24 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.

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 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 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.

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 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 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.

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.

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 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.

The invention discloses a melt and concrete multiphase heat transfer and ablation diffusion calculation method and system, and belongs to the technical field of pressurized water reactor serious accident analysis. The method comprises the following steps: calculating melt viscosity and diffusion radius through a fragment melt diffusion model, and determining spatial distribution; molten pool temperature and boundary heat flux density are calculated based on spatial distribution and debris decay heat; calculating heat transfer to the coolant layer through a water intrusion model and a melt eruption model; calculating an interface heat transfer coefficient by adopting a gas film or slag film model according to the decomposition state of the concrete; calculating the concrete ablation rate and distance based on the interface heat transfer coefficient; and finally, comparing the ablation distance with the safety thickness to realize the safety quantitative evaluation of the reactor pit. According to the invention, the whole interaction process of the melt and the concrete can be accurately simulated, and reliable technical support is provided for serious accident analysis.

The present invention relates to a molten salt reactor system comprising a reactor vessel comprising a reactor core in an upper compartment, and one or more drain tanks situated in a lower compartment below the upper compartment, and a molten fuel salt pump, and a heat exchanger, wherein a molten fuel salt flow direction is provided in a fuel salt loop, said fuel salt loop comprising the reactor core, the one or more drain tanks, the molten fuel salt pump, and the heat exchanger, the direction of the molten fuel salt flow is from the heat exchanger towards the reactor core, wherein (a) the one or more drain tanks are located between the reactor core and the molten fuel salt pump in the molten fuel salt flow direction and the molten fuel salt pump is located between the one or more drain tanks and the heat exchanger in the molten fuel salt flow direction and the heat exchanger is located between the molten fuel salt pump and the reactor core in the molten fuel salt flow direction and the reactor core is located between the heat exchanger and the one or more drain tanks in the molten fuel salt flow direction, or (b) the molten fuel salt pump is located between the reactor core and the one or more drain tanks in the molten fuel salt flow direction and the one or more drain tanks are located between the molten fuel salt pump and the heat exchanger in the molten fuel salt flow direction and the heat exchanger is located between the one or more drain tanks and the reactor core in the molten fuel salt flow direction and the reactor core is located between the heat exchanger and the molten fuel salt pump in the molten fuel salt flow direction, wherein the lower compartment is provided with a passive decay heat removal system.

The invention provides a method and a system for calculating personnel intervention time in a spent pool accident scene, and a medium, and relates to the field of nuclear power. The personnel intervention time calculation method comprises the following steps: acquiring refueling data of a nuclear power plant, and calculating the total decay heat power of a spent pool before an accident according to the refueling data; acquiring spent fuel pool data of the spent fuel pool and the number of newly added spent fuel assemblies corresponding to each accident moment in the total accident occurrence duration; according to the number of the newly added spent fuel assemblies, the spent pool data and the spent pool total decay heat power before the accident, the boiling moment is obtained through time step iterative calculation; according to the boiling moment, the number of the newly added spent fuel assemblies, the spent pool data and the spent pool total decay heat power before the accident, the water supplementing starting moment, the water supplementing ending moment and the total required water supplementing amount are obtained through time step iterative calculation.

The invention relates to the technical field of advanced nuclear energy engineering, and discloses a radioactive isotope nuclear power source for seabed fixation application, which comprises a shell located on the seabed, a sealed heat transfer cavity is arranged in the shell, and a gap is arranged between the shell and the sealed heat transfer cavity; the radioactive isotope heat source assembly is located in the middle in the sealed heat transfer cavity, the sealed heat transfer cavity is filled with liquid metal, the radioactive isotope heat source assembly generates heat through isotope decay, and the generated heat is transmitted to the sealed heat transfer cavity through the liquid metal; the temperature difference thermoelectric conversion device is located in the gap, the two ends of the temperature difference thermoelectric conversion device are arranged on the outer wall of the sealed heat transfer cavity and the inner wall of the shell correspondingly, heat transferred by the liquid metal is subjected to temperature difference power generation through the temperature difference thermoelectric conversion device, and waste heat generated after power generation is transmitted into seawater. The heat source heat is guided out through cooperation of the heat pipe and the liquid metal, decay heat is converted into electric energy through the temperature difference thermoelectric conversion technology, finally, seawater serves as a heat trap, and safe and stable long-term power supply can be achieved.

The invention relates to the technical field of nuclear reactors, in particular to a passive residual heat removal system and a lead-based fast neutron high-flux research reactor. The residual heat removal system comprises a reflecting layer structural part which is of a hollow structure with through holes formed in the two ends, the reflecting layer structural part is vertically arranged on the outer side of the fuel assembly in a surrounding mode in the working state, and an inner cavity and a cooling runner of the fuel assembly are arranged in parallel; the lower end of the closed air cooling container is provided with an air inlet channel, the upper end of the closed air cooling container is provided with an exhaust channel, and the air inlet channel can be opened and closed and is hermetically sleeved outside the reaction area of the reactor container in a working state; the discharge chimney is vertically arranged, and the lower end of the discharge chimney is connected with the exhaust channel, so that the residual fission heat and decay heat of the reactor core can be efficiently exported under the working condition of long-term normal shutdown or accident shutdown of the reactor, and the damage of in-reactor fuel elements and in-reactor components caused by overheating of the reactor core is avoided; and the inherent safety of the lead-based fast neutron high-flux research reactor is obviously improved. The research stack is loaded with the waste heat removal system.

The embodiments of this application relate to the field of nuclear reactor monitoring, and particularly to a method for determining the power of heat generated by the decay of spent fuel assemblies in a sodium-cooled fast reactor. The method includes the following steps: S10: placing the spent fuel assembly of the sodium-cooled fast reactor whose decay heat is to be determined in a measurement space; S20: placing a coolant in the measurement space, the coolant being configured to conduct away a portion of the decay heat generated by the spent fuel assembly; S30: determining the average temperature of the walls of the measurement space, determining the power of heat dissipation from the measurement space and the power of heat conducted away by the coolant; S40: determining the power of heat generated by the decay of the spent fuel assembly based on the power of heat dissipation from the measurement space and the power of heat conducted away by the coolant. This makes the decay heat power of the spent fuel assembly more accurate, reduces the uncertainty level, thereby reducing the design conservatism margin of the decay heat extraction system, and avoids safety problems such as coolant overcooling, improving reactor safety and economy.

This invention relates to light water nuclear reactors, particularly pressurized water reactors (PWRs) or boiling water reactors, including an integrated autonomous passive decay heat removal system. The invention discloses a light water nuclear reactor (LWR), particularly a PWR or BWR, comprising: - a reactor core; - a system for removing at least some decay heat from the reactor core, comprising: a first water reservoir or pool disposed above the reactor core; a heat exchange device immersed in the pool; an organic Rankine cycle; a fluid loop; a second water reservoir different from the pool; and a second pump. The invention lies in installing an organic Rankine cycle and a supplementary water reservoir different from the pool, wherein the energy stored in the pool is the heat source for the organic Rankine cycle evaporator, and the supplementary water reservoir directly supplies the organic Rankine cycle condenser via a dedicated pump to constitute the cold source for the organic Rankine cycle condenser.