62results about How to "Ensure Radiation Safety" patented technology

The invention aims to disclose an ambient neutron dose equivalent meter. The ambient neutron dose equivalent meter is characterized by comprising a neutron probe, a neutron energy response adjustor and a control system, wherein a neutron detection element is arranged in the neutron probe, the neutron detection element is europium-doped lithium iodide scintillation crystal, the signal output end of the neutron probe is connected with the signal input end of the control system, and the neutron probe is placed at the position of the geometric center of the neutron energy response adjustor. Because the photomultiplier with high sensitivity is used as a photoelectric converter, the meter has light weight, portability, sensitive neutron response and good angle response, can accurately measure the ambient neutron dose equivalent level, has sound and light alarm function, is used for ambient neutron radiation monitoring of large nuclear facilities such as reactors, high-energy particle accelerators and the like, ensures neutron radiation safety of the ambient and the public, and fulfills the purpose of the invention.

The invention discloses a vehicle-mounted mobile radiation safety inspection system which comprises a radiation source, a scanning arm and a detector. The system further comprises a geomagnetic switch which is arranged at an entrance from which a detected object comes into a scanning channel, a first light curtain switch which is arranged at the entrance of the scanning channel and is used for detecting whether the detected object enters and leaves the scanning channel, and a main beam detection switch which is arranged at a main beam position in the downstream direction of the scanning channel and is used for detecting whether the detected object reaches the main beam position. When the geomagnetic switch, the first light curtain switch and the main beam detection switch are triggered in turn, the radiation source is started to prepare a beam outgoing operation. In addition, the invention further relates to a control method of the vehicle-mounted mobile radiation safety inspection system.

The invention discloses a boron neutron capture therapy dose measuring method. The boron neutron capture therapy dose measuring method comprises: according to the difference of Cherenkov light intensity distribution per unit volume before injection of medicament containing boron and during the boron neutron capture therapy process, determining concentration distribution of the medicament containing boron in real time; and at the same time, according to the Cherenkov light intensity distribution per unit volume during the boron neutron capture therapy process, obtaining the Gamma total dose distribution, and finally obtaining the boron neutron capture therapy dose distribution according to the proportional relation between various doses which are generated because of reaction between thermal/epithermal neutron and different components. The boron neutron capture therapy dose measuring method has the advantages of being able to judge concentration distribution of the medicament containing boron, being high in real time online property and being simple in measurement, and can effectively guarantee the curative effect of boron neutron capture therapy and radiation safety for a patient.

The invention provides a sample preparation system of carbon-14 in a solid sample and belongs to the field of a radioactive sample preparation system. The sample preparation system is of an overall integrated cabinet type structure. All parts are arranged in an integrated cabinet. Compressed gas in the sample preparation system is controlled by a gas intake micrometering valve, an automatic valve, a pressure sensor and a mass flow meter and enters a buffering device and a sample tube. Pyrogenic decomposition is carried out on a sample in a decomposition bed in the aerobic atmosphere. A catalyzing bed is used for catalyzing the oxidized gas with heat and the gas is cooled by a cooling device. Tritium adsorption and water processing are carried out through a silica gel adsorption bed. The gas enters absorbers which are connected in series in three levels to be adsorbed through flow control. The re-processed gas is exhausted by a vacuum pump. According to the sample preparation system of the carbon-14 in the solid sample, a computer interface can be directly used for controlling a sample preparation process. In an automatic operation process, an operator can conveniently intervene in the whole operation process after authority of the operator is approved. The sample preparation system of the carbon-14 in the solid sample is economical, practical, capable of saving the energy and protecting the environment, high in practicability, capable of conveniently replacing parts, easy to clean and capable of achieving small scale and automation.

