62 results about "Neutron dose" patented technology

There is provided a neutron capture therapy apparatus (1) capable of measuring the neutron dose in real time while suppressing a reduction in the measurement accuracy of a neutron beam (N). A real-time dose output unit (40) including a scintillation detector (41), which detects the neutron beam (N) and outputs a signal in real time, and a neutron dose output unit (44), which converts the signal, output from the scintillation detector (41) into a neutron dose using a correction coefficient (45) and outputs the neutron dose in real time, and a correction coefficient setting unit (60) that sets the correction coefficient 45 are provided. The correction coefficient setting unit (60) includes a gamma ray detection section (62) that detects a gamma ray emitted from a gold wire (61) activated by irradiation of the neutron beam (N), and modifies the correction coefficient (45) based on a signal, which is output from the scintillation detector (41) by irradiation of the neutron beam (N), and a dose of the neutron beam (N), which is acquired based on the gamma ray detected by the gamma ray detection section (62).

The invention aims at providing a compensation type neutron dose instrument which comprises a packaging layer, a slowing-down sphere body and a heavy metal layer, wherein the slowing-down sphere body is arranged in the packaging layer, a neutron fission multiplication layer is arranged between the slowing-down sphere body and the packaging layer, the heavy metal layer is arranged in the slowing-down sphere body, a counter is arranged in the heavy metal layer, and an air rod is arranged in the slowing-down sphere body along a radial direction. According to the invention, neutron fluence response in a heat energy region can be improved, the neutron fluence response in an intermediate energy region can be correspondingly reduced, and accompanying Gamma rays in a neutron radiation field can be shielded.

The invention discloses a device and a method for [gamma] and neutron double ray power spectrum measurement. The [gamma]-ray signal and the neutron signal are distinguished by particle discrimination by using the feature of a scintillation crystal being sensitive to the [gamma]-ray and the neutron at the same time. The pulse amplitudes of the two signals are analyzed respectively to obtain the [gamma]-ray deposition spectrum and the neutron deposition spectrum. The [gamma]-ray has a peak response in the scintillation crystal, and the [gamma]-ray spectrum is the measured [gamma]-ray power spectrum. The neutron is continuously responsive in the scintillation crystal. The neutron power spectrum is solved by the least squares of the response matrix and the neutron deposition spectrum of a single energy neutron. Finally, by analyzing the energy spectrum data, the key nuclide is identified and the content thereof is calculated, and the [gamma] dose and the neutron dose are calculated. The portability of the instrument for mixed radiation field measurement is effectively improved and the mutual interference between the [gamma]-ray and the neutron is avoided.

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 belongs to the field of radiation measurement, and particularly relates to a plastic scintillator doped with the neutron-sensitive material lithium and a method thereof for measuring thermal neutrons. According to the invention, lithium-contained raw materials are added into the plastic scintillator. The lithium-doped plastic scintillator and a photomultiplier tube are connected and are sealed inside an opaque housing so as to form a probe. The probe is placed in a measurement environment and the nuclear reaction occurs between incident neutrons and the lithium-contained raw materials in the plastic scintillator, so that the materials are in an excited state. During de-excitation, photons are sent out. The number of photons is proportional to the number of incident neutrons. Meanwhile, through recording the electrical impulses semaphore of the photomultiplier tube per unit time, the neutron fluence can be calculated, so that the measurement of the neutron fluence is realized. The neutron fluence is multiplied by a fluence-dose conversion coefficient to obtain the neutron dose equivalent. The plastic scintillator and the method are low in cost, wide in application range and high in detection efficiency.

The invention provides a method for establishing a geometric model based on medical image data, comprising the steps of: inputting or reading medical image data; defining organization type and densityinformation; The step of giving 10B concentration information based on the measured overall 10B concentration spatial or temporal distribution; Establishing a 3D encoding matrix with tissue type, density and 10B concentration information; Steps to generate a geometric model. Another aspect of the present invention provides a dose calculation method based on medical image data, comprising the steps of calculating a 10B dose (DB), a neutron dose (DN), and a photon dose (D gamma) of the obtained geometric model according to a Monte Carlo method; The step of calculating a dose to tumor and tissuebased on the measured spatial or temporal distribution of the overall 10B concentration (Bcon). The geometric model building method based on medical image data of the invention can improve the accuracy and precision of the lattice information in the model. The dose calculation method based on the geometrical model can improve the reliability of dose calculation to improve the quality of treatment.

