39 results about "Decay heat" patented technology

A liquid metal cooled nuclear reactor includes a passive decay heat removal system having thermal insulation attached to a wall of a cold source reservoir that holds a phase change material, where the insulation is arranged to automatically fall by gravity from the wall in response to the wall reaching a predetermined temperature. The nuclear reactor may be a fast neutron reactor that incorporates an integral system having a final cold source with a reservoir incorporating an integral exchanger divided into a plurality of parallel tubes between which a phase change material is inserted, the reservoir being surrounded by a thermal insulating layer that can be detached in a passive manner in the event of reaching a predetermined threshold temperature.

The application provides a calculation method of a pebble bed high-temperature gas cooled reactor isotope irradiation target heat release rate, which comprises the following steps: constructing a reactor model of the pebble bed high-temperature gas cooled reactor through a special design application to calculate the nuclear fuel composition distribution of the reactor core; performing twice non-uniform arrangement on the fuel distribution of the reactor core through an MC application program to construct a reactor core MC model and transferring the nuclear fuel composition distribution of the reactor core to the reactor core MC model; calculating the energy deposition heat release rate of the isotope irradiation target through a Monte Carlo model for isotope irradiation production and the decay heat release rate of the isotope generated by irradiation; and adding the energy deposition heat release rate and the decay heat release rate to obtain the isotope irradiation target heat release rate of the pebble bed high-temperature gas cooled reactor. The method can accurately calculate the heat release rate of the isotope irradiation target in view of the double non-uniformity and the non-stop reactor refueling characteristics of the pebble bed high-temperature gas cooled reactor.

The method for analyzing the formation and evolution characteristics of a core melt debris bed based on a particle method mainly comprises the following steps: 1, selecting a particle size to adapt to geometric modeling and coolant modeling; 2, selecting a melt material and a coolant material, setting initial parameters, and starting calculation; 3, calculating melt column fragmentation by using a moving particle semi-implicit method basic model; 4, calculating melt particle solidification and combination by using a melting and solidification model; 5, calculating melt debris collision and settlement to form a debris bed by using a discrete element model; 6, calculating debris bed cooling or remelting by using a decay heat model and a boiling heat transfer model; 7, judging whether the end time is reached, if not, advancing by a time step, and if yes, outputting a calculation result; and the method can predict the melt fragmentation form, debris size distribution, debris bed accumulation form, debris bed coolability and debris bed remelting risk probability after a core melt accident.

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.

To provide a decay heat removal system for a gas-cooled nuclear plant and a gas-cooled nuclear reactor that can appropriately remove decay heat in an accident.SOLUTION: A gas-cooled nuclear plant is provided with: a reactor pressure vessel that accommodates a gas-cooled nuclear core; a reactor pressure vessel chamber that accommodates the reactor pressure vessel; a cooling panel that is provided in the reactor pressure vessel chamber to surround the periphery of the reactor pressure vessel; a thermoelectric element that is provided in the reactor pressure vessel chamber; and a fine particle reduction mechanism that is actuated by power generated in the thermoelectric element and reduces fine particles floating in the reactor pressure vessel chamber.SELECTED DRAWING: Figure 1

The application relates to a kind of passive residual heat removal systems of fuel salt discharge tank, heat transfer working medium is solid at room temperature and is liquid after heating, the passive residual heat removal system includes fuel salt discharge tank, salt discharge pipe, natural circulation loop, air cooler and circulation maintaining heat exchange device, fuel salt discharge tank is communicated with nuclear reactor core by salt discharge pipe, fuel salt discharge tank is communicated with air cooler and circulation maintaining heat exchange device respectively by natural circulation loop;Wherein, circulation maintaining heat exchange device is in thermal contact with molten salt reactor main loop pipeline, to compensate the heat loss of passive residual heat removal system by the heat energy of normal working loop.According to the passive residual heat removal system of the application, heat transfer working medium is not water, in the normal operation of reactor, it is standby state, maintain natural circulation fluid in molten liquid state and slowly natural circulation, once accident occurs, the fuel salt of reactor core is discharged into discharge tank, and the natural circulation fluid of molten liquid state can quickly establish natural circulation to remove decay heat.