The invention discloses a nuclear power plant point source radiation source intensity back-stepping method and a point source radiation source intensity back-stepping system. According to the method, dosage rates of multiple positions are acquired through detectors, a normalization radiation source intensity is dispersed in space, a ray tracking method is utilized to calculate an optical distance, equation set coefficient calculation is developed in combination with the material and accumulation factor information, and source intensity is back-stepped; dosage rate calculation is carried out for the positions of the detectors, linear regression analysis is carried out through measurement values and calculation values, key parameters such as standard deviation, gradient and intercept are acquired through calculation, and a quality factor is calculated to measure an acceptable degree of each calculation result; a weight iteration method is further proposed, errors caused by the detectors with relatively large uncertainties are reduced, the previous steps are repeated through utilizing an iteration mode till the quality factor reaches a preset constant value, and the expected radiation source intensity information is acquired.

The invention belongs to the technical field of radioactive solid waste disposal and in particular relates to a large-sized radioactive solid waste disintegrating protecting device. The large-sized radioactive solid waste disintegrating protecting device comprises a working box, a cutting platform and a roller bed. The box body is provided with a seal cover taking station and a cutting station, and a glove hole and a hagioscope are separately arranged in the two stations, a cutting station is arranged on the cutting platform, and the roller bed is higher than one side of the seal cover takingstation; four lugs are arranged on the top of the box body, and two forking holes are formed in the bottom; a working box door is arranged on the left side of the box body; an air inlet and an air exhaust port of the box body are formed in an upper-lower path manner, the air inlet is formed in the left upper side of the seal cover taking station, and the exhaust port is formed in the lower right side of the cutting station; an air blower is arranged in an outdoor clean area and is connected to an air inlet pipe through an air inlet hose; an air inlet filter and an air inlet manual valve are arranged on the air inlet pipe, and the air inlet filter is provided with an air inlet filter differential pressure gauge; and an exhaust pipe is provided with an air cap, an exhaust filter and an exhaust manual valve. The large-sized radioactive solid waste disintegrating protecting device provided by the invention integrates large-sized waste cutting and disintegrating and barreling, so that radioactive safety of waste cutting and disintegrating personnel is guaranteed.

The invention discloses a compound radiation source intensity backstepping method and a compound radiation source system for a point source-body source combination of a nuclear power plant. In the method, dosage rates of multiple positions are acquired through detectors, a normalization radiation source intensity is dispersed in space, a ray tracking method is utilized to calculate an optical distance, equation set coefficient calculation is developed in combination with information such as the material and an accumulation factor, and a source intensity is back-stepped; dosage rate calculation is carried out for the positions of the detectors, linear regression analysis is carried out on measurement values and calculation values, key parameters such as standard deviation, gradient and intercept are acquired through calculation, and a quality factor is calculated to measure an acceptable degree of each calculation result; a weight iteration method is further proposed, errors caused by the detectors with relatively large uncertainties are reduced, the previous steps are repeated through utilizing an iteration mode till the quality factor reaches a preset value, and the expected radiation source intensity information is acquired.

The invention belongs to the technical field of radioactive substance preparation, and relates to a freely combined low enriched uranium foil target with multiple single pieces. The freely combined low enriched uranium foil target is used for producing fission <99> Mo by means of irradiation. The freely combined low enriched uranium foil target comprises a target barrel, low enriched uranium foils, a target core plug, a pressure plug and sealing covers. The target core plug is placed in the hollow target barrel, so that the multiple pieces of freely combined low enriched uranium foils can be compressed between the inner wall of the target barrel and the target core plug; the target core plug is compressed by the pressure plug in the target barrel from the upper side of the target core plug; two ends of the target barrel are sealed by the sealing covers by means of welding. The freely combined low enriched uranium foil target with the multiple single pieces has the advantages that processes for preparing the freely combined low enriched uranium foil target can be simplified, the irradiation efficiency can be improved, self-shielding can be reduced, the freely combined low enriched uranium foil target is safe in reactor irradiation periods and is easy to release in follow-up procedures, and the quality of the produced <99> Mo can be guaranteed.

