39results about How to "Small absorption cross section" patented technology

The present invention discloses an end pumped order variable gradient adulterate composite lath laser amplifier, which comprises a laser gain medium, seed source, pumping source, coupling lens duct, lath cooling heat sink and indium transition layer, characterized in that the laser gain medium is N-order order variable gradient adulterate composite ceramic lath and is composed by 2N+3 segments that is sub-lath in middle section that has highest luminous ion concentration decreases, each N segment sub-lath with luminous ion concentration decreases gradually towards two ends, and sub-lath with zero adulterate on two ends, wherein N is positive integer. The present invention can improve temperature distribution homogeneity and decrease biggest thermal stress, and further realize laser output of high average power and high beam quality.

The invention aims at providing a nuclear fuel rod with ceramic cladding and metallic pellet. The nuclear fuel rod with ceramic cladding and metallic pellet has the advantages that the heat-conducting property of the nuclear fuel rod can be improved, the central temperature of the fuel pellet is lowered or the fuel power density is effectively improved, the fuel rod cladding mechanical property, corrosion resistance and radiation resistance are greatly improved, the breakage rate of the fuel rod is lowered, the discharge burnup of the fuel rod is remarkably increased and the fuel utilization rate is greatly improved; meanwhile, the fuel rod has good neutronics performance, and can be widely used in a fast reactor or a thermal reactor. The cladding of the nuclear fuel rod provided by the invention is a SiCf/SiC composite ceramic material cladding, the metallic fuel pellet adopts a U-Zr or U-Pu-Zr alloy fuel pellet; a gap between the cladding and the fuel pellet is filled with low-melting-point metal; and a lengthened air cavity is designed at the upper part of the fuel rod.

The invention relates to a refractory high-entropy alloy material with high plasticity and low neutron absorption cross section and a preparation method thereof, and belongs to the technical field of high-entropy alloys. The high-entropy alloy material is AlVTiNbZr high-entropy alloy, wherein the atomic percent of an aluminum element is 5%-10%, the atomic percent of a V element is 3%-15%, the atomic percent of a Ti element is 20%-40%, the atomic percent of an Nb element is 25%-35%, and the atomic percent of a Zr element is 20%-35%. The AlVTiNbZr high-entropy alloy is of a single-phase BCC structure. The AlVTiNbZr high-entropy alloy is obtained in a vacuum arc melting mode, has a high theoretical melting point, low alloy density, high room-temperature yield strength and deformation plasticity, can resist deformation failure in the service process, and brings sufficient safety margin in the service process.

The invention discloses a preparation method of a copper-based neutron absorbing material. Specifically, the spark plasma sintering technique is used for applying pulse currents to mixed powder materials, so that joule heat and point discharge plasma are generated between particles, and the particles are made to melt locally to be connected in a metallurgical bonding mode; and the chemical materials used in the preparation process mainly include copper powder, boron carbide powder, absolute ethyl alcohol, hydrochloric acid, hydrofluoric acid, deionized water, graphite paper, sand paper and diamond paste. The preparation method comprises the steps of pretreatment of the copper powder, pretreatment of the boron carbide powder, weighing of the raw materials, ball milling of mixed materials, filling of powder, prepressing of a mold, spark plasma sintering for preparation, cleaning of the surface of a sample, detection and analysis of the sample and the like. The compactness of the prepared copper-based neutron absorbing material can reach 99% or over, the boron carbide in the copper-based neutron absorbing material is distributed in a dispersed mode, the bending strength can reach 350+/-5 MPa, and the neutron absorbing performance can reach 98%.

The invention discloses ZrC inert matrix dispersed pellet nuclear fuel as well as a preparation method and application thereof and aims at solving the problems in the prior art that since thermal conductivity of an inert matrix of dispersed nuclear fuel is reduced rapidly under the irradiation condition, fault-tolerant performance is reduced and reactor safety is influenced. The ZrC inert matrix dispersed pellet nuclear fuel comprises the inert matrix, a fuel element and a fuel-free zone protective layer, wherein the inert matrix is ZrC, and the fuel-free zone protective layer is made from ZrC. The preparation method comprises the following steps: coating ZrC mixed slurry on TRISO particles by virtue of a spray deposition process to form a compact ZrC inert matrix, carrying out mould pressing on ZrC mixed powder to form a fuel-free zone biscuit, then carrying out complex pressing on the inert matrix and the fuel-free zone biscuit to form an IMDP biscuit, then sintering the IMDP biscuit in a vacuum furnace at high temperature, cooling the furnace, and then carrying out machining on a sintered billet according to the required size, so that a fuel assembly in a water reactor and a high temperature gas cooled reactor is obtained.

