190 results about "Nuclear grade" patented technology

The invention discloses a preparation method of a preparation for a radioactive gas purifying capability test in a nuclear power plant, which comprises the following steps of: mixing by using a methyl phosphate compound or a dimethylacetal compound, acetonitrile, trimethylchlorosilane and a radioactive iodine source as raw materials, and then reacting for 10-40 minutes at the temperature of 20-50DEG C to obtain the preparation for the radioactive gas purifying capability test in the nuclear power plant. The invention provides the preparation method of the preparation for the radioactive gas purifying capability test in the nuclear power plant, wherein reactants are not virulent, the reaction can be carried out at high speed and high efficiency without high temperature, and other productsexcept methyl iodide have no harm or small toxicity to nuclear-grade impregnated active carbon in an iodine filter.

The invention relates to the field of a manufacture method of a nuclear grade zirconium-niobium alloy cast ingot, adopting the following technical scheme: firstly, forming a pure niobium bar and sponge zirconium into zirconium-niobium master alloy scraps by the steps of pressing an electrode block, firstly smelting, secondly smelting and thirdly smelting, and the like; using the zirconium-niobium master alloy scraps and the sponge zirconium to form the zirconium-niobium alloy cast ingot by the steps of materials mixing, materials distributing, electrodes manufacturing and vacuum self-consumable electric arc smelting; and detecting that the zirconium-niobium alloy cast ingot has no defects such as components segregation and niobium impurity, and the like, to complete the manufacture of the nuclear grade zirconium-niobium alloy cast ingot. The nuclear grade zirconium-niobium alloy cast ingot has even components, has no metallurgical defects such as niobium impurity, and the like, has high raw materials yield, saves the production cost to a certain degree, and voids the waste.

The invention relates to the field of metal smelting, in particular to a method for smelting ultra-pure austenite nuclear grade stainless steel. The method is characterized by comprising the following steps: selecting high-quality raw materials, performing purification treatment, washing a furnace, and forging. The method has the advantages that the produced stainless steel meets the requirementsof RCC-M, 22Cn18-10 and AISI 304L, and also meets the requirement of OCT 04X18H10 standard and the requirement of GB/T1220-2007 00Cr19Ni10 standard, non-metallic inclusions are far lower than the standards of GB00Cr19Ni10, RCC-MZ2Cn18-10, AISI 304L, OCT 04X18H10, and the H and O gas content is low.

The invention relates to a method for producing enriched boric-10 acid from a trifluoride-anisole complex and application thereof. The method comprises the following steps: reacting a boron trifluoride-anisole complex with excessive sodium methoxide methanol solution, operating in an ice bath for 5-25 minutes, carrying out thermostatic water bath, reacting for 40-60 hours while keeping the reaction temperature within the range of 40-60 DEG C, stopping heating, and carrying out centrifugal stratification; fractionating the centrifugated supernatant: heating the mixture, starting to collect the fraction when the temperature rises to 50 DEG C, and stopping collecting the fraction when the temperature rises to 60 DEG C; and carrying out salting-out stratification on the collected fraction, mixing with deionized water, carrying out vacuum filtration on the mixture to obtain a solid, and drying to obtain the boric acid. The enriched boric-10 acid is used in the field of production of 10B-isotope-enriched downstream boric acid products, nuclear-grade boric acids and other enriched boric-10 acids by an anisole chemical exchange fractionation process. The production raw materials are from a closed system; and the invention has the impurity removal link, so the product purity is higher, thereby lowering the difficulty of subsequent boric acid production.

A device for accumulating, isolating, indicating and venting accumulated gas in a fluid system pipe includes a main pipe fitting affixed to a system pipe in which a hole has been drilled. A stand pipe attached to the pipe fitting houses a magnetic float. A magnetic-float level indicator exterior to the pipe and responsive to the magnetic float indicates the magnetic float's level. A valve attached to the stand pipe above the magnetic float allows controlled ventilation of the gas in the standpipe and thus in the piping system. Gas from the system pipe accumulates in the standpipe, removed from the primary fluid flow path of the system pipe. In the standpipe, as the liquid/gas interface drops, the float drops to a pre-determined level, at which point the user vents the gas from the piping system, causing the magnetic float to rise to a level indicating that gas is again at acceptable levels in the piping system.