The invention discloses a combined radiation source strength backward induction method and a combined radiation source strength backward induction system for a linear source and a surface source of a nuclear power plant. The method is characterized in that dosage rates at a plurality of positions are acquired through a detector; a normalized radiation source strength is diffused in a space; an optical distance is calculated by means of a ray tracking method; an equation set coefficient is calculated according to information such as material and accumulation factor, thereby performing backward induction for obtaining source strength; then dosage rate calculation is performed on detector position; through performing linear regression analysis on a measured value and a calculated value, key parameters such as standard deviation, gradient and intercept are calculated, and furthermore a quality factor is calculated for measuring acceptance degree of a calculation result at each time. Furthermore the invention provides a weighted iteration method. An error which is introduced by a detector with relatively high uncertainty is reduced. The steps are repeated by multiple times in an iteration manner until the quality factor reaches a preset fixed value, and anticipated radiation source strength information is obtained.

The invention discloses a combined radiation source strength backward induction method and a combined radiation source strength backward induction system for a point source and surface source of a nuclear power plant. The method is characterized in that dosage rates at a plurality of positions are acquired through a detector; a normalized radiation source strength is diffused in a space; an optical distance is calculated by means of a ray tracking method; an equation set coefficient is calculated according to information such as material and accumulation factor, thereby performing backward induction for obtaining source strength; then dosage rate calculation is performed on detector position; through performing linear regression analysis on a measured value and a calculated value, key parameters such as standard deviation, gradient and intercept are calculated, and furthermore a quality factor is calculated for measuring acceptance degree of a calculation result at each time. Furthermore the invention provides a weighted iteration method. An error which is introduced by a detector with relatively high uncertainty is reduced. The steps are repeated by multiple times in an iteration manner until the quality factor reaches a preset fixed value, and anticipated radiation source strength information is obtained.

The invention discloses a compound radiation source intensity backstepping method and a compound radiation source system for a point source-line source-body source combination of a nuclear power plant. In the method, dosage rates of multiple positions are acquired through detectors, a normalization radiation source intensity is dispersed in space, a ray tracking method is utilized to calculate an optical distance, equation set coefficient calculation is developed in combination with information such as the material and an accumulation factor, and a source intensity is back-stepped; dosage rate calculation is carried out for the positions of the detectors, linear regression analysis is carried out on measurement values and calculation values, key parameters such as standard deviation, gradient and intercept are acquired through calculation, and a quality factor is calculated to measure an acceptable degree of each calculation result; a weight iteration method is further proposed, errors caused by the detectors with relatively large uncertainties are reduced, the previous steps are repeated through utilizing an iteration mode till the quality factor reaches a preset value, and the expected radiation source intensity information is acquired.

The invention discloses a full-automatic tritium surface pollution inspection system and an application method thereof. The full-automatic tritium surface pollution inspection system comprises an inspection robot; and a control terminal which is in communication connection with the inspection robot through a cable, and provided with a display module. The inspection robot is configured to comprisea vehicle body, and at least two Mecanum wheel sets used for walking are oppositely arranged at the bottom of the vehicle body; and a driving controller is arranged on the vehicle body, is in communication connection with the control terminal and is configured to be in communication connection with the flow control module, the detection module, the vehicle body state information acquisition moduleand the storage module. The full-automatic tritium surface pollution inspection system and the application method thereof aim at sharply increased tritium surface pollution monitoring requirements, the full-automatic tritium surface pollution inspection and measurement robot is designed and processed, has manual and automatic measurement modes, can memorize tritium surface pollution measurement point locations, reads historical measurement tracks (point locations) and performs full-automatic inspection and measurement.