The invention discloses a high-level radioactive waste deep geological disposal and heat energy recovery collaborative method, which belongs to the field of high-level radioactive waste geological disposal, and comprises the following steps: S1, tunneling a disposal roadway in a complete bedrock of a disposal repository, excavating a disposal hole in a bottom plate of the disposal roadway, and tunneling the disposal roadway in a relatively complete crystal bedrock; s2, a heat insulation film is laid around the hole wall and the bottom of the treatment hole, and the thickness of the heat insulation film is 0.5 cm; s3, constructing a water injection station, a heat exchange pipe and a centralized heat exchange station; s4, filling a gap between the heat exchange pipe and the treatment tank with a bentonite buffer material; and S5, high-level radioactive waste is treated. The heat energy recovery pipe surrounding the high-level radioactive waste disposal hole is constructed on the basis of a traditional crystallized rock vertical hole high-level radioactive waste disposal method, the heat energy recovery pipe can effectively absorb decay heat of the high-level radioactive waste, the situation that the temperature of bed rock of crystallized rock rises and thermal stress is generated is avoided, meanwhile, decay heat recovery of the high-level radioactive waste is achieved, energy is fully utilized, and the energy utilization rate is increased. The carbon neutralization target is met, and wide application and popularization prospects are achieved.

The present invention relates to a method and system for commissioning management of a spent fuel pool cooling system. The method for commissioning management of the cooling system includes: obtaining the three-part decay heat power per unit mass of metallic uranium for each group of spent fuel assemblies; obtaining the mass of each group of spent fuel assemblies and calculating the total decay heat power of the spent fuel pool by using a VBA tool; obtaining a pre-established heat exchange model of the spent fuel pool and determining the inlet temperature limit of the cooling water of the currently commissioned cooling system according to the heat exchange model of the spent fuel pool and the total decay heat power of the spent fuel pool; measuring the inlet temperature of the cooling water of the currently commissioned cooling system to obtain an inlet temperature measurement value, and determining a commissioning plan for the cooling system according to the inlet temperature measurement value and the inlet temperature limit. Implementing the technical solution of the present invention can not only save manpower and time, but also provide scientific opinions on the commissioning of the cooling system for operators.

The invention discloses a passive residual heat removal system and method for a nuclear power plant. The problems that in the prior art, a pure water cooling mode is insufficient in initial heat exchange intensity and limited in long-term operation time, and a pure air cooling mode is poor in heat exchange capacity are solved. The passive residual heat removal system of the nuclear power unit comprises a passive residual heat removal system body, wherein a heat exchanger is used for guiding decay heat of a reactor core into the passive residual heat removal system body; the air guide plate is arranged opposite to the air inlet and used for creating a cold air introduction channel, and an air distribution skirt is arranged on the lower portion of the air guide plate and used for guiding cold air to the heat exchanger; and the sprayer is used for spraying cooling water to the surface of the heat exchanger under a passive condition when an accident occurs. Self-adaptive matching of heat exchange modes driven by decay heat removal requirements in different stages is achieved, and the reliability, safety and long-term operation capacity of a nuclear power unit waste heat removal system are remarkably improved.

Systems limit backflow through a nuclear reactor cores from a hot pool to a cold pool. Example systems include coolant risers and / or one-way flow structures before a coolant inlet for the reactor. The coolant riser provides coolant that will flow under gravity alone into the reactor despite hot pool pressure head. Fluidic diodes and / or check-valves prevent nearly all reverse flow. Such structures may be positioned with the flow limiter before the riser in the normal direction of coolant flow into the inlet. Such structures may be positioned anywhere they can functionally provide coast down cooling and prevent reverse flow, including as a pump annulus or pool. For a sodium-cooled fast reactor, the riser and fluidic diode may provide sufficient volume to cool the reactor with forward coolant flow for an entire reactor coast down period, when the reactor is shut down and generating only decay heat.

A nuclear waste heat generating system (10) comprises nuclear waste storage site (12) containing one or more storage containers (14) comprising radioactive nuclear waste. The heat pump system (16) comprises a closed working fluid circuit (26) filled with a working fluid, the working fluid circuit including an inflow pipe (22) for supplying the working fluid from a heat sink (18) of the heat pump system (16) to a heat exchanger (20) of the heat pump system (16), and an outflow pipe (24) for returning heated working fluid from the heat exchanger (20) to the heat sink (18). The heat exchanger (20) is thermally coupled to the nuclear waste storage site (12) such that the heat exchanger (20) is heated by decay heat produced by the one or more storage containers (14). Further, a use of one or more storage containers comprising radioactive nuclear waste as a heat source is described.