The invention provides a nuclear-grade dye check penetrant which comprises the following components in parts by weight: 45-70 parts of penetration solvent, 5-22 parts of cosolvent, 7-25 parts of other regents, 5-15 parts of surfactant and 1.5-5 parts of dye. The invention has the following benefits: the penetrant is made from raw materials which are free from three elements of fluorine, chlorine and sulfur; and all the raw materials are purified so that the contents of fluorine, chlorine and sulfur which are brought in the form of impurities are not more than 100ppm. The total content of fluorine and chlorine in the finished product is not more than 100ppm and the sulfur content is not more than 100ppm, and thus, the harm of the penetrant to workpiece is minimized and the penetrant is especially suitable for equipment detection in nuclear industry. In addition, the penetrant has the advantages of high sensitivity and high flashing point, can detect a flaw of the opening width of 0.4 mu m and can be washed with both organic solvents and deionized water.

The invention relates to nuclear grade ultrasonic inspection couplant, which is characterized by comprising the following components in parts by weight: 50-90 parts of deionized water, 0.3-5 parts of gel matrix, 0.3-8 parts of pH value modifier, 5-30 parts of moisturizing agent, 1-10 parts of anti-rusting agent and 0.1-11 parts of antibacterial preservative. The invention also provides a preparation method of the nuclear grade ultrasonic inspection couplant. The nuclear grade ultrasonic inspection couplant has low content of fluorine, chlorine and sulfur and has the advantages of having no toxicity or harm, pure and transparent appearance, no bubble, adjustable viscosity, good anti-rust capability, stable storage, good sound transparent performance and the like, and is easy to clear.

The application discloses a T3 test loop of a reactor protection system in a nuclear power station and an optimization method thereof. In the optimization method, the T3 test loop employs parallel resistance to check current loop, and comprises a plurality of test switches, wherein at least one test switch is corresponding to a multiple paths of devices, and the at least one test switch has at least two pairs of contact in parallel connection to form parallel contacts. The application has the beneficial effects that: at least two pairs of contact in the test switches are in parallel output to drive the test loop and reduce the risk of maloperation caused by high failure rate of contacts; in one embodiment, nuclear grade relay nodes are connected into the test loop to further improve the reliability of the equipment; in another embodiment, self-locking of test signal is added for interlocking test signal in the case of switch contact failure to reduce the maloperation; and in another embodiment, additional alarm signal sent to a main control enables the normal operation and monitoring on the loop, in order to reduce the risk of refusal-operation caused by scheme optimization.

The invention discloses a method of adopting a strain sensor to monitor the fatigue of nuclear grade pipelines. At first, a strain sensor (a strain sheet or a strain meter) is used to monitor the strain of the external surface of a nuclear grade pipeline; then according to the external surface strain of the nuclear grade pipeline, the properties of the material of the nuclear grade pipeline, and information of an existed monitoring device of the nuclear power plant, the strain of the internal wall of the nuclear grade pipeline can be determined; then alternation amplitude in strain circulations and corresponding circulation times can be determined; according to a preset epsilon-N curve, the fatigue use coefficient is calculated; then according to the pipeline strain history, the reasonable environmental influence coefficient (Fen) is determined, the fatigue use coefficient is corrected; finally, the fatigue use coefficient is output on the display terminal of the master-control room, and an accumulated fatigue use coefficient curve, which changes along the time, can be outputted and can be used to evaluate the change trend of accumulated fatigue use coefficient of a monitored position.

The invention belongs to the technical field of zirconium alloy preparation and particularly relates to a preparation method of a nuclear-grade zirconium alloy cast ingot. The preparation method comprises the following steps that (1) the alloy component and the total mass of the finished cast ingot are determined; (2) an intermediate alloy and an alloy cladding are prepared; (3) material mixing isconducted; (4) electrode pressing is conducted; (5) N electrodes are welded into one electrode; (6) vacuum self-consuming arc melting is conducted; and (7) surface machining and dead head cutting areconducted. According to a technology suitable for industrial scale production and used for preparation of the Zr-Sn-Nb-Fe/Cr zirconium alloy cast ingot, the technological economical efficiency can beobviously improved, meanwhile component evenness control over the cast ingot is superior to that of a traditional technology, and the preparation method can be used for industrial scale production ofthe nuclear-grade zirconium alloy cast ingot accordingly.