The invention provides a fully automatic shielding transfer door. The fully automatic shielding transfer door is used for transferring materials from the exterior of a hot room to the interior of thehot room, wherein a first through hole communicating with the exterior of the hot room is formed in the horizontal direction of a shielding block, cavities which are matched with the shape of a rotarydoor are formed inside a front seat and a rear seat, and inlets which communicate with the cavities and the first through hole separately are formed in the side, near the exterior of the hot room, ofthe front seat; outlets which communicate with the cavities and the interior of the hot room separately are formed in the side, near the interior of the hot room, of the rear seat, the rotary door isarranged in the cavities, and a second through hole is formed in the horizontal direction of the rotary door; the first through hole, inlets, second through hole and outlets communicate sequentiallyso as to form a passage, and a driving mechanism is used for driving the rotary door to rotate freely in the cavities around a vertical shaft of the rotary door, so that the second through hole in therotary door communicates with or is staggered with the inlets and outlets in the front and rear of the rotary door so as to open or close the transfer door. The transfer door can not only ensure radiation safety of personnel, but also realize automatic operation of the whole process.

The invention discloses a nuclear power plant non point source radiation source intensity backstepping method and a non point source radiation source intensity backstepping system. In the nuclear power plant non point source radiation source intensity backstepping method, the dosage rate of a plurality of positions is obtained through detectors, normalized radiation source intensity is dispersed in space, the optical distance is worked out by means of a ray tracing method, the coefficient of the equation set is calculated by combining information of materials and build-up factors and the like, the source intensity is backstepped, then the dosage rate of the detector positions is calculated, the measured value and the calculated value are subjected to linear-regression analysis, key parameters of the standard deviation, the slope, the intercept and the like are calculated, and furthermore the quality factor is calculated to evaluate the acceptance level of each calculating result . a weighted iteration method is provided at the same time, the error introduced by the detector with relatively-high uncertainty is reduced, the above steps are repeated for many times by means of an iteration manner until the quality factor reaches a preset value, and the expected radiation source intensity information is obtained.

The invention provides a marine electrical penetration assembly local shielding structure and a shielding method. A plurality of layers of honeycomb briquette type local shielding assemblies are additionally arranged in an electrical penetration assembly cylinder, and the honeycomb briquette type local shielding assemblies and a nuclear power ship reactor cabin shield form the electrical penetration assembly local shielding structure which is designed in an electrical penetration-shielding integrated manner and has a complete shielding function. Leakage of penetrating radiation in the reactor cabin to the outside of the reactor cabin caused by drilling of the electrical penetration assembly is reduced to the maximum extent, local shielding of the electrical penetration assemblies of a marine nuclear power platform and a ship nuclear reactor cabin is achieved, and the radiation safety of personnel and equipment outside the reactor cabin is guaranteed. The structure is simple, light in weight, convenient to assemble, low in cost and high in designability. Extra local shielding does not need to be arranged on the periphery of a junction box, the weight reduction effect and the size reduction effect are extremely considerable, and the shielding weight and the shielding size are minimized.

The invention discloses a WWAN antenna suitable for an all-metal rear shell of a notebook computer, the WWAN antenna comprises a notebook computer body, and a WWAN Main antenna, a WWAN MIMO 1 antenna, a WWAN Aux antenna and a WWAN MIMO2 antenna which are arranged on the notebook computer body, the WWAN Main antenna and the WWAN Aux antenna adopt the same structure, and the WWAN MIMO 1 antenna and the WWAN MIMO2 antenna adopt the same structure; the WWAN antenna suitable for the all-metal rear shell of the notebook computer further comprises a metal C shell, a metal D shell, first metal C shell antenna plastic, second metal C shell antenna plastic, a metal D shell antenna plastic window, a C shell WWAN MIMO2 metal ring and a C shell WWAN Aux metal ring. The WWAN antenna suitable for the all-metal rear shell of the notebook computer ensures that the all-metal rear shell notebook computer meets the network access efficiency in a power-on state through terminal wireless communication, induces human body approaching through a P sensor, and adjusts the module output power in time so as to ensure human body radiation safety.