82results about "Measurement with semiconductor devices" patented technology

The invention discloses a semiconductor neutron detector, a preparation method and a neutron flux measuring device, and relates to the field of neutron detection.The semiconductor neutron detector sequentially comprises an ohmic electrode, a substrate, a buffer layer, an epitaxial layer and a Schottky electrode from bottom to top; the substrate, the buffer layer and the epitaxial layer are all made of 4H-SiC; when the semiconductor neutron detector detects neutrons, the neutrons and the epitaxial layer are subjected to nuclear reaction, electron hole pairs excited by neutron energy deposition sequentially pass through the buffer layer and the substrate from the epitaxial layer to the ohmic electrode, and the ohmic electrode outputs neutron current signals which are converted into voltage signals in the sampling circuit so as to perform neutron detection. By adopting the characteristics of high hardness, high thermal conductivity, wide forbidden band and high displacement threshold energy of the 4H-SiC material, the high-fluence-rate neutron flux can be measured in real time, and the reliability and stability of the semiconductor neutron detector are improved.

The invention discloses a high-detection-efficiency groove silicon carbide neutron detector structure, and relates to the technical field of semiconductor neutron detection, a plurality of groove groups are etched on the top surface of a SiC substrate, and each groove group comprises groove units which are sequentially connected along the direction of a first central plane of the SiC substrate; each groove unit comprises a first groove and a second groove which are symmetrically arranged, and the bottom faces of the first grooves and the bottom faces of the second grooves are parallelograms which are symmetrically arranged. According to the neutron detector, the groove structure of a traditional neutron detector is improved, the filling area of the neutron conversion material is increased due to the arrangement of the through groove groups, meanwhile, the contact area of the neutron conversion material and the SiC region is also increased, the self-absorption effect of secondary particles is reduced, and the detection efficiency of the detector is remarkably improved.

A radiation dose quantification method and system, the method comprising: detecting, with one or more detectors with respective sensitive volumes, gamma-rays emitted as a result of capture of neutrons by a composition in a subject subjected to an irradiation program, the composition comprising one or more thermal neutron capture agents, the neutrons having been generated by non-elastic collisions between a primary beam of particles and nuclei in the subject, wherein the particles consist of any one or more of protons, deuterons, tritons and heavy ions, the irradiation program comprises at least one period of irradiation with a beam duration that includes beam-on periods and beam-off periods; applying at least one predefined energy window or filter configured to accept only detection events in the one or more detectors resulting from gamma-rays with an energy indicative of selected gamma-rays arising from the capture of thermal neutrons by the one or more thermal neutron capture agents, wherein the thermal neutrons are neutrons with energies below approximately 0.4 eV; applying a timing window configured to reject or ignore detection events in the one or more detectors resulting from at least prompt gamma-rays produced in non-neutron capture events; and determining the radiation dose of neutron radiation received by the subject during the irradiation program from at least the accepted detection events or determining a dose map of the radiation received by the subject from at least the accepted detection events.

A tape (100) is presented that includes a plurality of superconducting elements (110), such as pixels, distributed along the length of the tape, the tape having a size 101 along a first dimension, such as thickness, that is at least 10 times smaller, e.g., at least 100 times smaller, at least 1000 times smaller, than a size 102 along a second dimension, such as width, and the size 102 along the second dimension, such as width, that is at least 10 times smaller, e.g., at least 100 times smaller, at least 1000 times smaller, than a size 103 along a third dimension, such as length. Further provided are uses of the tape 100, methods of manufacturing the tape, and bolometers and / or kinetic inductance detectors that include the tape 100.

This invention relates to the field of neutron dose monitoring, specifically to a method and apparatus for monitoring the dose equivalent around neutrons over a wide energy range. The apparatus includes an outer moderator, a thermal neutron absorbing layer, a high-energy neutron compensation layer, and an inner moderator. The high-energy neutron compensation layer is composed of a heavy metal material capable of undergoing a significant neutron multiplication reaction with ultra-high-energy neutrons. The thermal neutron detector is a radiation-resistant solid-state semiconductor detector, comprising a silicon carbide semiconductor substrate and an enrichment layer closely adhering to its surface. 6 The device is equipped with a lithium fluoride thermal neutron converter; an initial thermal neutron detector group is deployed within the device to collect thermal neutron fluence information under different slowing degrees; based on the initial thermal neutron detector group, position optimization and weight coefficient output are performed, and multiple groups of thermal neutron detectors in the first depth order are selected to make the total dose response flat in a wide energy range for dose equivalent monitoring.

A neutron flux detector comprising a source material and a solid state radiation detector is disclosed. The source material is configured to produce gamma photons greater than or equal to 6.8 MeV during neutron capture. The solid state radiation detector comprises a Schottky diode and an emitter layer comprising a Compton and photoelectron source material. The emitter layer is configured to receive the gamma photons produced by the source material. The emitter layer is spaced apart from the Schottky diode a distance such that a gap is defined between the emitter layer and the Schottky diode. The distance is selected such that only electrons produced by the emitter layer as a result of the emitter layer absorbing the gamma photons will contribute to a measured output signal of the Schottky diode.

