219 results about "Radioactive gas" patented technology

A gas recovery and reuse/recycle method is disclosed which can readily and economically recover valuable and/or environmentally hazardous gases from a manufacturing or chemical process and then return the gas to the process for reuse, and repeat this many times without significant contamination or degradation of the gas or the produced products. All gas transport, compression and storage equipment is designed and maintained so that it is non-contaminating to the process gas. Commonly the process gas will be a Group VIII gas, preferably He, Ne, Kr or Xe, or a gas which comprises a hazard to the ambient environment or beings therein, such as a carbon oxide gas, a halocarbon gas, an acid-precursor gas, a biologically hazardous gas, or a radioactive gas.

The invention discloses a major accident mitigation system of a nuclear power station on the basis of the nano fluid characteristic, belonging to the technical field of nuclear power station equipment and safety. The system is composed of a superconductive tube, a heat exchanger, a radioactive grain removing device, a nano fluid preparation chamber, an ultrasonic oscillator, an air cooling tower and the like. The system can quickly adsorb energy in a containment under the major accident, hyperpressure in the containment can be alleviated by releasing and removing radioactive air and radioactive aerosol particles, nano particles can be obtained, and the efficient heat exchange medium nano fluid can be prepared to achieve the goal of alleviating the major accident of the reactor and protecting the integrity of the containment structure. A safety control function is performed and finished when major accidents happen. The system has the advantages of good backup safety, simple flow, stable performance and high reliability, is convenient to implement and simple to control, and treats wastes with wastes.

The invention provides a detection system for measuring beta-ray generated by a radioactive gas. The detection system comprises a detection device and a detection circuit which are connected, wherein the detection device comprises a main detector, and an anticoincidence detector wrapped at the periphery of the main detector; the main detector comprises a box body, at least one first plastic scintillator, a first glass light guide arranged between one first plastic scintillator and one inner wall of the box body, and a first photomultiplier connected between one first glass light guide and the detection circuit; the anticoincidence detector comprises a second plastic scintillator with a hollow middle for accommodating the main detector, and at least one second photomultiplier connected to the second plastic scintillator through a second glass light guide and connected with the detection circuit. According to the detection system disclosed by the invention, the influence of gamma-ray on the measurement result of beta-ray is greatly reduced in a manner of combining the main detector with the anticoincidence detector, thus the detection system is good in convenience on the premise of ensuring the measurement lower limit of the detection system.

The invention discloses a PIG sampling and monitoring system and a method, comprising a sampling loop unit and a radioactivity measurement analysis control unit. The sampling loop unit comprises an aerosol accumulation sampling subunit and a radioactive gas accumulation sampling subunit; the aerosol accumulation sampling subunit comprises a filter; the radioactive gas accumulation sampling subunit comprises a gas accumulation sampling device and a molecular filtering film; the radioactive measurement analysis control unit comprises an HPGe detector, an energy spectrum analysis meter and a computer. The HPGe detector performs measurement on the radioactivity of the above subset units at the same time; the energy spectrum analysis meter processes the measurement information and gives out the radioactivity energy spectrum information; the radioactivity energy spectrum information and the flow control information of the sampling loop unit are transmitted to the computer for processing to obtain activity concentration of all nuclides in the airborne fluid. The invention realizes the accumulated sampling measurement on the PIG in the airborne fluid, improves the accuracy of radioactivity nuclide measurement in PIG, and has better practicability.

The invention relates to an on-line monitoring system for radioactive gas and radioactive aerosol. The on-line monitoring system comprises a sampling measuring device, an inert gas cavity and a gas pump; the sampling measuring device is provided with a circulation channel, filter paper used for sedimenting the radioactive aerosol and a radioactive aerosol detector used for detecting the radioactive aerosol; the filter paper is arranged on a path of the circulation channel, and the radioactive aerosol detector is arranged over the filter paper. The inert gas cavity is communicated with the circulation channel of the sampling measuring device through a connecting gas tube, and a radioactive gas detector used for detecting the radioactive gas is arranged in the inert gas cavity. The gas pump is communicated with the inert gas cavity. The invention further relates to an on-line monitoring method. By means of the on-line monitoring system and method, measurement on alpha rays and beta rays of radioactive aerosol particles is carried out through the radioactive aerosol detector, measurement on beta rays of the radioactive gas is carried out through the radioactive gas detector, and therefore real-time monitoring on the radioactive gas and the radioactive aerosol is completed.

