48 results about "Research reactor" patented technology

The invention belongs to the field of environment protection, and in particular relates to an optimization design method of an anaerobic continuous flow agitator bath type biological hydrogen production reactor. The method comprises the following steps: researching reactor internal flow field characteristics of different types of stirring paddles under the conditions of different paddle-bath diameter ratios and different agitating speeds and influence on hydrogen production process, by using a numerical simulation software based on computational fluid mechanics technology; obtaining detailed flow field information, such as velocity field, turbulence energy and dissipation rate thereof, biogas integration rate and shearing rate distribution by adopting a two-phase flow model, and analyzing and comparing different flow field information obtained through simulation according to influence on hydrogen production process by the various flow field information, thereby determining an optimal stirring paddle style and agitating speed combination, and providing an effective method for the optimization design of the agitator bath reactor. The optimization design method of the anaerobic continuous flow agitator bath biological hydrogen production reactor is mature in technology, avoids the blindness of hydraulic design of the reactor using an empirical or semiempirical association method, and has the advantages of being visual in optimization cycle effect and short in optimization cycle, saving cost, and the like.

A special-designed motionless safety reactor study and low temperature heat input reactor facilities in nuclear reactor field. To the side of the reactor installed a large volume water pool whose top is 10-30 meters higher than the reactor top. To the bottom of the water pool a water injection pipe is linked to the reactor inlet, a pipe linked to the reactor outlet is connected to the pool and in a position higher than water level. Outside the plant there is an emergency water tank which is 25-28 meters higher than the main heat exchanger and an air cooler which is 15-18 meters higher than the main heat exchanger. Both the water tank and the air cooler are linked to the recycle pipe between the main heat exchanger and the second heat exchanger.

An instrumented capsule for nuclear fuel irradiation tests in research reactors. The capsule includes an outer shell; a rod tip assembly assembled to a lower end plate of the shell, thus absorbing impact; fuel rod assemblies supported in the shell by means of support tubes and housing therein sintered fuel bodies and in-capsule instruments, such as thermocouples and self-powered neutron detectors (SPND); a protective tube connected to the upper end plate and protecting instrument control cables; and a guide pipe connected to the inclined extension part of a junction tree and guiding the instrument control cables to a control unit provided outside the reactor. The capsule further includes upper and lower end caps provided around the upper and lower ends of the shell; and a lower stopper and an upper stopper fitted over the protective tube to be supported in the upper portion of the irradiation hole of the reactor and to be locked to the locking clamp provided at the upper portion of the chimney of the reactor, respectively. The capsule is used to measure in real time the properties of nuclear fuels irradiated in the reactor during a nuclear fuel irradiation test, thus providing nuclear fuel irradiation test data required for the design of nuclear fuels and the determination of in-pile performance and structural integrity of the nuclear fuels.

The invention discloses a metallic fast reactor fuel element irradiation test device. The device can be used for carrying out neutron irradiation test on metallic fast reactor fuel in a thermal spectrum research reactor and has the functions of neutron spectrum hardening, metallic fuel irradiation temperature control, fuel element heat release power measurement and the like. The device comprises aprotection tube, a cadmium tube, an irradiation test piece, a related gas pipeline, a thermocouple and the like. According to the device, a cadmium metal layer is adopted to absorb thermal neutrons to reduce the thermal neutron flux rate in the device, so that the neutron energy spectrum of a test fuel element is close to that of a fast neutron reactor; and the temperature difference is established between air gaps and a liquid metal layer outside a fuel rod, and the gas components in the air gaps in each irradiation part are independently adjusted, so that the temperature of fuel core bodiesin different irradiation parts are controlled within a required temperature range.

The invention discloses a modularized control rod driving system and a modularized control rod driving method for a horizontal reactor core structure. The invention provides the modularized control rod driving system which is used for a research reactor of the horizontal reactor core structure and has a passive shutdown function; a ball screw-nut structure is used for accurately controlling the position and the movement speed of a control rod assembly and considers experiment functions of the research reactor. The system comprises a horizontal reactor core passive shutdown mechanism, a modularized control rod driving device and a control rod driving system rack, wherein the horizontal reactor core passive shutdown mechanism is used for driving the control rod assembly by a traction counter weight, a traction steel wire and a pulley block to rapidly and horizontally move to be inserted into a horizontal reactor core; the modularized control rod driving device is high in integration degree, simple and convenient to maintain and reliable in performance; the control rod driving system rack is used for determining the position of the modularized control rod driving device in the reactor core and the height of the traction counter weight.