A container for transporting a reactor is disclosed. The container includes a loop thermosiphon including a chamber, a heat exchanger fluidically coupled to the chamber, and an actuator including an unactuated state and an actuated state. The actuator is configured to automatically transition to the actuated state. The transition is based on an event occurring within the reactor. A working medium is configured to remove heat from the reactor in the actuated state.

The application discloses a kind of lithium cold space reactor power system auxiliary thawing loop system and thawing method thereof, lithium cold space reactor power system is composed of one loop system and two loop systems.Lithium cold space reactor power system is thawed by auxiliary loop system, and auxiliary loop system is composed of auxiliary pipeline, auxiliary electromagnetic pump, auxiliary energy conversion unit.After the auxiliary pipeline comes out from reactor core, it is attached to the outer wall of one loop pipeline and two loop pipeline, for heating coolant pipeline.When thawing starts, core is operated at low power level;The heat generated is taken out of the core by the auxiliary loop system, heats up each component of lithium cold space reactor power system along the auxiliary pipeline, melts lithium working medium in each component, and after one loop system and two loop systems are completely thawed, lithium working medium starts to flow, the thawing process is completed, and lithium cold space reactor power system has the conditions for normal operation.The auxiliary loop system can take out the decay heat of the reactor core after the reactor is shut down, as a means of residual heat removal of the core.The auxiliary electromagnetic pump and the auxiliary energy conversion unit can provide working medium circulation and power generation functions respectively during the thawing process or after shutdown.

The invention belongs to the technical field of temperature monitoring experiments, and discloses an efficient heating and temperature monitoring experiment system suitable for multiple scenes, and the system comprises a visual experiment box body which is used for carrying out a visual experiment; the power adjusting and monitoring module is provided with a contactless solid-state relay SSR to realize constant-temperature adjustment and power protection; the power adjusting precision is 0.1 W, and the real-time output power is displayed through a touch screen; the exhaust and safety module is provided with a one-way air inlet valve and an automatic pressure relief exhaust valve, and it is ensured that the air pressure in the box is always within the safety range; the top of the box body is provided with a temperature early-warning lamp, and when the outer surface temperature exceeds a set value, the temperature early-warning lamp is automatically turned on and powered off. By simulating different power levels and decay heat conditions, the contact melting influence of heating assemblies in different shapes on a melting experiment piece is analyzed.

A nuclear reactor is configured with a primary coolant loop for transferring heat away from the nuclear reactor core. In a shutdown event, the primary coolant pump may stop pumping primary coolant through the reactor core, resulting in decay heat buildup within the reactor core. An inertial energy coast down system can store kinetic energy while the nuclear reactor is operating and then release the stored kinetic energy to cause the primary coolant to continue to flow through the nuclear reactor core to remove decay heat. The inertial energy coast down system may include an impeller and a flywheel having a mass. During normal reactor operation, the flowing primary coolant spins up the impeller and flywheel, and upon a shutdown event where the primary coolant pump stops pumping, the flywheel and impeller can cause the primary coolant to continue to flow during a coast down of the flywheel and impeller.

The present invention discloses a space-use decay heat Stirling conversion device, which relates to the field of nuclear energy utilization and includes: a double-ended free-piston Stirling motor, a fan-shaped liquid droplet radiation heat dissipation system for space, a Stirling expansion chamber, a plutonium-238 fuel ring, a thermal insulation layer, a shielding layer, a control and power transmission and distribution system, annular fins, a droplet emission stage, a radiation heat dissipation cavity, a droplet collection stage, a tube bundle channel, and an electromagnetic pump. The present invention utilizes the decay heat generated by the plutonium-238 fuel as a heat source, converts the decay heat into electrical energy through a double-ended free-piston Stirling motor, removes waste heat through a fan-shaped liquid droplet radiation heat dissipation system for space, and improves the utilization efficiency of the heat source using a dynamic Stirling converter. The opposing arrangement of the Stirling converters reduces noise and vibration, and the fan-shaped liquid droplet radiation heat dissipation system provides efficient and compact waste heat removal, further ensuring system efficiency and reducing the overall volume, thereby meeting the requirements of space missions for higher power and more efficient energy supply.

