217 results about "Nuclear facilities" patented technology

The invention provides a nuclear radiation diffusion analysis method based on a Gaussian mixture atmospheric diffusion model. The method comprises the steps that nuclear facility source item parameters are acquired; integrating meteorological, geography, remote sensing and radiation monitoring data by using a heterogeneous data fusion engine, extracting atmospheric turbulence, terrain and building interference characteristics and source item parameters, and forming a dynamic input parameter set; constructing an unsteady state diffusion model containing a multi-modal distribution function, processing features, obtaining a mixed distribution model, and correcting the mixed distribution model; predicting spatial and temporal distribution of radiation pollutants, and generating multi-scale diffusion field simulation; and optimizing a diffusion path, and generating a three-dimensional risk field. According to the method, multi-source data are integrated, an unsteady state diffusion model is constructed, Gaussian component weights are dynamically adjusted, the model is corrected by using real-time data, and a geographic information system is coupled to carry out spatial constraint optimization. And through multi-scale feature coupling, dynamic parameter adjustment and geographic space constraint correction, the depicting capability of the atmospheric diffusion model is remarkably improved, and the nuclear radiation diffusion prediction precision is greatly improved.

The invention relates to a shot blasting particle internal circulation regeneration method for decontaminating radioactive contaminated metal, which is applied to a shot blasting particle internal circulation regeneration system for decontaminating radioactive contaminated metal. The invention discloses a shot blasting particle internal circulation regeneration system for decontaminating radioactive contaminated metal. The shot blasting particle internal circulation regeneration system comprises a smelting furnace, a melting and pumping machine and a shot blasting machine which are sequentially used according to a closed-loop sequence, the method comprises the following steps: raw material pretreatment; smelting and decontaminating the raw materials; preparing pills by a melt extraction machine; shot-blasting and decontaminating the raw materials; and after decontamination, smelting and pelleting. The method has the advantages that radioactive contamination metal is converted into self-consumption shot blasting particles, cyclic regeneration is achieved, radioactive waste increment is remarkably reduced, near-zero emission of nuclear facility decommissioning waste is promoted, and the requirement for minimization of IAEA radioactive waste is met.

The invention relates to a nuclear facility radiation field model construction method and a dose evaluation system, and the system is configured to establish a three-dimensional space coordinate system, construct a three-dimensional radiation field model based on radiation field data and radiation source data, and correct the radiation field model at the current moment according to the radiation source data, introducing an attenuation coefficient of the three-dimensional nuclear facility on the original three-dimensional scene model to correct the three-dimensional scene model; correcting the radiation field model based on the corrected three-dimensional scene model; calling the three-dimensional scene model and the radiation field model according to an operation path selected by a user; and simulating the space range of each organ of the human body sweeping along with the movement of the human body in the working path in the three-dimensional scene model, obtaining the time of each organ staying on the path of the unit space, and calculating the real-time dose received by the human organ according to the radiation dose rate in the unit space. According to the invention, the change of the radiation field caused by the disassembly and assembly of the nuclear facility can be corrected, and the optimization of radiation protection is realized.

The invention relates to the technical field of pipeline decontamination, in particular to an online pipeline ultrasonic decontamination device and method suitable for various pipe diameters, and the online pipeline ultrasonic decontamination device comprises a numerical control generator, a transducer, an amplitude-change pole, an adaptive clamp, an online monitoring device and other key components. And through the designed adaptive clamp, the device can act on pipelines with various pipe diameters and can move along the pipelines, and the decontamination requirements in different scenes are met. The online ultrasonic decontamination device is suitable for maintenance and cleaning work of pipelines with different pipe diameters in nuclear power facilities, and has important significance in improving the safety of the nuclear facilities and guaranteeing radiation protection for operators. The problem that an ultrasonic device is poor in applicability is effectively solved, meanwhile, the flexibility of the device is guaranteed, effective online decontamination of pipelines with various pipe diameters can be achieved through a single set of device, the decontamination cost of radioactive contamination of nuclear facility pipelines is effectively reduced, online efficient decontamination of radioactive contamination deposited in the pipelines is achieved, and the service life of the pipelines is prolonged. The safe and stable operation of the reactor is ensured, and the radiation risk of operators is reduced.