The invention discloses a radiation supervision device for a core container of a high-flux research reactor, which comprises a lifting ring (1), a sample supporting body (2) and a bottom grid plate (3), and is characterized in that the inside of the sample supporting body (2) is provided with a small-sized CT sample, a tensile sample and a temperature monitor. The invention provides the radiation supervision device which is suitable for the core container of the research reactor and has a simple structure, small volume and reactor-following monitoring.

The invention discloses an aerobic composting stink in-situ monitoring device and an operation method thereof. The aerobic composting stink in-situ monitoring device comprises an aerobic composting system, a stink gas in-situ control system, a stink gas online monitoring system and an environmental condition online monitoring system. The device can be used for composting raw materials and auxiliary materials in different kinds, different technological parameters or a stink self-circulation mode is adopted, and efficient low-cost stink in-situ control research in the aerobic composting process is carried out; the composting stink pollution in-situ control effect can be monitored online, and the stink in-situ control operating process can be adjusted according to a monitoring result; compositing samples can be collected, the changing law of the compositing physical and chemical properties is researched, ecological analysis of pile body microorganisms becomes possible, and activity optimization and regulation and control of deodorization relevant microorganisms in a pile body are further achieved.

The invention discloses a method for researching eutectic reaction and high-temperature melting behaviors in a reactor core by using a particle grid mixing method. The method mainly comprises the following steps of: 1, setting initial arrangement of particles for specific problems; 2, performing mesh generation on the initial particle position; 3, solving a mass diffusion equation and an energy equation by using a finite volume method to obtain a mass and temperature solution of a next time layer; 4, solving a momentum equation by using a moving particle semi-implicit method to obtain the positions and speeds of the particles; 5, iteratively solving the steps 2-4 until a result at the required moment is obtained through solving; and 6, outputting a calculation result, and carrying out post-processing to obtain heat distribution, melt distribution and melt migration conditions in the reactor core. The method disclosed by the invention has the advantages of accurately processing the freesurface and simulating the phase change by adopting the moving particle semi-implicit method, has high precision of calculation by adopting the finite volume method, can be used for researching the melting process of the reactor core, and provides an important basis for researching the early melting safety characteristic of the reactor core in serious accidents of the nuclear power plant reactor.

The invention relates to a fuzzy control-based research reactor power automatic adjustment method. The method comprises the following steps: establishing a coordinate system by taking a reactor powermultiple period and a reactor power deviation percentage as horizontal and vertical coordinates respectively, dividing the reactor power multiple period and the reactor power deviation percentage intoa plurality of intervals, and setting different target rod speeds of an adjusting rod for intersection areas of different intervals; calculating a reactor power multiple period and a reactor power deviation percentage according to the current reactor power; and determining an area to which the reactor power multiple period and the reactor power deviation percentage belong in the coordinate systemaccording to a calculation result, and driving an adjusting rod according to a target rod speed set in the area. The method is simple and effective, the operation safety of a research reactor is ensured, and the long-term stable power operation of the research reactor is also ensured.

The invention provides a first criticality method for an anti-neutron trap research reactor. The first criticality method comprises the following steps: performing the subcritical extrapolation by adopting an element method; when the difference between an extrapolated result and an actual fuel assembly loading number of a core is less than 1, switching the element method critical extrapolation torod position method critical extrapolation, under the current core loading, inserting all control rods into the core, then loading a fuel assembly into the core, gradually lifting the control rods step by step, and carrying out the subcritical extrapolation by adopting the rod position of the control rods, gradually transiting the reactor to the critical state step by step, thereby realizing the first criticality. By combining the element method and the rod position method, the problem that a single criticality experiment method does not meet the requirements of the first criticality experiment of the anti-neutron trap research reactor can be solved.