A nuclear class cable sheath material comprises the following components by weight: 30-50 parts of ethylene-vinyl acetate copolymer; 50-70 parts of ethylene propylene rubber; 5-10 parts of silicone rubber; 70-90 parts of magnesium hydroxide; 5-10 parts of boron nitride and 4-12 parts of an antioxidant. After the nuclear class cable sheath material is used for 60 years at 90 DEG C, and a radiation dose of 2400KGy is applied, a sheath layer may not peel, and the nuclear class cable sheath material is high in flame retardant efficiency, low in adding amount and uniform in dispersion, may not agglomerate, and meets material mechanical properties, meanwhile can get through a single vertical combustion (B-type bundling) test.

The invention belongs to the nuclear process control system reliability detection field and relates to a pressurized water reactor nuclear power plant nuclear-grade digital instrument control system logical function test device. The pressurized water reactor nuclear power plant nuclear-grade digital instrument control system logical function test device includes a host computer, a switch, a printer and a network connection medium. According to the device, a network communication mode-based device structure is adopted. Before a logic function test is performed, a KCS is connected into the special test device, after the test is completed, the communication connection of the special test device and the KCS is cut off; a test network and a network which is adopted under a normal operation condition are relatively independent; and an independent communication card is adopted, so that the safety of the KCS can be ensured.

The present invention discloses a method of calculating strength of a nuclear grade pipeline support. The method comprises a first step of simplifying a pipe clamp mechanical model according to a pipe clamp structure size; a second step of deducing a section internal force and a bottom support reaction force of the pipe clamp under the action of external load according to a mechanical theory of a plane curved bar material; a third step of calculating a stress of each section of the pipe clamp according to a stress formula of a plane curved bar; a fourth step of calculating a bottom weld seam stress and a pipe clamp weld seam stress according to the load of the weld seam; a fifth step of calculating a bolt stress according to external load of the pipe clamp; and a sixth step of checking and assessing the pipe clamp strength and weld seam strength by using an RCC-M specification, and assessing the stress of the bolt, wherein if the assessment fails, the structure size is readjusted, and calculation and checking are performed again. The method of calculating strength of the nuclear grade pipeline support offers a complete strength calculation process, the whole process is expressed with a theory formula, so that the method is convenient for fast calculation and batch application, can be used for designing a new pipe clamp structure, and can determine a permitted load.

Relating to the technical field of nuclear power, the invention provides a nuclear reactor core neutron and temperature detection device, and solves the technical problems of reducing the sensitivitydeviation and improving the service life. The device comprises a plurality of reactor core detection components installed in a pressure vessel, the reactor core detection components are evenly laid along the radial direction of the pressure vessel, and each reactor core detection component includes at least one armored nuclear grade thermocouple thermometer for monitoring the reactor core fuel assembly outlet coolant temperature and multiple vanadium self-powered neutron detectors for monitoring the nuclear reactor core neutron fluence rate, also the vanadium emitters of the vanadium self-powered neutron detectors have an equal length, and the vanadium self-powered neutron detectors are laid from the bottom up in order along the axial direction of the pressure vessel. The device provided by the invention is suitable for a pressurized water reactor nuclear power station reactor core instrument system.

The invention belongs to the technical field of hydraulic shock-resistant supporting devices, and relates to a nuclear-grade large equipment damper. The damper solves the technical problems that the prior art is unreasonable in design and the like. A piston rod passes through a damping cylinder; a piston positioned in the damping cylinder is arranged on the piston rod; valve seats are respectively arranged at two ends of the damping cylinder; the two valve seats are respectively communicated with oil through holes at two ends of the damping cylinder; the two oil through holes are respectively communicated with a left chamber and a right chamber of the damping cylinder; valve cores, the outer ends of which are conical, are respectively arranged in the valve seats; throttling structures distributed symmetrically are arranged on the valve cores; springs for pressing the valve cores in the oil through holes are arranged between the valve cores and the valve seats; a connecting pipe is arranged between the two valve seats; an auxiliary cylinder is arranged at the outer side of the damping cylinder; and the auxiliary cylinder is connected with one of the valve seats. The damper has the advantages of reasonability in design, simple structure, high mechanical strength, stable working performance and capability of working normally for a long term even if the working environment is severe. The damper has good radiation resistance, and is easy in maintenance and convenient for changing a hydraulic medium.