The invention provides a small vehicle-mounted reactor passive residual heat removal system and method and a small vehicle-mounted reactor. The removal system comprises at least two groups of parallel heat exchange runners, each group of heat exchange runners comprises a residual discharge pipe, an inlet isolation unit and an outlet isolation unit, the inlet isolation unit and the outlet isolation unit are arranged at the two axial ends of the residual discharge pipe, and the residual discharge pipe is vertically arranged along a reactor cabin and penetrates through the reactor cabin; the inlet isolation unit and the outlet isolation unit are respectively used for controlling the on-off of the residual discharge pipe; when the inlet isolation unit and the outlet isolation unit are in an open state, external air at the bottom of the reactor cabin is conveyed upwards through the residual discharge pipe and flows through the heat exchanger in the reactor cabin, and the heated air is conveyed upwards to the outside of the reactor cabin. According to the passive residual heat removal system for the small vehicle-mounted reactor, the residual heat or decay heat of the reactor core can be reliably guided out in the starting and stopping stages of the vehicle-mounted reactor and under the accident working condition, the reactor core is prevented from being overheated, and the inherent safety and environmental adaptability of the operation of the vehicle-mounted reactor are remarkably improved.

The invention belongs to the technical field of temperature monitoring experiments, and discloses a high-precision constant-temperature control multifunctional heating experiment box and a control system, and the system comprises a visual experiment box body which is used for carrying out a visual experiment; the heating assembly and the temperature monitoring module are provided with two cast copper heating assemblies, and the cast copper heating assemblies are in a parabolic column shape and a cuboid shape and are used for different experiment scenes respectively; each heating assembly is provided with five thermocouple holes, and temperature data are collected in real time and fed back to the control system. The control system supports 48 temperature measuring points to work at the same time based on an Omron high-precision thermocouple and an itinerant detector data acquisition module; a user can set a temperature running program through a touch screen or a remote control interface, and 99 sections of different temperature and time combinations are supported. By simulating different power levels and decay heat conditions, the contact melting influence of heating assemblies in different shapes on a melting experiment piece is analyzed.

The invention relates to a device and a method for analyzing a heat extraction system outside a container for exporting waste heat from a liquid metal reactor, the device comprises a test body and a measuring point bracket, and the measuring point bracket is positioned in the test body; supercooled water or cold air is injected into an outer flow channel of a reaction container or an outer flow channel of a safety container filled with the liquid metal alloy pool, so that the liquid metal alloy pool simulating decay waste heat is subjected to container outer heat removal cooling; a cooling working medium is injected into a reactor pit, and the auxiliary enhancement effect of an immersion liquid metal alloy pool on heat discharge outside a container is researched; a cooling working medium is injected into the upper area of the outer flow channel of the reaction container or the outer flow channel of the safety container, and the enhancement effect of cooling of the upper liquid metal alloy on the natural circulation capacity of the liquid metal alloy pool is judged and enhanced. The device can fully consider and research different influences of test design of multiple factors such as different cooling working media, cooling flow channels, injection strength, immersion heat release, liquid metal alloy pool natural circulation capability promotion and the like on test results.

Systems limit backflow through a nuclear reactor cores from a hot pool to a cold pool. Example systems include coolant risers and / or one-way flow structures before a coolant inlet for the reactor. The coolant riser provides coolant that will flow under gravity alone into the reactor despite hot pool pressure head. Fluidic diodes and / or check-valves prevent nearly all reverse flow. Such structures may be positioned with the flow limiter before the riser in the normal direction of coolant flow into the inlet. Such structures may be positioned anywhere they can functionally provide coast down cooling and prevent reverse flow, including as a pump annulus or pool. For a sodium-cooled fast reactor, the riser and fluidic diode may provide sufficient volume to cool the reactor with forward coolant flow for an entire reactor coast down period, when the reactor is shut down and generating only decay heat.