The invention discloses a fission neutron converter. The fission neutron converter comprises an inner sleeve and an outer sleeve coaxially sleeved with the inner sleeve, wherein an upper grid plate and a lower grid plate are respectively mounted at two end parts between the inner sleeve and the outer sleeve, and cross-shaped fuel rods are mounted between the upper grid plate and the lower grid plate. According to the fission neutron converter disclosed by the invention, the bottleneck problem of material irradiation is solved by adequately utilizing technical conditions of the convention research reactor, the twofold effect can be achieved with half effort, and the current problems that on the one hand, the advanced research reactor cannot be built in one day, and on the other hand, a large amount of in-service research reactors are still operated are solved; by utilizing the cross-shaped fuel rods made of high-fission-density fuels, the fluence rate of local fast neutrons of a reactor core is substantially improved under a condition that the physical and thermotechnical safety of an HFETR (High Flux Engineering Test Reactor) is met, so that the technical potentials of the HFETR effectively are developed, and furthermore, the irradiation cost of materials is saved.

The invention provides a series technology for producing phosgene in industry. Different from the traditional phosgene production technology in which one protection reactor is connected to the rear part of one main phosgene reactor, the technology provided by the invention adopts the scheme that one protection reactor is connected to the rear parts of two main reactors; the catalyst loading capacity of each main reactor is halved; the feeding of chlorine gas as a feed gas is divided into two parts, namely, one part of the chlorine gas enters the first main reactor, and the other part of the chlorine gas enters the second main reactor; when catalysts are replaced, a catalyst in the rear reactor is put into the front reactor, and a new catalyst is put into the rear reactor. Through the adoption of the technology, catalysts can be saved; besides, by virtue of the separate feeding of chlorine gas, the total matching ratio of the feed gas and carbon monoxide can be reduced, so that the consumption amount of carbon monoxide is reduced.

The invention discloses a creep irradiation device suitable for stress static loading. The device comprises an upper shell, a lower shell, a threaded power transmission mechanism, a displacement rod,a force measuring part, a movable sample holder I, a movable sample holder II and a fixed sample holder, wherein the threaded power transmission mechanism and the displacement rod are arranged in theupper shell. The force measuring part is arranged in the lower shell. The device is suitable for a stress static loading creep irradiation device, a pipeline does not need to be connected to the outside of a research reactor, and the influence on the stability of the whole device due to the impact of cooling water in the research reactor on the pipeline is avoided. By utilizing the creep irradiation device suitable for stress static loading, simultaneous loading of test environments such as stress, temperature, neutrons and the like of a test sample is realized, meanwhile, the device is compact in structure, a pipeline does not need to be led out of a reactor, and the test reliability and safety are improved.

The invention provides a boron injection system for a second reactor shutdown system of a research reactor. The system comprises a boron liquid tank, an inlet valve, a boron injection pipe, an outlet valve and a boron liquid recovery tank, wherein a liquid outlet of the boron liquid tank is connected with a liquid inlet of the boron injection pipe through a first pipeline; a liquid outlet of the boron injection pipe is connected with a liquid inlet of the boron liquid recovery tank through a second pipeline; the inlet valve is arranged on the first pipeline; the outlet valve is arranged on the second pipeline; the boron injection pipe is arranged around a fuel assembly in the reactor. Compared with the prior art, the boron injection system for the second reactor shutdown system of the research reactor, provided by the invention, is simple in structure, quick in motion, easy to manufacture and mount, and convenient to overhaul; furthermore, the boron solution and a reactor coolant are mutually isolated, so that after the reactor shutdown function is executed, the reactor coolant is not required to be processed, and the normal operation of the reactor can be quickly restored.

The present invention relates to a multi-core fuel rod for research reactor and a manufacturing method thereof and, more particularly, to a multi-core fuel rod for research reactor in which monolithic fuel cores made of uranium-molybdenum alloy are disposed in an aluminum matrix in a multi-core form, and a manufacturing method thereof. The multi-core fuel rod in accordance with the present invention provides a minimized contact surface area between nuclear fuel and aluminum, and reduces the formation of pores and swelling by restraining formation of reaction layer to avoid excessive reaction between the fuel and aluminum. Therefore, improved stability of nuclear fuel can be obtained by minimizing temperature rise as well as achieving high density and thermal conductivity of the fuel.

