136 results about "Nuclear power reactor" patented technology

The present invention is directed to providing a method of producing synthetic fuels and organic chemicals from atmospheric carbon dioxide. Carbon dioxide gas is extracted from the atmosphere, hydrogen gas is obtained by splitting water, a mixture of the carbon dioxide gas and the hydrogen gas (synthesis gas) is generated, and the synthesis gas is converted into synthetic fuels and / or organic products. The present invention is also directed to utilizing a nuclear power reactor to provide power for the method of the present invention.

A nuclear fuel element for use in water-cooled nuclear power reactors and an improved multilayered silicon carbide tube for use in water-cooled nuclear power reactors and other high temperature, high strength thermal tubing applications including solar energy collectors. The fuel element includes a multilayered silicon carbide cladding tube. The multilayered silicon carbide cladding tube includes (i) an inner layer; (ii) a central layer; and (iii) a crack propagation prevention layer between the inner layer and the central layer. A stack of individual fissionable fuel pellets may be located within the cladding tube. In addition, a thermally conductive layer may be deposited within the cladding tube between the inner layer of the cladding tube and the stack of fuel pellets. The multilayered silicon carbide cladding tube may also be adapted for other high temperature, high strength thermal tubing applications including solar energy collectors.

The invention belongs to the field of composite shielding material technology, and especially relates to an aluminium-based composite material used for nuclear reactors to shield neutrons and gamma rays as well as a preparation method thereof. The composite material employs pure aluminium as a basis material, and shielding components are tungsten and boron carbide; the mass percent of tungsten is 20-70%; the mass percent of boron carbide is 1-10%; the balances are aluminium and unavoidable impurities. The aluminium-based composite material with certain mechanical strength as well as excellent nuclear reactor neutron and gamma ray irradiation resisting properties is developed by using an isostatic compaction method according to needs of some nuclear power reactors; the material whose thickness is 10cm can ensure that the transmittance of gamma rays whose power spectrum is 0.1-2MeV is reduced to 20% or less, and at the same time absorption effects of the composite material to thermal neutrons are substantially obvious.

Microstructured nuclear fuel adapted for nuclear power system use includes fissile material structures of micrometer-scale dimension dispersed in a matrix material. In one method of production, fissile material particles are processed in a chemical vapor deposition (CVD) fluidized-bed reactor including a gas inlet for providing controlled gas flow into a particle coating chamber, a lower bed hot zone region to contain powder, and an upper bed region to enable powder expansion. At least one pneumatic or electric vibrator is operationally coupled to the particle coating chamber for causing vibration of the particle coater to promote uniform powder coating within the particle coater during fuel processing. An exhaust associated with the particle coating chamber and can provide a port for placement and removal of particles and powder. During use of the fuel in a nuclear power reactor, fission products escape from the fissile material structures and come to rest in the matrix material. After a period of use in a nuclear power reactor and subsequent cooling, separation of the fissile material from the matrix containing the embedded fission products will provide an efficient partitioning of the bulk of the fissile material from the fission products. The fissile material can be reused by incorporating it into new microstructured fuel. The fission products and matrix material can be incorporated into a waste form for disposal or processed to separate valuable components from the fission products mixture.

A method for inspection and maintenance of an inside of a nuclear power reactor is provided, which can perform positioning of a remote operated vehicle relative to an inspection target surface in a nuclear power reactor in a short period of time with high accuracy, and enhance the positioning repeatability for the remote operated vehicle, thereby enabling performance of the inspection and maintenance for the inspection target surface securely and satisfactorily. For the inspection or maintenance for the inspection target surface 1a in a nuclear power reactor by using the remote operated vehicle 30, a detection mark 3 is formed in advance by providing a notching, marking, punching or engraving process to the inspection target surface 1a before performing inspection or maintenance of the inside of the nuclear power reactor by using the remote operated vehicle 30. Then, the remote operated vehicle 30 is driven to move in the nuclear power reactor filled with water. During the movement, the detection mark 3 formed in the inspection target surface 1a is detected by using the remote operated vehicle 30 so as to determine the position of the remote operated vehicle 30.