The invention aims to disclose a card type personal neutron dosimeter. The card type personal neutron dosimeter comprises a neutron probe and a control system, wherein the neutron probe is provided with a neutron detection element which is a 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; the signals of the neutron probe are processed by the control system; the 6LiI (Eu) scintillation crystal is used as the neutron detection element; and a matched photodiode is used as a photoelectric transducer. The card type personal neutron dosimeter is light in weight, is convenient to carry, can sensitively response neutrons, can determine the dose equivalent of the whole-body neutrons received by a person, has the alarming functions such as sound, light, vibration and the like, can be used for monitoring personal neutron dose of radioactive employees so as to ensure that the neutron radiant quantity at the surrounding environment of the radioactive employees is within the safe range, thereby achieving the purpose of the invention.

The invention provides a device and method for measuring neutron dose equivalent. The device comprises multiple mutually-isolated liquid scintillator counting zones, a light-transparent glass surface, a light guide, a photomultiplier, and a data processor. The detecting material of each liquid scintillator counting zone is a liquid scintillator used for interacting with neutron to generate photons. One end of each liquid scintillator counting zone is sealed by the light-transparent glass surface. The photons generated by the interaction of the neutrons and the liquid scintillators pass through the light-transparent glass surface to be output to the light guide and then are collected by the light guide to be output to the photomultiplier to be counted in order that the sum of the photons generated by all the liquid scintillator counting zones is acquired. The data processor is used for computing the sum of the generated photons in order to obtain the dose equivalent of a neutron radiation field. The device and method in the technical scheme of the invention may measure neutrons with different energies in segments and accurately acquire the neutron dose equivalent so as to achieve neutron dose equivalent measurement with high sensitivity and a wide range.

The invention discloses a fixed radiation monitoring system. The system comprises a control part, at most eight detectors and an ionization chamber, the detectors are located on the ionization chamberand connected with the control part, and the detectors are one or more selected from a low-range gamma detector, a wide-range gamma detector, a NaI gamma detector, a high-range gamma detector and a neutron detector. Thus, the structure is simple, the stability is high, the energy response is sound, the dose rate range is wide, the angular response and cosmic ray response are sound, and the mechanical property is reliable; and the system can work for a long time, does not reduce the monitoring performance due to the working time, and can be applied to measurement and real-time alarm of the environment X-gamma external irradiation air absorption dose rate and neutron dose equivalent, dose rate monitoring alarm of a radiation diagnosis and treatment workplace, monitoring of a medical X-ray machine and an industrial radioactive source, dose rate monitoring of an underground radiation environment, and the like.

An apparatus and method for improving the sensitivity and energy response of neutron detectors and neutron dose rate meters. A beryllium layer is added to neutron detector moderators to improve the sensitivity of the detector. Energy dependence of the sensitivity is optimized by controlling the amount of beryllium in the moderator and by specifying the geometrical design parameters. The beryllium layer, in combination with additional material layers in the moderator, makes the detector response function correspond to the theoretical one in a wide range of energies. Response parameters of the neutron dose rate meter are within 20% of the theoretical response function in the neutron energy range from 500 keV to 10 GeV, and also in the energy range corresponding to thermal neutrons (about 1-100 meV).

The invention provides a method and device for measuring neutron dose equivalent. The method comprises steps of: establishing a liquid scintillator counting zone, a central counting zone, and a peripheral counting zone, wherein the liquid scintillator counting zone comprises a liquid scintillator, the central counting zone comprises a first neutron counter, the peripheral counting zone comprises multiple second neutron counters; acquiring first neutron dose equivalent by means of the interaction of the neutron and the liquid scintillator; acquiring second neutron dose equivalent by means of the interaction of the neutron and the first neutron counter; acquiring third neutron dose equivalent by means of the interaction of the neutron and the second neutron counters; and computing the first neutron dose equivalent, the second neutron dose equivalent, and the third neutron dose equivalent in order to obtain the dose equivalent of a neutron radiation field. The method and the device in the technical scheme of the invention may measure low-energy, intermediate-energy, and high-energy neutrons relatively independently so as to achieve neutron dose equivalent measurement with high sensitivity and a wide range.

