31results about How to "Prevents stress corrosion cracking" patented technology

A photocatalytic substance having the properties of an n-type semiconductor is deposited on a surface of a metal base made of a stainless steel or Inconel. When necessary, the hydrogen concentration of the reactor water is increased. A current produced by the photocatalytic substance when the same is irradiated with light or radioactive rays in a nuclear reactor flows through the metal base to reduce corrosion current. When necessary, the photocatalytic substance is provided on its surface with at least one of Pt, Rh, Ru and Pd.

A laser peening process for the densification of metal surfaces and sub-layers and for changing surface chemical activities provides retardation of the up-take and penetration of atoms and molecules, particularly Hydrogen, which improves the lifetime of such laser peened metals. Penetration of hydrogen into metals initiates an embrittlement that leaves the material susceptible to cracking.

The method of preventing cold-worked stress corrosion cracking in the iron steels and alloyed steels including stainless steels comprises removing a residual tensile stressed layer, optionally generating residual compression, through the impact of irradiation with the ultra-short femtosecond pulse and kW class high average-power laser, and removing a hardened surface layer extremely susceptible to stress corrosion cracking, whereby this susceptible hardened surface layer with many concentrated dislocation defects and the likes is removed without generating or introducing any of such hardened layer in the process of its removal. The method is free from the risk of damaging other equipment by peening with steel balls or other shaped shots, is almost unlimited in actual environment, effectiveness, work, place and other conditions for application, is inexpensive, and functions continuously for a much longer period than an ordinary nuclear reactor's service lifetimes.

In a method of reducing corrosion of a material constituting a nuclear reactor structure, an electrochemical corrosion potential is controlled by injecting a solution or a suspension containing a substance generating an excitation current by an action of at least one of radiation, light, and heat existing in a nuclear reactor, or a metal or a metallic compound forming the substance generating the excitation current under the condition in the nuclear reactor to allow the substance generating the excitation current to adhere to the surface of the nuclear reactor structural material, and by injecting hydrogen in cooling water of the nuclear reactor while controlling the hydrogen concentration in a feed water.

A device for isolating, accumulating, indicating and venting gas in a fluid system pipe includes a coupling affixed to a system pipe. The coupling includes an isolation valve. A standpipe is attached to the coupling. The standpipe holds a float trapped between a flow retaining orifice and a closed upper end save for an angled hole therein that allows gas to flow around the float. An indicator exterior to the pipe indicates the float's level in the standpipe regardless of system pressure changes. A vent valve attached above the standpipe allows controlled ventilation of the gas flowing from the system pipe through the standpipe and through the vent valve. Accumulation of gas from the system pipe lowers the float in the standpipe, at which point the user vents the gas, causing the float to rise with the rising fluid level.

A control rod for nuclear reactors includes four wings including neutron absorbers containing hafnium, a front end structural member which has a cross shape in cross section and includes brackets bonded to the leading ends of the wings, and a terminal end structural member which has a cross shape in cross section and includes brackets bonded to the tailing ends of the wings. The four wings are bonded to a wing-bonding member including a cross-shaped center shaft so as to form a cross shape. The front end structural member and the wing-bonding member are made of a zirconium alloy. The wings include neutron-absorbing plates having neutron-absorbing portions and each have an outer surface which is opposed to a fuel assembly and at which a hafnium-zircaloy composite member covered with zircaloy is disposed. The neutron-absorbing plates are opposed to each other with trap spaces disposed therebetween.

In order to reduce a pressure loss in a check valve and simplify a manufacturing process, a check valve (10) configured to be disposed in piping of fluid is provided, which includes a valve housing (31) including a valve element (52) and a valve seat (54) therein, a casing (20) that is configured to place the valve housing therein, in which a casing flow path (60) is provided between the casing and the valve housing to make a flow of the fluid in an axial direction of the valve housing. In the valve housing, a communicating hole (38) and sealing portion (34) are formed. The communicating hole that is formed through the valve housing in a thickness direction thereof introduces the fluid, which flows through between the valve element and the valve seat at the open position of the check valve, into the casing flow path. The sealing portion that is provided on an outer surface of the valve housing upstream of the communicating hole in a flow direction of the fluid in the casing flow path, is configured to come into contact with an inner surface of the casing to seal the casing flow path and position the valve housing in a radial direction perpendicular to the axial direction relative to the casing.