The present invention is directed to providing a method of producing synthetic fuels and organic chemicals from atmospheric carbon dioxide. Carbon dioxide gas is extracted from the atmosphere, hydrogen gas is obtained by splitting water, a mixture of the carbon dioxide gas and the hydrogen gas (synthesis gas) is generated, and the synthesis gas is converted into synthetic fuels and / or organic products. The present invention is also directed to utilizing a nuclear power reactor to provide power for the method of the present invention.

The invention discloses a forging method of a lower sealing head forge piece of a one-mega kilowatt nuclear-power reactor pressure vessel. In the method, a double-vacuum steel ingot made of 16MND5 with a weight of 103 tons is forged by using a hydraulic press of 12,000 tons, wherein the forging process sequentially comprises the following steps of: drawing: heating the double-vacuum steel ingot to 1,220+ / -10 DEG C, and changing a length of 2,625mm into 4,600mm; gas cutting blanking: removing a section of the bottom end of the forge piece to change the length of 4,600mm into 3,450mm, wherein the bottom of the double-vacuum steel ingot is guaranteed to be sufficiently cut off; upsetting: raising the temperature to 1,240+ / -10 DEG C and changing the length of 3,450mm into 290mm; rough machining: roughly machining the forge piece blank into a cake shape, wherein the diameter of the cake is 5,000mm, and the thickness is 220mm; and stamping and forming: heating the forge piece to 1,000+ / -10 DEG C and putting the forge piece on a special die for stamping and forming until the forging ratio reaches 1.2, and the forge piece forms a spherical crown shape. The invention can ensure that the forged lower sealing forge piece has the advantages of dense material, uniform component, reasonable metal streamline distribution and stable property.

The invention discloses a method for manufacturing a steel ingot used for million-kilowatt nuclear power reactor core structural component, austenitic stainless steel ingot that weighs over 25 tons is manufactured by electroslag remelting process: firstly, preparation: a single-phase bipolar series connected or triphase bipolar series connected electroslag remelting furnace is adopted; a copper wall water-cooling crystallizer and a water-cooled base plate are adopted; a consumable electrode is assembled; slag charge is matched; SiGa and Al powder are adopted for deoxidation; secondly, refinement and fusion casting: a graphite electrode is picked up and replaced by the consumable electrode to remelt the fully melted slag charge after slagging and slugging. When the actual weight of the steel ingot is 2 tons less than the preset weight, the electric power is reduced and feeding operation in telophase is carried out; thirdly, the steel ingot is stripped. The steel ingot manufactured by the invention has extremely high purity, extremely fine uniformity, excellent high temperature resistance and tarnish resistance, as well as strong anti neutron radiation embrittlement sensibility.

The invention relates to a method for refining grains of heavy forging steel of a nuclear reactor evaporator, which belongs to the technical field of heat treatment of steel. The method comprises the following steps of: performing electric furnace smelting and ladle seconclary refining during steel production; performing vacuum pouring by an improved vacuum carbon deoxidation (VCD) process or a silicon killed deoxidation process and a subsequent multi-furnace pouring (MP) process to obtain 200 to 600 tons of steel ingots; accurately controlling the components of the steel in the smelting process, wherein the mass fraction of aluminum elements is between 0.02 and 0.04 percent, the mass fraction of nitrogen elements is between 0.005 and 0.015 percent, the mass fraction of silicon elements is less than 0.1 percent when the improved VCD process is used, and the mass fraction of the silicon elements is between 0.1 and 0.3 percent when the silicon killed deoxidation process is used; and homogenizing a forging state structure by a two-time normalizing preheating treatment process after the ingot is subjected to forging molding, and refining the grains. The method has the advantages that:the grains used for manufacturing the heavy forging steel of the nuclear reactor evaporator are obviously refined; and the level of the grain size is improved from a level 5 to a level 8.5.

A radiation-monitoring diagnostic hodoscope system for producing an approximate image of radiation-detecting components within or external to a pressure vessel of an operating, damaged, or shutdown nuclear-power plant.

An construction workflow tracking system is presented where the tracking system is capable of tracking component-level information associated with achieving closure for an Inspection, Tests, Analyses, and Acceptance Criteria (ITAAC) under 10 CFR Part 52 for licensing of new nuclear power reactors. Disclosed systems include an estimation engine configured to analyze workflow data points to estimate one or more compliance metrics, especially compliance dates (e.g., inspection dates, test dates, acceptance dates, etc.). A reporting engine can present estimated compliance dates with a confidence level or recommendations on improving confidence or reducing risk. Compliance dates can also include compromise dates indicating when a closed ITAAC is likely to be compromised.