The invention relates to a portable neutron detection device. The device comprises an outer sleeve (2), as well as a neutron scintillator conversion screen (5), a photoconductive material (6), an end window photomultiplier (7) and a base (3) which are sequentially arranged in the outer sleeve (2) and are in compression fitting joint, wherein an output port (4) extruding at the outer end of the base (3) is formed at one end of the end window photomultiplier (7), and connected with a preamplifier (9), a main amplifier (10) and a multi-channel analyzer (11) in sequence through leads; a screwable protective cover (1) with an inner thread is arranged at the front end of the outer sleeve (2); the protective cover (1) and the outer sleeve (2) are made of an aluminum material; the screwing progress can be adjusted by rotating the protective cover (1); and a magnetic shielding material (8) is coated on the outer surface of the end window photomultiplier (7). The portable neutron detection device can effectively shield an outer optical signal, has high efficiency in neutron detection, and is simple in structure and convenient to carry.

The invention discloses a modularized transverse prism type air-cooled micro-reactor core system which comprises a fuel assembly, a drum control rod, an absorber ball and a reflecting layer; the reflecting layer covers a reactor core formed by the fuel assembly; the drum control rod is arranged on the side reflecting layer; the reactor core is provided with a central graphite ribbon; an absorber ball channel is arranged in the central graphite ribbon; the fuel assembly is provided with fuel rod channels, coolant channels and beryllium oxide rods which are regularly arranged at intervals so as to enhance moderation. The system has the beneficial effects that the fuel is microencapsulated by the ceramic so that the fuel is prevented from being eroded; the system has the advantages of high heat capacity, high temperature resistance, high heat conductivity, high slowing-down ratio, small thermal neutron absorption cross section and the like; the inherent safety of automatic thermal shutdown only depending on temperature negative feedback is achieved; the drum control rod and the absorber ball not only can effectively control reactivity and ensure the safety of the reactor core, but also can save space, so that the reactor core system and the reactor can be arranged in a common container and are convenient to transport.

The invention relates to a porous rare earth titanate heat-isolating material and a preparation method and an application thereof. The porous rare earth titanate heat-isolating material is of a porousstructure and the composition of the material is (Re2O3)X(TiO2)(1-X), wherein Re is at least one of Y and lanthanide elements, and X is equal to 0.2-0.8, preferably 0.4-0.7. The porosity of the porous rare earth titanate heat-isolating material is 10-90% and the pore diameter is 0.1-500 microns.

The invention provides a compact reactor core applied to a land mobile power supply and the land mobile power supply, relates to the technical field of nuclear reactors, and is designed for solving the problem that the structure of the existing land small-sized gas-cooled nuclear reactor is not compact enough. The compact reactor core applied to the land mobile power supply comprises a reactor core active area, a reflecting layer wrapping the periphery of the reactor core active area and a pressure bearing layer wrapping the periphery of the reflecting layer, the reactor core active area comprises a plurality of hexagonal fuel elements, and the plurality of fuel elements are closely arranged in a honeycomb shape; the center of the fuel element is a fuel rod with a circular cross section, and the hexagon is a coolant inlet channel; the fuel rod comprises uranium carbide fuel; the cooling agent is helium and xenon mixed gas. The land mobile power supply comprises the compact reactor core applied to the land mobile power supply. The compact reactor core applied to the land mobile power supply and the land mobile power supply provided by the invention are more compact in structure on the premise of the same power.

The invention relates to a zirconium/niobium/copper/bismuth alloy used as such structural materials as a fuel cladding of a pressurized water reactor nuclear power plant and a positioning grillwork strip, and belongs to the technical field of zirconium alloy materials. The zirconium alloy comprises the following chemical components in percentage by weight: 0.7-1.2% of Nb, 0.05-0.6% of Cu, 0.05-1.0% of Bi, and the balance of Zr. The preferential range of the alloy elements is as follows: 0.8-1.2% of Nb, 0.1-0.4% of Cu, and 0.1-0.4% of Bi. The zirconium alloy is excellent in corrosion resistance in superheated steam of 400 DEG C/10.3 MPa and de-ionized water of 360 DEG C/18.6 MPa, is obviously superior to a Zr-1Nb alloy, is excellent in machinability, and can be used as such core structural materials as the fuel element cladding and the positioning grillwork strip in a pressurized water reactor of the nuclear power plant.

The invention discloses a structure for increasing one-way gain by side pumping and an implementation method. A non-absorptive cladding material is bonded on the outer side of a gain medium, so that pumping light is reflected on an upper reflecting surface, a lower reflecting surface and side surfaces of a cladding; and the two side surfaces have a relative angle, the light travels back and forthalong the gain medium while being reflected by the side surfaces, so that the light can pass through the gain medium for multiple times, the absorption of the gain medium is increased, and the size ofthe gain medium is effectively reduced while the pumping light absorption frequency is increased. Because the pump light is reflected for multiple times and passes through the gain medium, the energydistribution of the pump light is more uniform, the cladding material has high thermal conductivity, the temperature gradient of the gain medium can be effectively reduced, the possibility of damaging a device is reduced, and the quality of an output light beam is improved.