In a method of reducing corrosion of a material constituting a nuclear reactor structure, an electrochemical corrosion potential is controlled by injecting a solution or a suspension containing a substance generating an excitation current by an action of at least one of radiation, light, and heat existing in a nuclear reactor, or a metal or a metallic compound forming the substance generating the excitation current under the condition in the nuclear reactor to allow the substance generating the excitation current to adhere to the surface of the nuclear reactor structural material, and by injecting hydrogen in cooling water of the nuclear reactor while controlling the hydrogen concentration in a feed water.

The invention discloses a method for connecting a pressure vessel nozzle and a head for reducing welding residual stress, and belongs to the technical field of sealing vessel processing and manufacturing. It is characterized in that it includes the following steps: a. drilling the head and processing the connecting pipe (2); b. extending one end of the connecting pipe (2) into the pressure vessel (1); c. welding the sealing head and the connecting pipe (2); d. Preheat the welding seam (3) on one side of the pressure vessel (1), and then surfacing welding at the root of the connecting pipe (2) in the pressure vessel (1); e. After preheating the inner wall of the connecting pipe (2) Then surfacing welding on the inner wall of the connecting pipe (2); f, cutting the connecting pipe (2) in the pressure vessel (1); g, tempering the head. The pressure vessel nozzle and head connection method for reducing welding residual stress reduces the welding residual stress during the pressure vessel manufacturing process, improves the structural integrity of the pressure vessel, and enhances the stress corrosion cracking resistance of the pressure vessel nozzle welded joint.

An object of the present invention is to prevent stress-corrosion cracking of a header (40) of a reactor. A reactor for producing trichlorosilane by causing metal silicon powder and a hydrogen chloride gas to react with each other includes a cooler (70), the cooler including a plurality of heat transfer medium pipes (30) and a header (40), the plurality of heat transfer medium pipes being provided in a fluid bed (60) inside the reactor, the header being provided in a freeboard section (50) inside the reactor, the header being comprised of a corrosion-resistant material.

The alloy powder capable of producing compression stress in the fused-on layer consists of Cr 6.0-12.0 wt%, Ni 5.0-11.0 wt%, Mn 0.5-1.8 wt%, Mo 0.1-0.9 wt%, Nb 0.1- 0.9 wt%, Ti 0.1-0.9 wt%, Si 0.1-0.9 wt%, except Fe. It has grain size of 50-200 microns, solid phase change temperature after melting and solidifying of 100-300 deg.c, and structure comprising martensite in 50-90 % and austenite in 10-50 %. The present invention is used in raising the fatigue resistance, stress corrosion resistance, safety and service life of iron and steel material, mechanical structure and equipment. The method and material of the present invention may be used in producing composite material product with compression stress in the surface.

The invention relates to a tube plate of a steam generator and a preparation method thereof. The tube plate of the steam generator supports heat transfer tubes of the steam generator. In particular, it relates to a steam generator tube plate with an anti-corrosion layer and a preparation method thereof. Specifically, the steam generator tube sheet with an anti-corrosion layer includes: a steam generator tube sheet, which supports the heat transfer tube of the vaporizer; and a corrosion-resistant metal, which is plated on the surface of the primary side tube sheet and the surface of the secondary side tube sheet superior. In addition, a method for preparing a steam generator tube sheet with an anti-corrosion layer includes the steps of: coating the surface of the primary side tube sheet and the surface of the secondary side tube sheet with a corrosion-resistant metal; heat-treating the above-mentioned coated tube sheet; and drilling holes, forming heat transfer tube mounting holes on the heat-treated tube sheet; thereby inhibiting the corrosion of the surface of the primary side tube sheet and the surface of the secondary side tube sheet, and preventing the depression and stress corrosion cracking of the heat transfer tube, thereby improving The reliability of the steam generator.

The invention belongs to the field of building structures, and in particular relates to a nuclear facility stainless steel cladding structure and a construction process. The stainless steel cladding structure includes stainless steel cladding, connectors, anchors, ceramic components and concrete, wherein the ceramic components are located between the stainless steel cladding and the concrete for use. The stainless steel cladding and concrete are isolated from each other. After adopting the nuclear facility stainless steel cladding structure of the present invention, the construction process can be simplified, the construction quality level can be improved, the construction period of the nuclear facility stainless steel cladding can be shortened, and the corrosion leakage of the stainless steel cladding can be reduced.