The invention relates to a fixed integrated device for monitoring regional gamma neutron radiation. The device comprises a gamma detector, a neutron detector and a host system connected with the gammadetector and the neutron detector through an RS-485 communication network, wherein the host system is connected with a detection signal processing and display system, the gamma detector comprises a plurality of gamma probes, the neutron detector comprises a plurality of neutron probes, the detection signal processing and display system comprises a signal processor connected with the gamma probesand the neutron probes, and the signal processor converts detector signals of the gamma probes and the neutron probes into digital signals. The device is advantaged in that gamma and neutron dose monitoring and dose data management in complex radiation places such as a nuclear reactor are completed by using one dose monitoring system, a GM detector and a 3He proportional counter are selected as main detection instrument probes and are organically combined into a region gamma and neutron detection device, the monitor is beneficial to simplifying equipment complexity of a radiation monitoring system in a nuclear power facility place, and the arrangement space is saved.

The invention relates to a system and method for automatically drawing a nuclear power station neutron dose equivalent rate distribution diagram, and belongs to the technical field of neutron dose detection devices. The system comprises a wireless network base station, a wireless neutron dose equivalent rate instrument, a wireless communication terminal, a wireless neutron dose equivalent rate instrument server, a touch display screen terminal, a voice terminal, a wireless network server, a wireless voice server and a wireless network controller, wherein the wireless network base station, the wireless neutron dose equivalent rate instrument server, the touch display screen terminal, the voice terminal, the wireless network server, the wireless voice server and the wireless network controller are all in wireless communication with one another through an interconnected generic network, and the wireless network base station is in wireless communication with the wireless neutron dose equivalent rate instrument and the wireless communication terminal. According to the system, the neutron dose equivalent rate and the geographical information in the position where the system is located can be automatically obtained, and therefore the nuclear power station neutron dose equivalent rate distribution diagram can be rapidly and accurately drawn and visually displayed.

The invention relates to a method for measuring neutron dose equivalent, and in particular relates to a method for measuring neutron dose equivalent by utilizing a semiconductor array. The method comprises the following steps of: three semiconductor array pipes are inserted a polyethylene degradation body according to a three-dimensional coordinate axis manner, wherein the sum of counting of semiconductor detectors is fluence of an incident neutron; the slowing-down distance of each semiconductor detector is different, therefore an energy spectrum of the incident neutron is obtained in a reverse thrust manner; and the dose equivalent of the neutron can be obtain through calculating by combining the fluence and the energy spectrum. According to the invention, the uncertainty of a measurement result is much less than that of a tour gauging type instrument, meanwhile, the structure of the three semiconductor array pipes are adopted, therefore the tour gauging type instrument maintains the advantages of light weight and small volume.

The invention discloses a neutron dose rapid determining method, which solves the problems of long time and low accuracy rate of determining three-dimensional irradiation dose of boron neutron capture in the prior art. The neutron dose rapid determining method comprises the steps of: 1) forming phantom data of neutron irradiation on a CPU platform, carrying out discretization on the phantom data, forming calculation voxels, and determining a substance mark number of each calculation voxel in a three-dimensional space; 2) selecting a GPU platform used for determining a dose of a neutron beam, and transmitting calculation data to the GPU platform by means of the CPU platform; 3) and parallelly calculating doses of units of a three-dimensional irradiation dose matrix D by using a Monte Carlo program on the GPU platform, and returning data of the irradiation dose matrix D to the CPU platform by means of the GPU platform. The neutron dose rapid determining method realizes the effect of modifying an original method of operating in the CPU into the method of operating in the GPU, greatly accelerates the operation speed, and effectively ensures the precision of calculation.

The invention provides a neutron and gamma integrated electronic dose monitoring device which realizes the function of simultaneously monitoring the neutron external irradiation dose and gamma external irradiation dose of a worker in a neutron and gamma mixed complex radiation field by integrating the measurement functions of neutron dose and gamma dose measurement, is convenient to carry. Instrument carrying burden of workers is reduced, and radiation safety of the workers is ensured. According to the invention, the neutron dose value and the gamma dose value are measured and displayed in real time and are visual and easy to read, an alarm is given when the dose exceeds the limit, and the personal radiation dose of workers is ensured to be at a reasonable and as low as possible level.