The invention provides a multi-nuclide automatic separation and detection system and method, and particularly relates to the technical field of radioactive substance detection. The detection system comprises a cascade separation and regeneration unit, a dynamic regulation and control unit, an online detection unit, an eluent conveying unit and a control system, and the eluent conveying unit is matched with the cascade separation and regeneration unit to separate and purify radionuclide and regenerate resin. The dynamic regulation and control unit is used for controlling a flow path between the eluent conveying unit and the cascade separation regeneration unit and feeding back changes of pressure, temperature and pH value in the flow path in real time, and the online detection unit is used for detecting concentration of nuclides in separation liquid and feeding back monitoring data to the control system. And the control system controls the operation of each functional unit and automatically plans a separation path according to a measurement requirement to form a separation-detection-optimization closed-loop process. The device is suitable for the fields of nuclear facility liquid effluent detection, environmental water radioactivity detection and the like, and has high efficiency, environmental protection property and intelligence.

The invention discloses an edge computing gateway system based on radiation monitoring equipment, and belongs to the field of radiation monitoring, the edge computing gateway system comprises the radiation monitoring equipment, an edge computing gateway and a cloud Internet of Things platform which are connected in sequence, and the edge computing gateway is connected with the cloud Internet of Things platform through an MQTT protocol; the edge computing gateway mainly comprises a dual-architecture processor, a protocol library plug-in, an edge computing engine, an offline data storage module and a device configuration module. The system is dynamically compatible with a multi-source heterogeneous protocol, supports Internet of Things platform standard communication, simplifies configuration management, realizes local edge calculation, provides abnormal data discrimination and other functions, and solves the problems of poor protocol compatibility, insufficient data reliability, complex configuration management, lack of intellectualization, weak support of mobile equipment and the like in the prior art. The system is especially suitable for monitoring and management of radiation environments such as nuclear facilities, hospitals and industrial flaw detection.

To dismantle a nuclear facility affected by a severe accident in an underwater environment by submerging the facility in water so that the water shields radiation from the nuclear facility.SOLUTION: The method for dismantling a nuclear facility includes: stopping water leakage at a leakage portion by using a particle dispersion liquid containing fine particles having a specific gravity greater than that of water, when dismantling a nuclear facility in which the water leakage portion has occurred due to a severe accident; and submerging at least a part of the nuclear facility. This makes it possible to submerge a reactor well and to perform dismantling work from above a reactor building.SELECTED DRAWING: Figure 1

The utility model discloses a cleaning isolation cabin for installing a modular integrated reactor. The cleaning isolation cabin comprises a main cabin body, a movable ceiling, a bearing base, air purification equipment, an electric control box body, an air shower, an illuminating lamp and a monitoring camera, the main cabin body is mounted on the bearing base, the upper end of the main cabin body is opened to form a hoisting opening, and the movable ceiling is correspondingly and horizontally arranged on the hoisting opening; an isolation cabin access port communicated with the outside is formed in the circumferential direction of the main cabin body, the air shower is correspondingly arranged on the outer side of the isolation cabin access port, and an outlet of the air shower is communicated with the isolation cabin access port; the air purification equipment is communicated with the interior of the main cabin body, the electric control box body is arranged on the inner peripheral wall of the main cabin body, and the illuminating lamp and the monitoring camera are arranged at the top of the inner peripheral wall of the main cabin body. The clean isolation cabin has the advantages that repeated investment can be reduced, large-range isolation space is avoided, manpower and material resource investment can be reduced, the construction period of nuclear facility engineering is shortened, and economic efficiency is improved.