The invention discloses a heat-treatment technology method for a lower end socket forge piece of a million kilowatt grade nuclear power reactor pressure vessel by adopting 16MND5 alloy steel. The method comprises the following steps: heating a furnace to 400-500 DEG C, putting a forge piece in the furnace and keeping the temperature for 1-3h; continuously heating the furnace, wherein the rate of temperature increase is less than or equal to 80 DEG C / h; when the furnace is heated to 650-700 DEG C, keeping the temperature of the forge piece for 3-5h; continuously heating the furnace at a power rate of temperature increase, and when heating to 880-900 DEG C, soaking the forge piece and keeping the temperature for 5-8h; lifting the forge piece out from the furnace and putting into circulatingwater for cooling for at least 90 minutes, wherein the quantity of water supplied per hour is larger than or equal to 1000 tons; after the furnace is heated to 300-350 DEG C, putting the forge piece into the furnace and keeping the temperature for 1-3h; continuously heating the furnace, wherein the rate of temperature increase is smaller than or equal to 60 DEG C / h; when the furnace is heated to 640-660 DEG C, keeping the temperature of the forge piece for 5-8h; and lifting the forge piece out from the heat-treatment furnace, and carrying out air cooling until reaching the room temperature. The invention can enable the final mechanical property of the forge piece to reach the requirements of the related technical specification.

The invention discloses a hydrogen-explosion-preventing serious nuclear accident relieving device and a hydrogen-explosion-preventing serious nuclear accident relieving method, and belongs to the fields of safety equipment and technology of a nuclear power station. A hydrogen adsorption device is installed on the inner wall and in an upper space of a safety shell; by using the physical adsorptionproperty of a carbon nanofiber material, the hydrogen adsorption device is used for adsorbing and storing hydrogen without the help of external power; when a large amount of hydrogen is generated in the safety shell, the hydrogen adsorption device reduces the concentration of the hydrogen in the safety shell by adsorbing the hydrogen continuously and igniting the hydrogen, so that the explosion of the hydrogen is avoided; and the pressure in the safety shell is reduced, so that the aim of relieving a serious accident of a reactor is fulfilled. In the running process of the hydrogen-explosion-preventing serious nuclear accident relieving device, the external power is not required; the device can work stably for a long time, is reliable in performance, good in spare safety, convenient to implement and easy to control and meets the design requirements of a current novel nuclear power reactor.

A remotely controllable apparatus and technique for the removal of “CRUD” from surfaces within a nuclear Reactor Pressure Vessel includes a flexible, high pressure water jetting lance assembly for delivering intense water pressure (up to 20,000 psi) via nozzle(s) directly to the areas containing the highly radioactive “CRUD” material. The jetting lance assembly is positioned to the contamination site with either a guide tube having any predetermined shape and into which the lance assembly is disposed or a positioning member attached to the lance assembly, either of which may be torqued and manipulated until the nozzle end of the lance assembly is in position. Various other embodiments include multiple nozzles, adapters that have a specific shaped lumen, one or more sensor devices is communication with processing units near the proximal end of the apparatus.

A floating nuclear power reactor includes a self-cooling containment structure and an emergency heat exchange system. The containment structure of the reactor may be flooded upon the temperature or pressure in the containment structure reaching a certain level. The reactor vessel may also be flooded upon the temperature or pressure in the reactor vessel reaching a predetermined level. The reactor includes a heat exchange system and a filtered containment venting system. The reactor also includes a multi-compartment containment structure. Multiple steam by-pass pipes extend to the filtered containment vent chamber.

The invention relates to a chromium-aluminum ceramic alloy material, discloses a manufacturing plate of a shielding plate for a new fuel storage rack, a spent fuel storage rack and a fuel temporary storage rack of an AP1000 nuclear power plant and particularly relates to the manufacturing method of a boron carbide particle-reinforced aluminum-based composite material neutron absorption shielding plate for a nuclear power reactor. The preparation method comprises the following steps of: mixing materials according to the proportion, adopting a two-way extrusion technology and pressing for forming a high-density cylindrical biscuit; placing the biscuit formed by pressing into a heating furnace for preheating, clamping a stock which is uniformly preheated into an extrusion mold which is preheated to the temperature of 400 DEG C, extruding the plate at the pressure of 7000MPa by using a 2000KN press, and finishing the extruded stock to get the neutron absorption shielding plate.