The invention discloses a high response neutron dose equivalent meter. The probe of the equivalent meter is selected from 3He counter tube, the length thereof is 95-110 centimeter, the diameter thereof is 5-6 centimeter, the air pressure thereof is 3-4 atmospheric pressure. At the same time, the cadmium absorber of probe is evenly drilled, wherein the hole area is 0.9-1 square centimeter, the hole area accounts for 30-40% of the whole cadmium slice area. The sensitivity of the neutron dose equivalent meter can be 48.9cps / uSv h-1 to satisfy the requirement of the detection of the environment level neutron radiation field.

The invention belongs to the field of radiation measurement, and particularly relates to a plastic scintillator doped with the neutron-sensitive material uranium and a method thereof for measuring thermal neutrons. According to the invention, uranium-contained raw materials are added into the plastic scintillator. The uranium-doped plastic scintillator and a photomultiplier tube are connected and are sealed inside an opaque housing so as to form a probe. The probe is placed in a measurement environment and the nuclear reaction occurs between incident neutrons and the uranium-contained raw materials in the plastic scintillator, so that the materials are in an excited state. During de-excitation, photons are sent out. The number of photons is proportional to the number of incident neutrons. Meanwhile, through recording the electrical impulses semaphore of the photomultiplier tube per unit time, the neutron fluence can be calculated, so that the measurement of the neutron fluence is realized. The neutron fluence is multipled by a fluenc-dose conversion coefficient to obtain the neutron dose equivalent. The plastic scintillator and the method are low in cost, wide in application range and high in detection efficiency.

Provided are a method for measuring dose distribution in a mixed radiation field of neutrons and gamma rays, and a method for measuring beam uniformity of a mixed radiation field of neutrons and gamma rays. The planar dose measuring method includes: a step of obtaining a total dose of neutrons and gamma rays by measuring with a dosimeter; and a step of analyzing a neutron dose. The method may effectively measure the doses of neutrons and gamma rays, may be applied to beam measurement and treatment plan validation, and thus improve the quality of treatment.

The invention discloses a method for measuring neutron dose rate by means of a cerium bromide detector and a neutron dose instrument, wherein according to the method for measuring the neutron dose rate, a deterministic function relation between a net counting rate of a neutron characteristic gamma energy peak which is generated by the neutron in the cerium bromide detector and the neutron dose rate which is caused by the neutron at the point, through measuring the gamma energy spectrum and using the deterministic function relation, the neutron dose rate is calculated and obtained. Therefore, the method for measuring the neutron dose rate according to the embodiment of the invention can acquire the neutron dose rate more conveniently, quickly and accurately.

The invention provides special protective glass which can effectively protect thermal neutrons. The special protective glass comprises the following ingredients by weight percent: 20.0-50.0% of SiO2, 5.0-20.0% of PbO, 5.0-35.0% of B2O3, 3.0-12.5% of Na2O, 1.0-15.0% of Gd2O3, 5.0-30.0% of BaO and 2.0-10.0% of Li2O. Through reasonable proportioning of all the components and the component contents, the special protective glass with the 550 nm transmissibility of higher than 85%, the knoop harness of larger than 560*107 MPa, the neutron dose rate lowered to 50%, the thickness of smaller than 40 mm and the capability of effectively protecting thermal neutrons is obtained.

The invention discloses a silicon PIN neutron dose detector and a manufacture method thereof. The silicon PIN neutron dose detector comprises a high-resistance silicon chip, as well as a P type active area and an N type active area, which are respectively positioned on the front surface and the back surface of the silicon chip, wherein one of the active areas is circular, and the other active area is annular. According to the silicon PIN neutron dose detector disclosed by the invention, the P type active area and the N type active area are designed into of an upper and down asymmetric structure, so that the distance between the P area and the N area is greater than the thickness of the silicon chip, the distance between the two active areas is effectively increased, namely the effective thickness of an I region of the detector is increased, and the restriction of the thickness of the silicon chip on the thickness of the I region of a PIN tube is also broken through. In addition, the annular active area can generate more uniform current distribution through the symmetry of the structure of the annular active area.