The invention discloses an inductive coupling plasma atomic emission spectrometry for determining elements in a nuclear-grade zirconium alloy. The inductive coupling plasma atomic emission spectrometry is used for determining 4 types of macroelements and 13 types of trace elements in the nuclear-grade zirconium alloy. The macroelements including tin, niobium, iron and chromium, and the trace elements including aluminum, cobalt, copper, molybdenum, magnesium, manganese, nickel, lead, silicon, tantalum, titanium, vanadium and tungsten in the nuclear-grade zirconium alloy are successfully determined by carrying out base body match on high-purity sponge zirconium and main alloy elements and selecting a suitable spectral line as an analysis line of detected elements. The determination of tin, iron, chromium and nickel in a 360b zirconium alloy in NIST is carried out and determination results are the same as standard values of a standard substance certificate. The nuclear-grade zirconium alloy is subjected to a marking recovery test; and a result shows that the recovery rates of niobium, cobalt, copper, molybdenum, magnesium, manganese, silicon, titanium, vanadium and tungsten are 92%-108% except that the recovery rate of tantalum is low and the recovery rates of aluminum and lead are high. According to the method, the determination result is stable; and the relative standard deviations (RSD) of the determination results of 3 days are lower than 6% and can meet the requirements of Westinghouse certification standards (RSD is less than 10%).

This present invention provides a two-step process for producing nuclear grade, active uranium dioxide (UO₂)powder in which the first step comprises reacting uranium hexafluoride (UF₆) with steam in, for example, an integrated dry route (IDR)-type kiln or a flame reactor to yield uranyl fluoride (UO₂F₂); and the second step comprises removing fluoride and reducing UO₂F₂ to uranium dioxide (UO₂) in a second kiln under a steam/hydrogen atmosphere. The two-step process tightly controls the exothermicity of the reaction, which allows for a very tight temperature control which controls the growth of the particles and results in UO₂ powder that is active.

The invention discloses a nuclear-grade ion-exchange resin-based composite material and a preparation method thereof. The composite material is prepared by the steps: firstly reducing a platinum precursor into polymer-packaged nano platinum particles by using a reducing agent in existence of a polymer packaging agent; adding a silicon source and an aluminum source and then carrying out hydrothermal reaction employing a polymer as a structure-directing agent to obtain a platinum-silicon aluminum molecular sieve core-shell structure compound, wherein the inner core is nano platinum particles and the shell layer is a silicon-aluminum molecular sieve; and finally assembling the platinum-silicon aluminum molecular sieve core-shell structure compound to nuclear-grade ion-exchange resin to obtain the nuclear-grade ion-exchange resin-based composite material. The composite material has relatively high ion exchange property, relatively high catalytic removal efficiency, relatively high environment durability and extremely high recycling property, and has a good application prospect in the field of nuclear industry circulating water treatment.

The invention relates to nuclear-grade sponge zirconium, in particular to a method for preparing nuclear-grade sponge zirconium and nuclear-grade sponge hafnium. Chlorine generated by hydrochloric acid electrolysis is used for producing zirconium tetrachloride and zirconium oxychloride by a zircon sand boiling chlorination method; zirconium and hafnium are seperated to obtain zirconium oxide and hafnium oxide through a TBP-Hcl-HNO3 extraction method; and then, the chlorination and the purification are performed to obtain the nuclear-grade sponge zirconium and the nuclear-grade sponge hafnium. The method radically solves the environmental protecting environment difficultly treated in the zirconium oxychloride industry and the nuclear-grade sponge zirconium industry, and realizes complete industry chain circulation for combined production of the zirconium oxychloride industry, the gas-phase method silicon dioxide industry, the nuclear-grade sponge zirconium industry and the hydrochloric acid electrolysis industry.