The present application relates to a kind of spent fuel assembly nuclear material weighing method, system and readable storage medium, the nuclear material weighing method includes: obtaining the initial nominal enrichment of fuel assembly, the type of spent fuel assembly, the burnup value of spent fuel assembly, irradiation duration, decay duration;It is input to the fuel assembly decay heat model established in advance, and obtains the nuclear material mass table exported by fuel assembly decay heat model;Receive the specific nuclide input by user, and from the nuclear material mass table, the mass of specific nuclide generated by unit mass of metal uranium nuclear reaction is inquired;The initial mass of metal uranium of fuel assembly is obtained, and the mass of specific nuclide generated by unit mass of metal uranium nuclear reaction is combined, the mass of uranium-235 in spent fuel assembly, the mass of plutonium-239, the mass of uranium, the mass of plutonium are calculated.Calculation of the mass of nuclear material at any time point from just shutdown time to spent fuel assembly external handling can be realized by the technical scheme, and human error can be reduced.

Fill systems introduce coolant against nuclear reactor exteriors to transfer heat away from the reactor. Coolant is stored inside or outside a containment and allowed to fill the same up to the reactor with a proper amount of coolant. Natural, artificial, fuel pool, underground, and other sources and flows of coolant may be used. Valves or other flow control structures allow an operator or automated plant control systems to determine when and how much the coolant fills around the reactor. Other systems are not blocked by the filling coolant. Thermally-blocking materials or structures can be included as an exterior of reactor to limit risks in temperature difference between the coolant and pressure vessel. Fill systems may be implemented at plant construction or as a retrofit. Other safety systems such as an RVACS may be relieved by the fill systems and differently operated, such as with active systems and security barriers.

A nuclear fuel decay heat utilization system usable for space heating in one embodiment comprises a nuclear generation plant building housing a spent fuel pool containing submerged fuel assemblies which emit decay heat that heats the pool. Plural fluidly isolated but thermally coupled heat removal systems comprising pumped flow loops operate in tandem to absorb thermal energy from the heated pool water, and transfer the thermal energy through the systems in a cascading manner form one to the next to a final external heat sink outside the plant building from which the heat is rejected to the ambient environment. A programmable controller operably regulates the intake and flowrate of water from the heat sink into the heat removal systems and monitors ambient air temperature inside to building. The flowrate is regulated to maintain a preprogrammed building setpoint air temperature by increasing fuel pool water temperature to a maximum permissible limit.

The present application relates to the technical field of spent fuel storage, and discloses a spent fuel dry storage vault, which comprises a top air inlet vertical shaft for introducing natural ventilation, a bottom air chamber in communication with the air inlet vertical shaft, a storage chamber above the bottom air chamber, a plurality of shafts arranged in the storage chamber in a vertical direction, a heat generating body sleeved in the shafts in a spaced manner, an annular gap formed between the shafts and the heat generating body, the heat generating body containing spent fuel, the bottom air chamber in communication with a top air chamber through the annular gap, and an air outlet vertical shaft in communication with the top air chamber. Air for natural ventilation enters the air inlet vertical shaft through an air inlet, then enters the bottom air chamber, and then flows into the annular gap and the top air chamber, and is discharged from an air outlet through the air outlet vertical shaft, so that the flow path of the internal air is optimized, the storage chamber enhances the ventilation efficiency and heat exchange performance in a combined manner of heat conduction, convection and radiation heat exchange, the decay heat of the spent fuel in the heat generating body is discharged, and the safety and stability of the spent fuel dry storage vault are ensured.

The invention belongs to the technical field of reactor design and operation, and particularly relates to a nuclear power plant low-power reactor core power measurement method, which comprises the following steps: step 1, measuring the thermal power of a heat source outside a reactor core during the first zero power debugging period; 2, measuring the power of decay heat in a zero power period after power operation; 3, calibrating the time from the current input parameter of the decay heat semi-empirical formula to the shutdown time of the reactor by adopting the decay heat obtained by measurement; and step 4, measuring the reactor core power under low power after operation. According to the method disclosed by the invention, mutual power correction is realized by utilizing heat balance or quasi-heat balance of the primary side and the secondary side of the steam generator, and accurate reactor core power under low power is obtained.