The invention belongs to the field of nuclear material preparation and provides a preparation technology of a neptunium oxide target pellet for solving the problems that an existing preparation technology of a neptunium oxide target is relatively low in overall neptunium loading capacity, relatively high in production cost and relatively complicated in process. The technology comprises the following steps: (1) taking neptunium oxide powder and magnesium oxide powder as a target pellet material; (2) taking out part of neptunium oxide powder and mixing with the magnesium oxide powder to obtain preliminarily mixed powder; (3) adding the remaining neptunium oxide powder and further mixing with stearic acid to obtain mixed powder; (4) carrying out cold press molding to prepare a target pellet body; and (5) carrying out pressureless sintering to prepare the neptunium oxide target pellet. The preparation technology has the advantages of being simple in technological process and low in cost; the prepared neptunium oxide target pellet is high in overall neptunium loading capacity and high in neutron utilization rate, has relatively high thermal conductivity and good in-pile irradiation performance, and is convenient to chemically dissolve and extract and suitable for massive production of 238Pu fuel.

The invention discloses a ZrCSiC ceramic and TC4 titanium alloy brazing method, and relates to a ZrCSiC ceramic and TC4 titanium alloy brazing method. The problems that when existing ZrCSiC ceramic and TC4 titanium alloy are brazed, wettability of brazing filler metal to a base metal is poor, brittle compounds are prone to being generated in brazing seams, a metal base metal is excessively dissolved, and connector residual stress is large are solved. The method comprises the steps: adding ZrH2 powder into Cu powder, wherein graphene grows on the surface of the Cu powder, and adding ZrH2 for tabletting to obtain a brazing filler metal foil; and performing brazing after assembling. According to the method, the vertical few-layer graphene grows on the surface of the Cu powder in situ, so thatthe reinforced composite brazing filler metal containing the graphene is prepared, the proportion of brittle compounds in brazing seams is reduced, the joint residual stress is relieved, and the wettability of the brazing filler metal to base metal is improved. Generation of the brittle compounds is avoided, and the problem of excessive dissolution of the base metal is solved. The method is suitable for brazing the ZrCSiC ceramic and the TC4 titanium alloy.

The invention discloses a neutron deceleration composite material. The neutron deceleration composite material comprises a mixed material, a material which accounts for 0.1-5% of the weight of the mixed material and contains a 6Li element, and Cu which accounts for 0.1-5% of the weight of the mixed material, and the mixed material comprises one or a mixture of more of Al, MgF2 and AlF3; the mixed material comprises the following components in percentage by mass: 10-95% of pure aluminum powder, 20%-80% of aluminum fluoride powder, 0.1%-15% of magnesium fluoride powder; the material containing the < 6 > Li element is < 6 > LiF powder; the mixed material further comprises pure magnesium powder, and the mass percent of the pure magnesium powder is 0.1-5%. The radioactivity of the material can be attenuated to an exemption level or below within a short time when the material is decommissioned at the last stage of the device or equipment, and the neutron deceleration composite material has obvious advantages in the aspects of radioactive waste disposal method and economy.

The invention provides a zirconium alloy. The zirconium alloy comprises by weight 1.1%-2.1% of Nb, 0.2%-1.5% of Sn, 0.02%-0.1% of Cr, 0.01%-0.1% of Ni, 0.02%-0.1% of Ce and the balance of Zr. Nb is added into the zirconium alloy, and Nb is an alloy element which can improve corrosion resistance and mechanical property of the zirconium alloy; N is a harmful impurity element in the zirconium alloy due to the fact that N3- can replace oxygen ions in oxide crystal lattices and generate additional vacancy, after Sn is added into the zirconium alloy, N3- and oxygen ion vacancy attempt to be reservednear Sn3+; ions, and mobility is poor after the three components are combined, so that vacancy mobility is reduced, and Sn can counteract harmful effects of N and improve corrosion resistance; Cr andNi are further added into the zirconium alloy, Cr can effectively improve the creep resistance of the high alloy, and Ni can effectively improve the high-temperature steam corrosion resistance of thehigh alloy.

The invention discloses a preparation method of a SiC-based composite ceramic solidified body, which is characterized by the following steps: taking ZnO, CaO, Al2O3 and SiO2 as raw materials, ball milling, drying, calcining, ball milling and drying to obtain ZCAS glass; using silicon powder, graphite and ZCAS glass as raw materials, ball milling, drying and calcining to obtain SiC-ZCAS composite powder; adding polyvinyl alcohol sol, mixing, granulating and pressing and forming; performing heat treatment on the formed blank body and placing the blank body into a vacuum hot-pressing sintering furnace for sintering, so as to obtain the SiC-based composite ceramic solidified body. The invention adopts safe and cheap raw materials, simple and practical vacuum hot pressing sintering technology is used to prepare the SiC-based composite ceramic solidified body with excellent radiation stability, chemical stability, mechanical stability and thermal stability at lower temperature. The SiC-basedcomposite ceramic solidified body prepared by the invention is suitable for safely treating high-radioactive graphite waste generated due to decommissioning of a nuclear reactor.