The invention relates to a slag material for electroslag remelting of nuclear-grade stainless steel and a method for electroslag remelting by the slag material. The slag material is prepared from thefollowing components in percentage by weight: 40 to 70% of CaF2, 1 to 10% of MgF2, 10 to 20% of Al2O3, 10 to 20% of CaO, 1 to 10% of MgO, and 0 to 3% of BaO. The preparation method of an electroslag ingot mainly comprises the following steps of preparation of an electrode rod, preparation of the slag material, and electroslag remelting. The prepared electroslag ingot has the advantages that the content of harmful elements of Co, B, Pb, Sn, H, P, S and the like is low, the surface quality is high, the components are uniform, the segregation is little, the microstructures are fine and dense, thesum of nonmetallic inclusions is lower than or equal to Grade 2.0, and the metallurgical quality requirement of the nuclear-grade stainless steel is met.

The invention discloses a modification method for a filter material used in a filter core of a nuclear grade water filter. To improve leaching resistance of a filter material in a filter core of a water filter used in a nuclear power station, especially the problem of leaching of SiO2, the invention provides a following scheme: a commercially available glass fiber filter material and a commercially available paper-based filter material are treated with a polyacrylamide polymer aqueous solution, then the filter materials are dried in a baking oven so as to allow polymers on the filter materials to undergo crosslinking, and finally, the filter materials are washed and dried for subsequent usage. With the modification technology for the filter material in the invention, leaching resistance of a plurality of trace elements (especially SiO2) in the filter material can be improved, and filtration efficiency of the filter material on a subsize particle can be enhanced.

The invention discloses a method for measuring the content of a hydrogen element in a nuclear-grade zirconium material. The method comprises the following steps: 1, cutting the nuclear-grade zirconium material into 0.1-0.3g of a solid test sample; 2, firstly, pickling the solid test sample for 5-30s so as to remove an oxygen enrichment layer on the surface of the solid test sample and hydrogen and other gas elements adsorbed in the machining process, subsequently washing the pickled test sample for 10-50 seconds by using acetone, subsequently putting the test sample into an oven to be dried at 50-60 DEG C, and removing a contaminated layer on the surface of the test sample so as to ensure the cleanness of the test sample; 3, putting the test sample pretreated in the step 2 into a hydrogen measuring instrument so as to measure the content of the hydrogen in the solid test sample by using the hydrogen measuring instrument. By adopting the method, the defects of the conventional measuring method are overcome, and requirements of practical detection on the nuclear-grade zirconium material are met.

The invention relates to a pressure resistance valve for a nuclear grade air passage, which is used in a ventilation system of a nuclear power station and comprises a valve body, a valve seat, blades, an automatic blade reset mechanism of the blades and an emergency protection mechanism, wherein the blades and the automatic blade reset mechanism of the blades comprise more than three blades in vertical arrangement, and rotating shafts, a rocker arm and a pair of symmetrical left and right reset sub mechanisms corresponding to each blade, the valve seat comprises an upper sealing valve seat, a lower sealing valve seat and a side sealing seat, the emergency protection mechanism comprises an electric device and an electric fitting rocker bar, the electric device is fixed on a steel plate welded on the valve body through screw bolts, the electric rocker bar is arranged on an output shaft of the electric device, and one end of the electric fitting rocker bar presses the upper end of a connecting rod when the electric fitting rocker bar leftwards rotates for 90 degrees. The structure is simple, the action is flexible, the emergency protection capability is realized, and the back ventilation system can be ensured not to be damaged when pipe bursting accidents occur in a cabin chamber of the nuclear power station, so the ventilation requirement of the nuclear power station is met.

The application of the invention discloses a nuclear safety grade 316L stainless steel gauge pipe. A round pipe blank is adopted as a raw material. The manufacture method comprises the steps of a cold rolling procedure, a solid solution thermal treatment procedure and storage; the cold rolling procedure comprises a cogging cold rolling step, an intermediate-product cold rolling step and a finished-product cold rolling step; the relative deformation of the cogging cold rolling step is 70%-75%; the relative deformation of the intermediate-product cold rolling step is 65%-70%; the relative deformation of the finished-product cold rolling step is 60%-65%; and the treatment temperature of the solid solution thermal treatment procedure is 1060 DEG C-1080 DEG C. The dimension precision, the surface quality and the comprehensive performance of products reach a higher level; the long-term safe service requirement in a nuclear grade environment is met; and the production efficiency and the qualification rate are both at a high level.