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 belongs to the nuclear radiation monitoring technical field and provides an airborne radioactive monitor. The airborne radioactive monitor includes an aerosol iodine detection system, a radioactive gas detection system and a sampling system; a sampling gas inlet end is connected with the aerosol iodine detection system and the sampling system; the outlet of the aerosol iodine detection system is divided into two paths, wherein one pipeline is connected with the radioactive gas detection system, and the other pipeline is connected with a sampling gas outlet end and is provided with a manual shut-off valve; and the outlet of the sampling system is connected with the sampling gas outlet end. The airborne radioactive monitor of the invention has the advantages of high sensitivity, low detection limit and short response time, and can better realize fast and effective measurement of airborne radioactive material.

The invention belongs to an apparatus for nuclear reactor spent fuel element damage detection, specifically an underwater sipping apparatus for the spent fuel element damage detection. The apparatus comprises a sipping container, a quick coupling, air inlet pipes, exhaust pipes, air inlet valves, air release valves, drainage pipes and the like. A main material of the apparatus is stainless steel.According to the apparatus provided by the present invention, batch radioactive gas samples required by the spent fuel element damage detection can be collected in a short time so as to provide high detection efficiency.

The invention discloses a gamma radioactive gas body source absolute detection efficiency calibration device and method. The device comprises a standard face source, a solid small face source, a solid big face source, a gas big face source, a gas body source and a detector. When the height H of the inner cavity of the gas big face source is smaller than a certain specific value Y, the gamma ray absolute detection efficiency deviation of the solid big face source and the gas big face source is smaller than 0.5%, and the detection efficiency of the solid big face source and the detection efficiency of the gas big face source are equivalent within 0.5%. The value Y is taken to prepare the gas big face source, the equivalent detection efficiency of the gas big face source is obtained through the solid big face source with the detection efficiency easy to measure, and the detection efficiency of the gas body source can be obtained through the ratio of the size of the gas big face source to the size of the gas body source. The absolute detection efficiency calibration of any gas body source with an unchanged shape is achieved by ingeniously utilizing quality transition, efficiency equivalency, size transition and other methods. The device and method are convenient and easy to achieve.

A low-temp. nuclear reactor using depleted fuel for nuclear power station is disclosed. The reactor core is in a water tank. A depleled fuel package without any processing is used for the reactor core. The sealing cover of the water tank for the said core has at least one air-tight shielding and is filled with gas under a certain pressure. A water channel is arranged between two water tanks respectively for reactor core and depleted fuel. Its advantages include raised utilization value of uranium resource, high environemnt protection effect, low cost, and high safety.

A shield has a detector installing hole having a central axis orthogonal to a central axis of a detection tube. A radiation sensor of a columnar scintillation detector to which a low concentration side measurement range within all required measurement ranges is allocated is arranged inside the detector installing hole. An ionization chamber to which a high concentration side measurement range is allocated is arranged side by side in parallel with the detection tube. The measurement range of the columnar scintillation detector and the measurement range of the ionization chamber are adjusted so as to be overlapped with each other.

