136 results about "Nuclear power reactor" patented technology

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 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.

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.

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.

A floating nuclear power reactor including one or two nuclear power reactors positioned in a floating vessel such as a barge or the like. Means is disclosed for flooding the containment structure of the nuclear reactor and for flooding the reactor vessels to cool the same.

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.