The invention discloses a manufacturing process of large 14Cr17Ni2 stainless steel forgings. The process comprises steps as follows: (1) optimization of chemical components: N elements with the content wt being 0.055%-0.065% are added, the content wt of Ti elements are controlled to be smaller than or equal to 0.010%; (2) forging: the large 14Cr17Ni2 stainless steel forgings are forged and heated at 1,100-1,180 DEG C, and the finish forging temperature is controlled to be larger than or equal to 950 DEG C; (3) thermal treatment: the large forgings are subjected to annealing, quenching and tempering, wherein annealing comprises steps as follows: after forging ends, the forgings are subjected to thermal insulation in a waiting furnace; a gate is closed, the furnace is cooled, and thermal insulation is performed; the temperature is increased to 690 DEG C plus or minus 10 DEG C for thermal insulation; the forgings are cooled and discharged from the furnace, and annealing is completed; quenching and tempering comprises steps as follows: the forgings are heated to 1,020 DEG C plus or minus 10 DEG C for thermal insulation, oil cooling is performed after the forgings are discharged from the furnace, and tempering is performed; the forgings are heated to 600 DEG C plus or minus 10 DEG C for thermal insulation and immediately subjected to oil cooling after being discharged out of the furnace. With the adoption of the manufacturing process, the problem that cracks are easily caused during forging of the large 14Cr17Ni2 forgings can be solved, mechanical properties of the large forgings are remarkably improved, and the use requirements of internal components of nuclear power reactors are completely met.

The invention discloses a heat treatment method of a martensite stainless steel forge piece for a nuclear power reactor internals pressure spring, comprising the following steps: step 1, preheating: rapidly heating up a workpiece to 990-1030 DEG C with a speed of 90-110 DEG C / h, preserving heat, and then carrying out air cooling on the workpiece; step 2, quenching: rapidly heating up the workpiece to 950-990 DEG C with a speed of 90-110 DEG C / h, preserving heat, and then carrying out oil cooling to cool the workpiece surface to 100 DEG C; and step 3, high temperature tempering: rapidly heating up the workpiece to 630-670 DEG C with a speed of 90-110 DEG C / h, preserving heat, and then carrying out air cooling on the workpiece to cool to room temperature. According to the invention, by adding the preheating step and optimizing the technological parameters of quenching and tempering, the impact toughness is successfully raised to 120-160 J while raising the yield strength, thus the comprehensive performance of the forge piece is raised.

The invention discloses a heat treatment method for austenitic stainless steel pie forgings for nuclear power reactor internals. The method is used for performing heat treatment on a solid forging, the material of the solid forging is 0Cr19Ni10, the diameter of the material is not smaller than 4000 mm, and the thickness of the material is not smaller than 400 mm. The method comprises the following steps: step 1, heating the forging to 400-450 DEG C for preserving heat by using the maximum power of a heat treatment furnace, wherein the preserving heat time is 0.5-1 minute for each millimeter of forging thickness; step 2, heating the forging to 670-700 DEG C for preserving heat at the temperature rise rate of at most 50 DEG C / h, wherein the preserving heat time is 0.5-1 minute for each millimeter of forging thickness; step 3, heating the forging to 1040-1080 DEG C for preserving heat at the temperature rise rate of at most 150 DEG C / h, wherein the preserving heat time is 3-5 minutes for each millimeter of forging thickness; step 4, discharging the forging from the furnace and cooling with water. The crystalline grains of the 0Cr19Ni10 steel forging subjected to heat treatment by the method are not grown, organization is uniform, and the strength of room temperature and 350 DEG C of high temperature meets the design requirement of the nuclear power reactor internals.