The invention discloses a nuclear fuel irradiation test method, and belongs to the technical field of research on reactor fuel irradiation. The nuclear fuel irradiation test method comprises the following steps: packaging fuel, wherein filling the fuel into pre-processed and installed and welded fuel assemblies each with a one-way opening, arranging each a pair of the assemblies to be one group between clamping block assemblies side by side; conveying a test section into a reactor, wherein hoisting the test section in the reactor from the reactor top, irradiating the test section to be at a reactor core active section, and connecting a irradiation device section with the reactor top through a flange in sealed connection; carrying out in-reactor test, wherein testing and monitoring relatedparameters after the fuel enters the reactor along with the irradiation device; and discharging the test section out of the reactor, wherein discharging the fuel after a irradiation neutron fluence specified by the test is reached through irradiation. The nuclear fuel irradiation test method is invented aiming at the irradiation requirement of novel nuclear fuel in reactors; the nuclear fuel irradiation test method is more economical and reliable, can reflect the irradiation condition of the nuclear fuel in the reactor in real time, and accurately obtain related irradiation parameters.

The invention discloses a multi-purpose heat pipe reactor prototype and belongs to the technical field of reactor experiment devices. The multi-purpose heat pipe reactor prototype comprises a hexagonal matrix used for simulating a solid reactor core of a heat pipe reactor, a plurality of heating rods inserted into the hexagonal matrix and used for simulating a reactor fuel rod bundle, a Stirling thermoelectric converter used for simulating a dynamic thermoelectric conversion process, a silent thermoelectric converter used for simulating a static thermoelectric conversion process, a plurality of high-temperature alkali metal heat pipes used for simulating a reactor core cooling and energy transmission process, and a pressure-bearing protective shell covering the hexagonal matrix, the heating rods, the high-temperature alkali metal heat pipes, the Stirling thermoelectric converter and the silent thermoelectric converter. According to the multi-purpose heat pipe reactor prototype, the feasibility of the multi-purpose heat pipe reactor nuclear power technology can be verified, and the prototype can be used for studying the response of the operation characteristics of the heat pipes and the operation characteristics of the energy conversion device when the reactor core power changes, and studying the coupling characteristics of the reactor core-heat pipe-energy conversion device.

The invention discloses a miniaturized material irradiation device which comprises an irradiation chamber A. One end of the irradiation chamber A is provided with an upper top cover, the upper top cover is provided with an air inlet hole and an air outlet hole, the air inlet hole and the air outlet hole are both located in the circumferential direction of the upper top cover, an inflation screw plug is arranged at the air inlet hole, and an exhaust screw plug is arranged at the air outlet hole. According to the device, an air inlet pipe and an air outlet pipe which are communicated with the irradiation chamber and the outside of the research reactor are avoided, and the upper top cover is additionally arranged on one side of the irradiation chamber A, the air inlet hole and the air outlethole are formed in the upper top cover, the control plug is adopted to seal the air inlet hole and the air outlet hole, and the screw plug is utilized to seal the upper top cover after the upper top cover is pre-filled with air, so that the longitudinal length of the irradiation device is reduced, and the miniaturization design is realized.

The invention belongs to the field of nuclear radiation, and particularly relates to a method for producing carrier-free <99>Mo by a heavy water reactor. At present, global <99>Mo is mainly carrier-free <99>Mo produced by a research reactor through fission, production of carrier <99>Mo by using the heavy water reactor has been developed and utilized internationally, and an example for producing carrier-free <99>Mo by using the commercial heavy water reactor has not been seen up to now. By means of the method, large-scale stable production and supply of <99>Mo are achieved through the heavy water reactor, and the produced <99>Mo is mainly applied to the field of nuclear medicine diagnosis. The method comprises the following steps: 1, confirming loading conditions; 2, determining a loading position; 3, putting a target piece into a pile; 4, carrying out in-pile irradiation; and 5, discharging the target piece from a reactor core. On the premise that safe operation of the reactor is not affected and fuel loss is reduced as much as possible, the carrier-free <99>Mo is produced through irradiation of the heavy water reactor, and the requirement of the nuclear medicine field for <99>Mo is met.