The application provides a personnel intervenable time calculation method, system and medium under a spent pool accident scenario, and relates to the field of nuclear power. The personnel intervenable time calculation method comprises the following steps: obtaining refueling data of a nuclear power plant, and calculating total decay heat power of the spent pool before an accident according to the refueling data; obtaining spent pool data of the spent fuel pool and the number of newly added spent fuel assemblies corresponding to each accident time within the total duration of the accident; calculating a boiling time through time step iteration according to the number of newly added spent fuel assemblies, the spent pool data and the total decay heat power of the spent pool before the accident; and calculating a water supply starting time, a water supply ending time and a total required water supply amount through time step iteration according to the boiling time, the number of newly added spent fuel assemblies, the spent pool data and the total decay heat power of the spent pool before the accident.

The present invention discloses a system and method for the decay heat removal of a sodium-cooled fast reactor. The water outlet of the steam-water separator is connected to the inlet of the heat-releasing side of the regenerative heat exchanger. The outlet of the heat-releasing side of the regenerative heat exchanger is connected to the inlet of the drain cooler. The outlet of the drain cooler is divided into two paths. One path is connected to the inlet of the heat-absorbing side of the regenerative heat exchanger through the condensate polishing system, and the other path is connected to one end of the temperature regulating valve. The other end of the temperature regulating valve is connected to the outlet of the heat-absorbing side of the regenerative heat exchanger through a pipe and then connected to the inlet of the deaerator. The gas outlet of the steam-water separator is connected to the inlet of the auxiliary steam header. The outlet of the auxiliary steam header is connected to the inlet of the deaerator. The outlet of the deaerator is connected to the inlet of the steam-water separator through the steam generator. The system and method can effectively ensure the qualified water quality of the feed water of the steam generator of the sodium-cooled fast reactor, reduce the fuel consumption, and avoid wasting a large amount of demineralized water.

The invention relates to a three-flow structure heat exchanger auxiliary molten salt reactor cooling system which is characterized in that a nuclear reactor core is communicated with the primary side of a primary loop heat exchanger through a primary loop pipeline, and the secondary side B of the primary loop heat exchanger is communicated with the primary side of a secondary loop heat exchanger through a secondary loop pipeline; a secondary side A of the primary loop heat exchanger is communicated with a primary side of the air cooler through a residual discharge loop pipeline, primary loop fluid in the primary loop pipeline provides a first natural circulation loop in a shutdown state, and residual discharge loop fluid in the residual discharge loop pipeline provides a second natural circulation loop. A certain height difference is arranged between the center of the nuclear reactor core and the center of the primary loop heat exchanger, and a certain height difference is arranged between the center of the primary loop heat exchanger and the center of the air cooler. According to the molten salt reactor cooling system provided by the invention, a heat transfer working medium is non-water, and during planned shutdown and most accident shutdown, a reactor does not need to discharge fuel salt out of a reactor core, and decay heat is discharged out of the reactor core completely depending on passive action.

The present invention relates to a liquid metal cooled nuclear reactor comprising a fully passive decay heat removal (DHR) system with a modular cold source. The invention mainly consists in manufacturing a nuclear reactor comprising a system which ensures at the same time: - decay heat removal in a fully passive way from the moment of the beginning of the accident; - heat removal through the primary vessel; - reduction of the risk of chemical interaction between sodium (or NaK) and the material used as final cold source; - presence of a final cold source which provides the same function as the sodium / air or NaK / air exchanger used in the state of the art but with different properties (storage with PCM).

The application discloses a high-level radioactive waste disposal pit multi-field coupling condition full cycle simulation system and a test method, and relates to the field of rock mechanics tests.The system comprises a true triaxial dynamic and static loading device, which can bear a fractured rock sample to be tested, and can move the fractured rock sample to a set position, and can apply dynamic and static loads in multiple directions to the fractured rock sample; a vertical excavation and tunneling device can realize simulation of an excavation process of dynamic disturbance drilling of the center of the fractured rock sample under a deep geostress environment; a disposal pit reaction source device is placed in the disposal pit to simulate a nuclide leakage process under conditions of decay heat release, corrosion and gas production of a high-level radioactive waste tank; and a fractured seepage control device can apply a water head difference to left and right end surfaces of the fractured rock sample in a horizontal direction through a rock sample pad to drive seepage solution to flow along original fractures in the fractured rock sample.The application can truly reproduce disposal pit full cycle disturbance effects, integrate multiple barrier coupling boundary conditions, and consider seepage influence on storage tank surface corrosion differences.