The invention relates to the technical field of compressed air system intelligent control, and discloses an intelligent control and air supply distribution method and system for a nuclear facility compressed air system pipe network. Carrying out numerical simulation on the distributed arrangement and operation conditions of different air compression stations and the flow and pressure characteristics of the compressed air pipe network under the intermittent load of different process equipment; on the basis of data generated by numerical simulation, the pipe network tail end pressure of the compressed air pipe network system is adjusted through an intelligent fuzzy self-tuning PID control method, and the operation state of an air compressor unit of the compressed air pipe network system is optimized on the basis of a flow matching strategy; pipe network operation parameters of the compressed air pipe network system are collected in real time, and the control strategy is fed back and adjusted. The whole compressed air system can operate efficiently and stably, the requirements of all links of nuclear facilities for compressed air are guaranteed, meanwhile, the operation intelligence and reliability of the system are improved, and adverse effects caused by unstable pressure and the like are reduced.

The invention discloses a parameter optimization method and system based on an agent model and multi-algorithm gradient collaboration, and relates to the technical field of nuclear engineering and nuclear security, and the method comprises the steps: firstly determining a nuclear facility parameter combination, carrying out the high-fidelity calculation through a Monte Carlo program to obtain a Keff value, and constructing a training set; training a micro-proxy model capable of predicting a Keff value and gradient information through a feedforward neural network in combination with a back propagation algorithm, taking the micro-proxy model as a target function, taking the minimum absolute error of a Keff predicted value and a target value as an optimization target, and generating candidate parameter solutions through global exploration of a hybrid optimization algorithm; and an optimal parameter combination candidate solution is obtained through gradient-assisted local optimization, and finally, through Monte Carlo program verification, if the error reaches the standard, the optimal parameter combination candidate solution is used as a final parameter combination solution. The problem that a parameter optimization method in the prior art cannot give consideration to calculation efficiency and precision is solved.

The invention belongs to the technical field of robot teleoperation and haptic interaction, and provides a multi-mode force feedback control method and system for nuclear facility robot operation, and the method comprises the steps: collecting the state information of a master arm and a slave arm in real time, and dynamically scheduling a haptic calculation strategy according to a task state; a fast estimation strategy is enabled in the fast transit mode. Further, real environment interaction force is obtained based on dynamic model compensation and net interaction force extraction; combining the dynamically selected multi-mode force feedback parameters, and synthesizing and outputting a touch instruction to the main arm. Meanwhile, admittance correction and feed-forward compensation are carried out on a slave arm motion instruction through net interaction force and dynamic internal force, synchronous output is carried out after safety optimization, and the slave arm is controlled to move. The problems of poor task adaptability, force sense distortion and insufficient safety in nuclear facility teleoperation are solved, and intelligent, high-fidelity and safe robot teleoperation is achieved.

The embodiment of the invention provides a nuclear facility decommissioning project three-dimensional waste cutting and boxing collaborative simulation method and system. The method comprises the steps of obtaining three-dimensional model data, waste storage container data and cutting constraint data of a target nuclear facility; performing format conversion on the three-dimensional model data and the waste storage container data to generate target three-dimensional data and target container data; based on a preset optimization engine, the target three-dimensional data, the target container data and the cutting constraint data, cutting and boxing co-simulation is carried out, and a simulation scheme is generated; optimizing the simulation scheme based on preset optimization parameters, and generating a Pareto frontier solution set; in response to selection of a user in the Pareto frontier solution set, determining a final selection result; and generating a cutting and boxing simulation project report based on the final selection result. In this way, the waste can be flexibly and efficiently treated.

The invention discloses an electrochemical denucleation decommissioning facility oxide layer method based on a synergistic acidic electrolyte, and relates to the technical field of cleaning. The method comprises the following steps: by taking a polluted metal component as an anode and an inert electrode as a cathode, immersing the polluted metal component into a synergistic acidic electrolyte of a low-concentration acidic medium, and stripping an oxide layer on the surface of the polluted metal component until a clear grain boundary is exposed at the temperature of lower than 40 DEG C in a high-voltage direct-current electrolysis mode. By developing a synergistic enhanced acidic electrolyte system, high decontamination rate is realized in a low-concentration sulfuric acid medium, radionuclide stripping is accurately controlled to avoid matrix pitting corrosion damage, and high-valence nuclides in a compact oxide layer are deeply removed, so that the decontamination efficiency is improved. Finally, an efficient, low-damage and stable-process radioactive contamination metal treatment scheme is provided for nuclear facility decommissioning engineering. The method is suitable for removing radioactive contamination oxidation layers on the surfaces of the inner walls of stainless steel pipelines and reactor containers in nuclear facility decommissioning projects, and radionuclides are enriched in the oxidation layers.