A system for generating a source of neutrons from a thermonuclear fusion reaction includes a reaction chamber and a number of particle beam emitters. The reaction system has at least four particle beam emitters supported spatially around oriented toward a common focal region of the reaction chamber for directing the plurality of plasma beams that are spatially symmetrical in three dimensional space. Each of the plasma beams are directed towards a plasma region in the geometric center. A stable collapse of the plasma region permits a controllable and sufficiently long confinement time, which in combination with necessary temperature and density conditions may ignite and sustain fusion reactions and achieve a net energy output. Optionally, laser beams or other input energy devices may also be oriented around and toward the common focal region to direct high-energy laser beams at the plasma ball to assist with instigation of the fusion reaction. The thermonuclear reaction system may be used as a neutron source for nuclear power reactors.

the invention discloses a novel reactor control rod and rod pair. The novel reactor control rod and rod pair comprises an outer-layer rod body and an inner-layer rod body capable of sliding relative to the outer-layer rod body; the outer-layer rod body is vertically through and is composed of neutron absorbers A and non-neutron absorbers B in the axial direction of the outer-layer rod body, wherein the total number of the neutron absorbers A and the non-neutron absorbers B is 2n, and the neutron absorbers A and the non-neutron absorbers B are equal in length and are arranged in a staggered mode; the inner-layer rod body is composed of neutron absorbers C and neutron absorbers D in the axial direction of the inner-layer rod body, wherein the total number of the neutron absorbers C and the neutron absorbers D is 2n+1, and the neutron absorbers C and the neutron absorbers D are equal in length and are arranged in a staggered mode; the mass distribution of the neutron absorbers A of the outer-layer rod body is non-uniform distribution; the mass distribution of the neutron absorbers C of the inner-layer rod body is non-uniform distribution, according to the specific mass distribution design, after the rod pair is inserted into a reactor, the distorted peak of the axial power of a reactor core can be decomposed and flattened, and the decoupling control over the power level and the axial power deviation of he reactor core can be realized. Compared with a traditional uniform control rod of which the complex rod lifting and inserting procedures are needed for the decoupling controlover the power level and the axial power deviation of the reactor core, the novel reactor control rod and rod pair have the advantages of being simpler, more convenient to use and efficient and has higher application value on the control over the nuclear power reactor.

The invention discloses a tripping oil-return system for a steam turbine for a nuclear power station, which is used to trigger a protective system of the steam turbine to operate when a nuclear power reactor or the steam turbine breaks down so that a cut-off valve of the steam turbine protects that an oil return path returns oil rapidly to release pressure and shutdown. In the tripping oil return system for the steam turbine for the nuclear power station, three groups of oil valves are arranged. In each group, two oil return chambers of the oil valves are connected in series. The oil return chambers among the groups are connected in parallel. Three electric control assemblies are arranged. The electric control assembly I controls the oil values of the oil valve groups II and III, the electric control assembly II controls the other oil valve of the oil valve group III and one oil valve of the oil valve group I and the electric control assembly III controls the other oil valve of the oil valve group I and the other oi valve of the oil valve group II. In the invention, oil return of the oil valves is controlled by the three electric control assemblies. When two or more than two of the three electric control assemblies are in a good condition, the electric control assemblies return oil normally when receiving a protective action signal of the steam turbine. When only one electric control assembly receives the protective action signal of the steam turbine, the electric control assembly cannot return oil. Therefore, the malfunction probability can be effectively reduced and both operational safety and reliability of the steam turbine are considered.

The invention relates to a portable passive nuclear power reactor, and belongs to the technical field of reactors. The portable passive nuclear power reactor comprises a shell, a neutron reflecting layer, a thermoelectric converter, a hydrogen inlet valve, a hydrogen regulator, fuel, a hydrogen conduction pipe, a heat pipe group and a current output terminal, wherein the shell defines a cylinder structure with an accommodating chamber; the accommodating chamber is divided into a hydrogen regulation chamber and a fuel chamber by the neutron reflecting layer; the thermoelectric converter is positioned on the upper part of the fuel chamber; the hydrogen inlet valve is positioned on the top of the shell; the hydrogen regulator is positioned in the hydrogen regulation chamber; the fuel is positioned in the fuel chamber; the hydrogen regulation chamber is communicated with the fuel chamber by the hydrogen conduction pipe; the heat pipe group is communicated with the thermoelectric converter and the fuel; the current output terminal is positioned outside the accommodating chamber, and penetrates through the shell to be communicated with the thermoelectric converter. According to the reactor disclosed by the invention, by utilizing the characteristic that hydrogenation and dehydrogenation performance of metal uranium can be changed along with the temperature, the design of the passive reactor utilizing the hydrogen as a neutron moderator and an operation control agent is realized.