The invention discloses a method for infield transporting and unloading of spent fuel in a nuclear power plant. The method comprises the following steps: S1, transporting a spent fuel loaded transportation container to an unloading area in a horizontal state; S2, placing the transportation container into a cleaning pool of the unloading area in a vertical state; S3, disassembling an outer cover of the transportation container, and arranging a shield plate on an inner cover of the transportation container; S4, opening covers on a drainage hole and an exhaust hole on the inner cover, connecting a radioactive detection device, detecting if the transportation container has radioactive gases or not, if so, stopping unloading, and if not, removing the radioactive detection device, and executing the step S5; S5, filling water into the transportation container and carrying out cooling; S6, disassembling the inner cover of the transportation container; and S7, taking the spent fuel out of the transportation container. The method provided by the invention realizes infield transporting and unloading of the spent fuel in the nuclear power plant, and guarantees empty framework to receive new fuel in the nuclear power plant.

The invention is applied to the technical field of nuclear power and provides a nuclear power station radiation distribution prediction system. The nuclear power station radiation distribution prediction system comprises multiple sensor nodes, a gateway, a server base station and remote monitoring processing centers. The multiple sensor nodes gather acquired data to the server base station through the gateway, the server base station transmits the data to the remote monitoring processing centers, and the remote monitoring processing centers predict radiation distribution of a nuclear power station according to the data, wherein the prediction includes windless condition prediction on a nuclear radiation concentration C of time t with a space coordinate position (x,y,z) on the basis of an expression shown in the descriptions, P0 is a concentration of a radioactive gas leaked at a space coordinate origin when t=0, k is a diffusion coefficient, and m is the mass of the radioactive gas in a volume V. According to the invention, the prediction accuracy is effectively improved.

The invention discloses a containment pressure relieving system of an underground nuclear power station in serious accidents. The system comprises a containment, a pressure relieving pipeline and a pressure relieving chamber, wherein the containment is arranged in a reactor building; the pressure relieving pipeline is used for communicating the containment, the reactor building, the pressure relieving chamber and an outside chimney respectively by branches; each pressure relieving branch is provided with a related valve and a radioactive gas treatment device. After the containment pressure relieving system is adopted, automatic pressure release can be automatically realized by different pressure relieving ways according to the pressure rise in the containment, and radioactive air can be filtered and discharged in the period of the whole accident, so that great convenience is brought to the rescue and spot clean after the accident.

The invention relates to a beta-ray self-absorption correction method for radioactive gaseous nuclide. For beta-gamma coincidence measurement of radioactive gas activity, a self-absorption problem inevitably exists in the beta rays, and if the self-absorption is not corrected, the accuracy of the beta-gamma coincidence measurement can be influenced. Through the use of the self-absorption correction method disclosed by the invention, a self-absorption curve of the beta rays in target gases in different contents is researched through experiments, so that self-absorption factors of the beta rays in any gas components are obtained; the self-absorption correction method disclosed by the invention solves the difficulty of the beta-ray self-absorption correction during the beta-gamma coincidence measurement of the radioactive gas activity.

The invention discloses a method for improving the adsorption quantity of radioactive gases such as radon gas and the like, and is characterized in that adsorption devices (two or more classes) constitute a series circuit, when the radioactive gas adsorbed by the adsorption device (at the previous class) in the circuit is transferred to the adsorption device (at the next class), under the action of a pump and a throttling valve, the pressure of the adsorption device (at the previous class) is lower than an atmospheric pressure, the pressure of the adsorption device (at the next class) is higher than the atmospheric pressure, the pressure ratio of the adsorption device (at the next class) to the adsorption device (at the previous class) is n, and the adsorption coefficient thereof is proportional to the pressure ratio, so that the radioactive gas adsorption quantity of the adsorption device (at the next class) is n times larger than that of the adsorption device (at the next class). Therefore, the aerosols in the air can be filtered, and the radioactive gas in the air can be adsorbed. Because the hydrogen gas is difficult to be adsorbed by adsorbing materials at normal temperature, the high-concentration hydrogen gas from nuclear power plants can be quickly discharged with the qualified air.