The invention discloses a preparation method of a starch-based gel color developing material for visual detection of airborne iodine nuclides, which comprises the following steps: adding starch, a polymer matrix material and a plasticizer into deionized water for soaking and uniformly stirring to obtain a starch-based gel matrix solution; placing the starch-based gel matrix solution at a certain temperature, and stirring to obtain gel paste; and injecting the gel paste liquid into a mold, and carrying out freeze-thaw cycle to obtain the starch-based gel color developing material for airborne iodine nuclide visual detection. The starch-based gel color developing material has the characteristics of quick response, stable color development and lasting color and luster for airborne iodine nuclide at normal temperature, has excellent mechanical strength, machinability and recycling performance, can be flexibly processed into forms such as a film and a test piece, or can be integrated in various detection devices, and has a wide application prospect. The method is suitable for places needing airborne iodine nuclide detection, such as nuclear medicine laboratories, nuclear facilities / nuclear equipment and the like, and meets the requirements of real-time visual detection and early warning.

The invention discloses an anti-leakage intelligent monitoring double-layer casing pipe, assembling equipment and an assembling method thereof, and relates to the technical field of nuclear facility or chemical equipment matching, the anti-leakage intelligent monitoring double-layer casing pipe comprises an inner pipe and an outer pipe, the outer pipe is arranged on the outer side of the inner pipe in a sleeving mode, and multiple sets of supporting rings are arranged between the inner pipe and the outer pipe at equal intervals; the multiple sets of supporting rings are connected with the inner pipe. A monitoring device is arranged on one side of the outer pipe and used for monitoring leakage of the inner pipe. The assembling equipment comprises a supporting and guiding assembly, a welding assembly, an inner pipe milling cutter, an inner pipe ironing plate, an outer pipe milling cutter and an outer pipe ironing plate. The assembling method comprises the steps of preparation, inner pipe butt joint and appearance butt joint. A single safety barrier is upgraded into a double-independent safety barrier by arranging a double-layer sleeve structure formed by combining the inner pipe and the outer pipe.

The invention relates to the technical field of nuclear facility pipeline heating. The invention provides a pipeline heating device and an in-reactor test facility. The pipeline heating device comprises a magnetic conduction layer, a heat conduction layer, a heat preservation layer and an electromagnetic induction coil, the magnetic conduction layer is made of a magnetic alloy material, and the magnetic conduction layer is enclosed and internally forms a heating area used for containing a pipeline; the heat conduction layer is located on the side, facing the heating area, of the magnetic conduction layer and used for transmitting heat of the magnetic conduction layer to the pipeline. The heat preservation layer wraps the side, away from the heating area, of the magnetic conduction layer. The electromagnetic induction coil is wound around the side, away from the heating area, of the heat preservation layer. The magnetic conductive layer made of the magnetic alloy material serves as a main absorption and conversion carrier of magnetic field energy, the eddy current effect can be remarkably enhanced, the problem that the heating efficiency is insufficient due to low magnetic conductivity of the austenitic stainless steel pipeline is solved, and the heating efficiency is greatly improved. The arrangement of the magnetic conductive layer can fundamentally avoid material microscopic damage caused by direct action of a magnetic field, and the safety is higher.