The invention discloses a plant seed neutron irradiation dose distribution device based on a compact D-D neutron generator. The device comprises a side reflector and a plurality of seed placement laminates, the side reflector is a frame structure with an opening at one end, the seed placement laminates are clamped in the side reflector, the seed placement laminates are arranged in the side reflector, and a neutron irradiation area is formed by an area between two adjacent seed placement laminates and the side reflectors. The neutron irradiation area can contain a large number of seeds at a time, the seed placement laminates are evenly distributed up and down in the vertical direction, the neutron absorption dose of the placed seeds appears in a uniform gradient, and after each position where the seeds are placed is numbered, the dose of neutrons absorbed by each seed can be obtained through the Monte Carlo simulation technology, the application range of the compact D-D neutron generator in neutron radiation biological research and irradiation mutation breeding is expanded, and meanwhile a basis can be provided for the neutron radiation biological research and the irradiation mutation breeding.

The invention provides a neutron dosimeter based on a current integration type electronics system, and belongs to the technical field of radiation dose detection. The neutron dosimeter comprises a probe module and a circuit module. Wherein the probe module comprises an outer-layer moderator, a metal layer, a polymer layer, an inner-layer moderator and a counter which are sequentially arranged from outside to inside, and neutrons can generate nuclear reaction in the counter to generate current signals. The circuit module comprises a signal amplification circuit and a data processing circuit, the signal amplification circuit can convert the current signal into a voltage signal, and the data processing circuit can convert the voltage signal into the current signal and can determine the radiation dose value of neutrons according to the current signal. In the detection process, the current neutron radiation dose value can be determined by acquiring the total charge quantity generated by the current neutrons without recording the number of the current pulse signals. Therefore, missed counting is avoided, and the detection precision of the neutron dosimeter is improved.

The invention relates to a plutonium material neutron radiation field calculating method, which comprises the following steps: analyzing radionuclides and possible neutrons in a highly-enriched plutonium material; analyzing a neutron generation mechanism of the highly-enriched plutonium material, and calculating plutonium isotope spontaneous fission radiation neutron yield and neutron yield obtained after alpha released in alpha decay process is subjected to (alpha, n) reaction; and when neutrons pass through unit thickness material, analyzing probability of the neutrons separating from a neutron group higher than some energy threshold and entering a neutron group having lower energy, and based on the removal cross section theory, calculating neutron dose equivalent rate of the highly-enriched plutonium.

A method for measuring radiation intensity includes measuring the radiation intensity received by the protein in a radiation field based on degree of protein degradation in the radiation field, wherein the degree of degradation is a ratio of the molecular weight of the protein before and after irradiation. The measuring method is simple in operation, small in number of steps, small in error, and capable of measuring radiation doses of various radiation fields or even mixed radiation fields. Use of a biological dosimeter for measuring the radiation intensity by the method in a neutron capture therapy system can not only assess radiation contamination in the irradiation chamber, but also evaluate the neutron dose.

The invention relates to a multipurpose neutron terminal, which is mainly applied to dose calibration of a neutron dose rate instrument, material shielding performance and radiation detection in the field of radiation protection as well as energy measurement performance evaluation and position resolution evaluation of a neutron detector in the field of application. The multipurpose neutron terminal mainly comprises a neutron source, a control system and a collimation moderation system. The control system is used for hoisting the source to a proper height, and the multipurpose neutron terminalcan be used for instrument calibration with the application of a shadow cone. If the neutron source is matched with a movable collimation moderation system, a moderation shielding body can be freely replaced according to requirements, the neutron source can be perfectly embedded into the system, so that beams of the source cannot leak outwards, and neutron beams with different energy spectrums, different positions and different shapes can be obtained through using collimation holes. The multipurpose neutron terminal has the advantages of safety, stability, multiple functions, convenience, easiness in implementation and the like.

The invention provides a wide-energy neutron dose equivalent rate instrument based on a gamma ray energy spectrum detector. The wide-energy neutron dose equivalent rate instrument comprises a sphere,the outermost layer of the sphere is a neutron non-elastic scattering material layer, and the sphere center of the sphere is provided with the gamma ray detector; a plurality of layers of neutron moderation bodies and a plurality of layers of thermal neutron capture materials are arranged between the gamma ray detector and the neutron non-elastic scattering material layer, and the neutron moderation bodies and the thermal neutron capture materials are arranged in an interlayer staggered mode; and the gamma ray detector is a lanthanum bromide detector. After neutrons enter the measurement system, instantaneous gamma rays generated by inelastic reaction or radiation capture of the neutrons and an intermediate material are detected by the gamma detector, a neutron energy spectrum can be obtained through a gamma energy spectrum unfolding technology, and finally the neutron energy spectrum is converted into a neutron dose equivalent rate.