The invention provides a full-closed type double-ball type interlocking separation valve which is valve equipment capable of effectively separating toxic, harmful, combustible, explosive and radioactive gas. The full-closed type double-ball type interlocking separation valve is suitable for industries of metallurgy, electric power, chemical industry, municipal administration, environmental friendliness, fuel gas and the like. In order to overcome the problems of a floating and hanging straight-line transversely-inserted type horizontal movable opening/closing switching mechanism with a valve body which has a framework box type structure and a valve plate which has a flat plate bearing structure of the existing separation valve technology that the weight is great, the volume is large, the occupied pipeline space position is large, a process pipeline is not easy to optimize, the bearing pressure is low, the rigidity of a main body is small, the strength is low, the gravity center of the valve deviates, the stress is not uniform, the walking is not stable, the clamping is easy to generate, the route is long, the opening and the closing are slow and the limitation flexibility has defects and disadvantages. The double-ball type interlocking separation valve is composed of a double-ball valve plate overturning mechanism, an interlocking pressure-balancing mechanism, a spherical valve body, a replacing device, a dirt discharging device, a cooling system, a spherical valve plate clamping mechanism, a blowing device and a diffusing device. The full-closed type double-ball type interlocking separation valve has the advantages of high performance, high parameters, multiple functions, low cost, reliable operability and the like.

The invention discloses a nuclear power station radioactive waste gas treatment system. The nuclear power station radioactive waste gas treatment system comprises a gas cooler, a gas-water separator, a compressor, a delay treatment unit and a radioactive monitoring and control device, wherein the gas cooler is connected with an upstream system pipeline through a gas inlet pipeline, the gas-water separator, the compressor, the delay treatment unit and the radioactive monitoring and control device are sequentially and serially connected to the downstream of the gas cooler through a pipeline. The compressor is arranged in front of the delay treatment unit, so that the running pressure of the system is effectively increased, the adsorption coefficient of the adsorbent on the gas can be greatly increased, the delay attenuation time of the radioactive gas can be prolonged, the radioactive waste gas treatment capacity of the system can be multiplied, the radioactive waste gas treatment efficiency is improved, the application range of the system is enlarged, the treatment effect of the system is improved, and the popularization and application prospect is promising.

The invention discloses a method for calculating activity of radioactive gas diffused to the atmospheric environment under the accident condition of an underground nuclear power station. The method comprises steps as follows: establishing a radioactive gas diffusion database of the underground nuclear power station through research of existing literature materials; consulting data to acquire source item data of each radionuclide escaped from a reactor building cavern of an underground nuclear power station containment; determining all diffusion paths of the radioactive gas diffused from the reactor building cavern to the atmospheric environment and calculating the concentration of each type of radioactive gas passing each diffusion path; sequentially evaluating the diffusion behavior of the radioactive gas of each path in each diffusion link to obtain the radioactive activity of each radionuclide diffused to atmosphere. According to the method, one set of complete and scientific radioactive gas diffusion evaluation system of the underground nuclear power station is established, quite high universality is realized, calculation software is formed easily through programming, user's use and secondary development are facilitated, and data support is provided for site selection, formulation of a contingency plan and the like of the underground nuclear power station.

The invention is suitable for the field of nuclear power plants, and provides a boron recovery system of a nuclear power plant, comprising a preposing temporary storage box, a filter desalination device, a degasser and an intermediate temporary storage box which are sequentially connected. A reactor coolant exhausted by a nuclear island of the nuclear power plant is collected by the preposing temporary storage box, is treated by the filter desalination device, and is stored in the intermediate temporary storage box so as to carry out boron-water separation treatment when hydrogen and radioactive gases in the reactor coolant are deprived by the degasser. The recovery system also comprises a bypass pipeline device, one end of the bypass pipeline device is connected with an inlet pipe of the degasser, the other end of the bypass pipeline device is connected with a water inlet pipe of the intermediate temporary storage box, and the reactor coolant in the preposing temporary storage box is directly exhausted to the intermediate temporary storage box. The boron recovery system of the nuclear power plant directly exhausts the reactor coolant going through the preposing temporary storage box and the filter desalination device to the intermediate temporary storage box through the bypass pipeline device, reduces the running time of the degasser when a machine set is overhauled, and saves the running cost at the same time.