The invention provides a combined automatic treatment system and method for tritium and carbon-14 in gaseous effluent, and particularly relates to the technical field of nuclear facility gaseous effluent monitoring. The treatment system comprises a gas inlet unit, a catalytic oxidation unit, a tritium trapping unit, a carbon trapping unit and a circuit control unit, a gas inlet of the catalytic oxidation unit is communicated with the gas inlet unit, and the catalytic oxidation unit is used for converting tritium gas and organic tritium in gaseous effluent into tritiated water vapor and converting organic carbon into carbon dioxide; the tritium trapping unit is communicated with a gas outlet of the catalytic oxidation unit, and can be used for condensing tritiated water vapor into liquid water and automatically sampling and detecting; the carbon capture unit comprises a neutralizer and a carbon dioxide sensor, a NaOH solution for absorbing carbon dioxide is arranged in the neutralizer, and the carbon dioxide sensor detects the absorption efficiency of carbon dioxide in real time; and the circuit control unit controls the operation of each unit. According to the system, catalytic oxidation, tritium condensation and carbon absorption are integrated, and synchronous treatment and monitoring of tritium and carbon-14 can be achieved.

A nuclear installation comprising at least one modular nuclear reactor (SMR) and a reactor vessel well delimiting a water basin in which the SMR reactor unit is immersed. The invention relates to a nuclear installation comprising: at least one SMR reactor unit (4) for heat generation, delimited by a reactor vessel (40, 41, 42); a reactor vessel well (100) delimiting a water-filled basin (B) forming part of a secondary circuit, in which the reactor vessel is immersed; at least one heat exchanger (49) between the primary circuit of the SMR reactor and the secondary circuit, arranged outside the reactor vessel and in the water basin. Figure for the abstract: Fig. 6

The invention relates to the technical field of neutron measurement, and particularly discloses a multi-activation foil and method for measuring a neutron energy spectrum. The multi-activation foil comprises a shell and an activation sheet placed in the shell, and the activation sheet sequentially comprises Ni, Sc, Au, W, Cu, Pt, In, Zn and Fe from top to bottom. The activated foil can realize high-resolution measurement of a neutron spectrum in a very small space, has the characteristics of size miniaturization, wide energy response range, wide neutron fluence rate application range, controllable manufacturing tolerance, optimized material ratio and the like, and is suitable for neutron spectrum and neutron dose measurement of a small-scale region of a nuclear facility and neutron dose measurement of a small-scale region of a nuclear facility. The method can also be popularized and applied to fine dose field distribution measurement of a large-scale area of a nuclear facility, and provides accurate data support for safe operation, shielding verification and radiation protection of the nuclear facility.

The invention discloses a liquid-phase deuterium-doped anti-radiation optical fiber, a preparation method and application, and belongs to the technical field of manufacturing of anti-radiation optical fibers. According to the method, deuterium is stably doped into a fiber core in the form of OD groups in the preparation stage of the optical fiber preform through an improved chemical vapor deposition combined solution doping technology, so that the problems that gas is easy to leak and the radiation resistance is attenuated along with time in a traditional gas loading technology are solved. The preparation process comprises the steps of pretreatment of the deposition tube, preparation of the barrier layer and the porous loose layer, solution soaking OD group doping, sintering collapse and wire drawing treatment. An OD group can be detected in the prepared optical fiber, the radiation resistance of the optical fiber is remarkably improved, after the optical fiber is irradiated by gamma rays with the total dose of 200Gy (Si), the radiation-induced attenuation is remarkably lower than that of a common optical fiber without the OD group, and a material basis is provided for long-term reliable application of the optical fiber in radiation environments such as space exploration and nuclear facilities.

The invention belongs to the technical field of nuclear facility decommissioning and three-waste treatment, and particularly relates to an automatic laser decontamination system and method for polluted metal waste of a nuclear power plant. The decontamination / cutting workbench is transferred to all stations and an operation table through a transfer trolley; the operation table is used for controlling the transfer of the transfer trolley, controlling the operation of the workpiece turnover device, measuring the size of a to-be-decontaminated item and measuring the surface contamination level and radiation level of the to-be-decontaminated item before and after decontaminating; and the controller is used for recording a measurement result, presetting cutting parameters and decontamination process parameters of a to-be-decontaminated item, and controlling the laser cutting device, the laser cleaning device and the laser stripping device to carry out laser cutting and automatic laser decontamination operation. According to the invention, the labor intensity of personnel can be effectively reduced, the exposure dose of the personnel is reduced, the decontamination working efficiency is improved, waste degradation or clean control of surface polluted metal waste is realized, and waste minimization is further realized.

