosion product based on the Zeta potential distribution; a discrete phase model is adopted to track the transportation track of the particle-state corrosion product, and the force borne by particles in the transportation process comprises the electric field force; according to the discrete phase model, calculating to obtain a particle mass flow rate; and according to the particle mass flow rate, the deposition rate and the erosion rate of the particle-state corrosion product on the wall surface are calculated. The method can be used for relieving the problem that a current calculation model cannot accurately predict the CRUD deposition rate, and the authenticity of CRUD growth prediction and the credibility of reactor safety analysis can be improved.

An integrated PWR nuclear reactor that requires no refueling and is cooled by natural convection. It is housed within a reactor containment unit which is transportable to a placement site and placed within an underground building that allows heat transfer from the reactor to the surrounding soil. Tire reactor containment unit encloses an integrated PWR nuclear reactor, a pressure and chemical control system, a safety system and a turbo generator. After its operational life, the reactor's water and air are replaced with inert gases, circulating by natural convection to remove decay heat to the surrounding soil. This enables the integrated PWR reactor, initially designed to generate electricity, to act as a dry storage facility for uranium fuel elements after the end of its operational life.

The present application is applicable to the field of nuclear reactor safety technology in nuclear power plants, and provides an efficient nuclear reactor safety margin assessment method, system, terminal device, and computer program product, including: constructing a training data set based on a sample data set of uncertainty parameters and time series data of key safety parameters; training an initial proxy model based on the training data set to obtain a target proxy model; efficiently calculating the sequence exceedance probability of the key safety parameters through the target proxy model; calculating the shutdown risk of the nuclear reactor based on the sequence exceedance probability and using it as the safety margin assessment result of the nuclear reactor. The method provided in the present application achieves efficient, rapid, and accurate assessment of the safety margin of the nuclear reactor by efficiently calculating the sequence exceedance probability of the key safety parameters and calculating the shutdown risk of the nuclear reactor based on the sequence exceedance probability as the safety margin assessment result of the nuclear reactor.

The present invention provides an online assessment method for the safety risk of fully mixed flow reactor operation. The method is a risk assessment for the dynamic operation of exothermic reactions. The complete set of work includes defining the scope, collecting safety data, determining safety conditions and critical limits, and finally determining the coefficient indicating the safety deviation during the operation, that is, the real-time safety total index of the reactor. The method provides a real-time safety total index for evaluating the reactor temperature, while taking into account the reactor safety risk caused by the increase in reaction temperature and the reactor concentration accumulation safety risk caused by the decrease in reaction temperature, providing a warning for the operation of the reactor. The evaluation method is simple to implement. Compared with the traditional design-type reactor safety risk assessment, it can be used for the online assessment of the continuous operation safety risk of fully mixed flow reactors for exothermic reactions.

The invention belongs to the field of nuclear materials, and particularly relates to a nuclear fuel and a preparation method thereof and a reactor, the nuclear fuel comprises a nitride matrix and an MAX phase material distributed in the nitride matrix; the chemical formula of the MAX phase material is Mn + 1AXn, M is selected from at least one of Ti and Cr, A is selected from at least one of Si and Al, X is selected from C, and n is 1 or 2. According to the nuclear fuel disclosed by the invention, the MAX phase material is introduced into the nitride matrix, and a corrosion-resistant oxide layer is generated in situ by utilizing the MAX phase material to block water permeation, so that the nuclear fuel has the properties of high thermal conductivity, relatively strong water corrosion resistance, steam oxidation resistance and radiation resistance, and can be applied to a light water reactor; the application risk of the existing uranium nitride fuel in the light-water reactor is reduced, and the safety and economy of the reactor are improved.

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.

The invention is applicable to the field of nuclear power stations, and provides a pressure test system for a reactor containment, which comprises an internal transmission device, a regulating valve and exhaust equipment, the internal transmission device is arranged in the reactor containment, is connected with the regulating valve through a first pipeline and is used for transmitting gas in the reactor containment to the regulating valve; the adjusting valve is arranged outside the reactor containment, connected with the exhaust equipment through a second pipeline and used for guiding the gas to be transmitted to the exhaust equipment through the second pipeline by adjusting the opening degree of the valve to the set opening degree, and the designed flow range of the gas transmitted between the first pipeline and the second pipeline is 0-32,400 m < 3 > / h; and the exhaust equipment is used for exhausting gas. According to the pressure test system, the valve opening degree of the regulating valve is regulated to the set opening degree to guide gas to be quickly transmitted to the exhaust equipment, so that the pressure relief rate of the gas in the containment is greatly improved, the duration required by a CTT test is shortened, and the operation efficiency of a unit is improved.