The invention belongs to the field of nuclear radiation, and particularly relates to a method for producing carrier-free < 99 > Mo by a heavy water reactor. At present, global < 99 > Mo is mainly carrier-free < 99 > Mo produced by a research reactor through fission, production of carrier < 99 > Mo by using a heavy water reactor has been developed and utilized internationally, and an example for producing carrier-free < 99 > Mo by using a commercial heavy water reactor has not been seen up to now. By means of the method, large-scale stable production and supply of < 99 > Mo are achieved through the heavy water reactor, and the produced < 99 > Mo is mainly applied to the field of nuclear medicine diagnosis. The method comprises the following steps: 1, confirming loading conditions; 2, determining a loading position; 3, the target piece is put into a pile; step 4, carrying out in-pile irradiation; and 5, discharging the target from the reactor core. On the premise that safe operation of the reactor is not affected and fuel loss is reduced as much as possible, the carrier-free < 99 > Mo is produced through irradiation of the heavy water reactor, and the requirement of the nuclear medicine field for < 99 > Mo is met.

The invention discloses a resonance calculation method for a research reactor based on an equivalent geometric theory. The resonance calculation method comprises the following steps: establishing an independent isolated lattice cell model for each fuel lattice cell in a target research reactor; solving the fixed source equation to obtain a neutron escape probability and a collision probability; and then the non-rod-shaped geometric fuel is equivalent to one-dimensional rod-shaped equivalent fuel based on escape probability conservation. Then, based on collision probability conservation, the surrounding structural material is converted into a ring shape from a non-ring shape; obtaining a Dankov correction factor of a fuel zone through a neutron current method; and finally, the outer radius of the equivalent moderator area is obtained based on the Dankov correction factor conservation of the fuel area. An equivalent one-dimensional lattice cell model is established through the one-dimensional rod fuel, the annular structure material and the moderator area, and then resonance calculation is conducted to obtain the effective self-shielding section of the problem. Compared with the prior art, the method can greatly improve the resonance calculation efficiency of non-rod and non-plate geometric fuel in numerical reactor simulation calculation on the premise of ensuring the precision.

The invention relates to a preparation method of a nuclear-grade boron carbide pellet. By adopting the preparation method of the nuclear-grade boron carbide pellet provided by the invention, the nuclear-grade boron carbide pellet of which the density is 2.1-2.2 g / cm < 3 > and the 10B enrichment degree is greater than 85% can be obtained through hot pressing sintering of the 10B boron carbide powder, so that the obtained nuclear-grade boron carbide pellet can meet a neutron absorption function required by reactivity control of a research nuclear reactor. Meanwhile, the prepared nuclear-grade boron carbide pellet has high physical and mechanical properties, can ensure that the pellet is complete in structure and stable in performance and operates under various working conditions of high temperature, irradiation, impact, vibration and the like within the service life of a reactor, and is used as a neutron absorber pellet in a domestic research reactor control rod assembly.

The present invention relates to a research reactor power automatic adjustment method based on fuzzy control. The method takes the reactor power doubling period and the reactor power deviation percentage as the abscissa and ordinate respectively, establishes a coordinate system, and takes the reactor power doubling period and the reactor power deviation percentage Divide it into several intervals, set different target rod speeds for the adjustment rods for the intersecting areas of different intervals; calculate the reactor power doubling period and reactor power deviation percentage according to the current reactor power; determine the position in the coordinate system according to the calculation results area, the adjusting rod is driven according to the target rod speed set in this area. The method is simple and effective, which not only ensures the operational safety of the research reactor, but also ensures the long-term stable power operation of the research reactor.

The invention discloses fault tree based initial event frequency evaluation method and device and a computer-readable storage medium. The evaluation method shows a framework and steps for initial event frequency evaluation, and can overcome the difficulty that there is no operation experience and data for support during PSA analysis with a new reactor type or a research reactor. In addition, the method solves a problem that how to make a reasonable evaluation for frequency of occurrence of the initial event by using logistic operation of the fault tree and reasonably defining the reliability models of all basic events under the situation that there is no basic event with known frequency of occurrence. Meanwhile, the evaluation method has good compatibility with common widely-applied commercial software (i.e., Risk Spectrum), so that existing modeling software based on fault tree can be fully utilized. The method is fairly acceptable by nuclear power engineering staff, and brings convenience to engineering implementation.