The objective is to provide a method for dissolving solids that is excellent at dissolving solids generated in nuclear facilities. [Solution] A method for dissolving a solid generated at a nuclear facility, comprising dissolving the solid in a solution containing at least one of carbonate ions and bicarbonate ions, and at least one of potassium ions and cesium ions.

The invention relates to anti-radiation ultra-high performance recycled concrete and a preparation method thereof, and belongs to the technical field of building materials and solid waste resource utilization. According to the concrete, a multi-source silicon-aluminum precursor mainly comprising ground blast furnace slag and an alkali activating solution are subjected to synergistic reaction, Portland cement is not contained, and a compact structure and excellent mechanical, durability and radiation protection performance are formed by combining ferronickel slag reclaimed sand with refined particle size, compounded shielding reinforcing filler (lead glass and barite) and high-performance copper-plated steel fibers. According to the method, high-valued recycling of the industrial solid waste is achieved, and the method has excellent environment friendliness, engineering adaptability and long-term stability and is suitable for high-protection-demand scenes such as nuclear facilities and radioactive places.

The invention relates to the technical field of neutron measurement, and particularly discloses a thermoluminescence field dosimeter for measuring neutron energy spectrum.The thermoluminescence field dosimeter comprises a thermoluminescence field dosimeter body, and the thermoluminescence field dosimeter body comprises a front-end assembly, a rear-end assembly and a center assembly located between the front-end assembly and the rear-end assembly; the center assembly and the rear end assembly are detachably connected with the front end assembly. Wherein the distance between the front surface of the central assembly and the front surface of the front end assembly is 1mm, the distance between the front surface of the rear end assembly and the front surface of the front end assembly is 7mm, and the front surface of the central assembly and the front surface of the rear end assembly are respectively provided with a 6LiF thermoluminescence dosage piece and a 7LiF thermoluminescence dosage piece. The thermoluminescence field dosimeter is wide in neutron energy response range, covers 0.0253 eV-20 MeV, can be suitable for neutron energy spectrum measurement of nuclear facilities such as fission reactors, fusion reactors and medical accelerators, and is particularly suitable for measurement in measurement occasions with narrow spaces and large neutron field gradients.

The application provides an automatic sampling and measuring device for radioactive aerosol and gaseous tritium in radioactive gaseous effluent. The measuring device comprises a radioactive aerosol measuring module and a gaseous tritium measuring module integrated in the measuring device, and the radioactive aerosol measuring module and the gaseous tritium measuring module are communicated with a gaseous effluent pipeline through sampling pipelines. The measuring device of the application can greatly save labor, realize unattended and long-time online monitoring, flexibly adjust the sampling and monitoring frequency, and improve the monitoring efficiency. In addition, in the case of a nuclear accident, the emergency monitoring mode can be quickly started to quickly respond to the accident monitoring requirements, so as to realize continuous, online and automatic monitoring of the concentration of radioactive aerosol and gaseous tritium in the effluent of a nuclear facility.

Embodiments of the present application provide a method and system for three-dimensional waste cutting and packing simulation for nuclear facility decommissioning engineering. The method comprises obtaining three-dimensional model data, waste storage container data and cutting constraint data of a target nuclear facility; performing format conversion on the three-dimensional model data and the waste storage container data to generate target three-dimensional data and target container data; performing cutting and packing simulation based on a preset optimization engine, the target three-dimensional data, the target container data and the cutting constraint data to generate a simulation scheme; optimizing the simulation scheme based on a preset optimization parameter to generate a Pareto frontier solution set; determining a final selection result in response to user selection in the Pareto frontier solution set; and generating a cutting and packing simulation engineering report based on the final selection result. In this way, waste can be efficiently processed flexibly.