The invention provides a method for monitoring 16N leaking from a nuclear power reactor vapor generator. The method comprises the following steps: S01. reserving a scale peak of a stable gamma energy spectrum after a scintillant detection device internally containing a gamma source is stripped at the background, monitoring 16N leaking from the nuclear power reactor vapor generator by the scale peak and generating a monitoring signal; and S02. processing the monitoring signal by using a signal processing machine, and outputting a monitoring result. By using a scintillant internally containing the gamma source, the uncertainty factor of the scintillant caused by packaging process operations such as an embedded source can be avoided, and the technical problem that a monitoring result is influenced due to false alarm easily caused by low sensitivity of the monitoring system is solved.

Disclosed is a zirconium alloy material for a nuclear power reactor core, falling within the technical field of special alloy materials, and comprising the following components in percentage by weight: 0.60-1.00% of Sn, 0.80-1.10% of Nb, 0.10-0.40% of Fe, 0-0.1% of Cu or Bi or Ge, 0-0.03% of Si or S, 0.06-0.15% of O, less than 0.008% of C, less than 0.006% of N and the balance being Zr. Based on a Zr-Sn-Nb system alloy, other components used for improving the performance of the alloy are added, and appropriate component contents thereof are selected, so that the performance of the alloy provided meets the requirements of high burnup of the nuclear power reactor on the structural materials of the core. A product prepared from this alloy improves uniform corrosion resistance in ex-core pure water, especially in a lithium hydroxide aqueous solution, and improves nodular corrosion resistance in high-temperature steam.

A floating nuclear power reactor includes a self-cooling containment structure and an emergency heat exchange system. The containment structure of the reactor may be flooded upon the temperature or pressure in the containment structure reaching a certain level. The reactor vessel may also be flooded upon the temperature or pressure in the reactor vessel reaching a predetermined level. The reactor includes a heat exchange system and a filtered containment venting system.

The invention discloses a heat treatment process of million-kilowatt nuclear power reactor pressure vessel reactor core cylinder forgings made of 16MND5 alloy steel, which comprises: raising the temperature of a furnace to 400 to 450 DEG C, placing the forgings in the furnace and keeping the temperature for 2 to 5 hours; continuing to heat the furnace at a temperature rise speed of less than or equal to 80 DEG C per hour; when the temperature of the furnace rises to 670 to 700 DEG C, keeping the temperature of the forgings for 4 to 6 hours; continuing to heat the furnace at a power temperature rise speed, and keeping the temperature of the forgings for 6 to 10 hours after the temperature of the forgings is uniform when the forgings are heated to 870 to 900 DEG C; lifting the forgings out of the furnace and placing the forgings in circulating water for cooling, wherein the water supply per hour is more than or equal to 1,500 tons; raising the temperature of the furnace to 300 to 350 DEG C, placing the forgings in the furnace, and keeping the temperature of the forgings for 2 to 5 hours; continuing to heat the furnace at a temperature rise speed of less than or equal to 60 DEG C per hour; when the temperature of the furnace rises to 635 to 665 DEG C, keeping the temperature of the forgings for 6 to 10 hours; and lifting the forgings out of the heat treatment furnace, and cooling the forgings in the air to room temperature. When the method is used, the final mechanical performance of the forgings can meet the requirements of related technical standards.

The invention provides a parameter uncertainty analysis method and system for a nuclear power reactor system, and the method comprises the steps: building a system evaluation model according to a system design value through employing an optimal estimation program, and carrying out the steady-state debugging of the system evaluation model; carrying out sensitivity analysis on the input parameter / physical model by utilizing a system evaluation model to obtain a key input parameter / key physical model, and obtaining a target change range of the key input parameter / key physical model according to characteristics of the key input parameter / key physical model; constructing a target parameter sample point set according to a target change range of the key input parameter / key physical model, and performing uncertainty analysis on the target parameter sample point set by using a system evaluation model to obtain an uncertainty quantification result of the key input parameter / key physical model, according to the method, the excessive conservative allowance of the nuclear power reactor system is effectively released, and the economical efficiency of the nuclear power reactor system is improved to the maximum extent.