Safety container cooling system (10) comprising a sealed safety container (12) of a nuclear facility, - with a heat exchanger for heat transfer between a medium to be cooled and a cooling medium, - with first means to extract the medium to be cooled from the safety vessel (12) in a first cooling circuit and supply it to the heat exchanger and, after it has flowed through the heat exchanger, return it to the safety vessel (12), wherein the first means comprise a first pumping device (38) to pump the medium, now cooled by the heat exchanger, back into the safety vessel (12), - with second means to supply the recoolant from a coolant reservoir to the heat exchanger and to return it to the coolant reservoir after it has passed through the heat exchanger, wherein the second means comprise a second pumping device located outside the safety vessel (12) to set the recoolant in motion, wherein the second means comprise a turbine (40) which is driven by the flow of the cooling medium, wherein the first pump device (38) is coupled to the turbine (40) in such a way that it is driven by it, characterized by the fact that the first pumping device (38) and the turbine (40) are arranged outside the safety container (12), that the heat exchanger is a condenser (30) arranged outside the safety container (12), that the first means have a withdrawal device (24) in the safety container (12) for withdrawing the medium to be cooled from the safety container (12), which has a steam component and an air component, and at least the steam component is directed to the condenser (30), that the steam component is condensed to water by the condenser (30), that the water is pumped back into the safety container (12) as a cooled medium by the first pumping device (38).

The present application relates to the technical field of nuclear power plant ventilation purification, and discloses an equivalent test bench for evaluating the performance of iodine removal and a verification method, the verification method comprising: obtaining three carbon samples with different iodine removal efficiencies; based on the three carbon samples, performing a first test under the French standard system, a second test under the American standard system, a third test under different humidity conditions, and a fourth test under accident conditions on the same test bench, and recording the corresponding test results; comprehensively analyzing the test results corresponding to the four tests to obtain the test equivalent relationship between the French standard system and the American standard system, the numerical relationship between humidity and iodine removal efficiency, and the change rule of iodine removal efficiency after accident radiation. The present application accurately establishes the quantitative relationship between the French standard and the American standard, humidity and iodine removal efficiency, and accident radiation and performance decay based on a unified test bench; provides a unified scientific basis for the performance evaluation of iodine adsorbers in nuclear power plants, and improves the safety and reliability of gaseous radioactive contamination protection of nuclear facilities.

A hybrid power generation system which includes a nuclear facility comprising a nuclear reactor and an exclusion zone. A thermal energy storage, a nuclear steam supply system located, and a solar energy collection system are all located within the exclusion zone. The thermal energy storage vessel contains a thermal mass composition operable to store thermal energy. The nuclear steam supply system has a nuclear reactor and a working fluid loop. The working fluid loop is configured to circulate a working fluid from a steam generator through the thermal energy storage vessel to absorb thermal energy and heat the working fluid for introduction to an electricity generating system. The solar energy collection system includes a heat transfer loop heated via a solar collector. The heat transfer loop is configured to circulate a heated heat transfer fluid to add thermal energy to the thermal mass composition in the thermal energy storage vessel.

The invention relates to a method for the radioactive decontamination of the internal face of the metal wall of an item of fluid-circulating equipment in a nuclear facility, the wall being made of a metal alloy and the internal face of this wall being covered with a surface layer of metal oxides, the method comprising at least one step a) of dissolving the surface layer of metal oxides, followed by a step b) of corroding the internal face of the wall, and being characterised in that: (1) steps a) and b) comprise circulating through the item of equipment a reducing aqueous solution at a temperature of at least 75°C; (2) step a) comprises, simultaneously with circulating the reducing aqueous solution through the item of equipment, biasing the wall to a potential lower than the corrosion potential of the metal alloy in contact with the reducing aqueous solution; and (3) the biasing of the wall is stopped at the end of step a). The invention is intended for use in the nuclear field, in particular for the decommissioning of items of metal equipment as conventional waste or for carrying out thorough decontamination of these items of equipment.