A tape includes a plurality of superconducting elements, such as pixels, distributed along a longitudinal direction of the tape. The tape has a size along a first dimension, such as a thickness, which is at least 10 times smaller, than a size along a second dimension, such as a width, and where the size along the second dimension, such as the width, is at least 10 times smaller, than a size along a third dimension, such as a length. There is also presented a use of the tape, a method of manufacture of the tape, and a bolometer and / or a kinetic inductance detector comprising the tape.

A radiation dose quantification method and system, the method comprising: detecting, with one or more detectors with respective sensitive volumes, gamma-rays emitted as a result of capture of neutrons by a composition in a subject subjected to an irradiation program, the composition comprising one or more thermal neutron capture agents, the neutrons having been generated by non-elastic collisions between a primary beam of particles and nuclei in the subject, wherein the particles consist of any one or more of protons, deuterons, tritons and heavy ions, the irradiation program comprises at least one period of irradiation with a beam duration that includes beam-on periods and beam-off periods; applying at least one predefined energy window or filter configured to accept only detection events in the one or more detectors resulting from gamma-rays with an energy indicative of selected gamma-rays arising from the capture of thermal neutrons by the one or more thermal neutron capture agents, wherein the thermal neutrons are neutrons with energies below approximately 0.4 eV; applying a timing window configured to reject or ignore detection events in the one or more detectors resulting from at least prompt gamma-rays produced in non-neutron capture events; and determining the radiation dose of neutron radiation received by the subject during the irradiation program from at least the accepted detection events or determining a dose map of the radiation received by the subject from at least the accepted detection events.

The invention discloses a neutron-gamma mixed radiation screening method based on a radiation response signal. The neutron-gamma mixed radiation screening method depends on a nano-coating CMOS sensor neutron measurement system. The neutron measurement system of the nano-coating CMOS sensor comprises a detector, the detector comprises a CMOS sensor; a LiF thin film is attached to the light-sensitive surface of the CMOS sensor; the method comprises the following steps that a part of neutrons enable the CMOS sensor to generate a radiation response signal A; the other part of neutrons and 6Li are subjected to nuclear reaction to generate alpha rays and tritium ions, so that the CMOS sensor respectively generates a radiation response signal B and a radiation response signal C; the gamma ray enables the CMOS sensor to generate a radiation response signal D; and according to the difference of the four radiation response signals in morphology and / or gray values, discriminating the neutron-gamma mixed radiation field. The detector formed by combining the LiF film and the CMOS sensor is used for discriminating a neutron-gamma mixed radiation field and can be used for monitoring nuclear leakage of a reactor.

The invention discloses a monitoring device and an experimental device for a neutron poison layer in a spent fuel bearing device, and relates to the technical field of nuclear chemical engineering, the monitoring device comprises a first support and a second support, the bottoms of the first support and the second support are respectively provided with a moving device, and the first support and the second support are arranged in parallel at an interval; the radiation device is connected to the first support, the radiation end of the radiation device faces the second support, and the radiation end is provided with a first moderation layer; the detection device is connected to the second support, the detection device is provided with a second moderation layer, and a plurality of detectors are sequentially inserted into the second moderation layer. According to the technical scheme, the radiation device and the detection device are supported through the first support and the second support respectively, the radiation device and the detection device can move relatively through the moving device, and the monitoring device can flexibly monitor neutron poison layers at different positions of the spent fuel bearing device.

The application provides a parameterized real-time measurement method for field neutron energy spectrum and dose, which comprises the following steps: using a three-section function model to represent the field neutron energy spectrum; constructing simulated field neutron energy spectrums of multiple moderation levels, and using a detector to measure and calculate a thermal-fast ratio; simulating and calculating the simulated field neutron energy spectrum of a test point by using a Monte Carlo model to obtain model parameters; establishing a corresponding relationship between the model parameters and the thermal-fast ratio; measuring the fast neutron pulse count rate and the thermal neutron pulse count rate by using the detector in the field, determining the value of the model parameters by looking up a table or interpolation, and the like according to the corresponding relationship between the model parameters and the thermal-fast ratio; calculating the field neutron energy spectrum according to the model parameters, and calculating the field neutron dose by using the field neutron energy spectrum and a neutron fluence-dose conversion coefficient.

The invention discloses a wide-range neutron measurement method, and belongs to the technical field of neutron detection and measurement, and the method comprises the steps: selecting a wide bandgap semiconductor material, an electrode metal material, a neutron conversion material and a neutron conversion material coating method, obtaining a wide bandgap semiconductor radiation detector coated with the neutron conversion material, and taking the wide bandgap semiconductor radiation detector as a neutron detector; a neutron detector is connected with a pre-amplifier, the pre-amplifier is connected with a rear-end digital sampling circuit, the rear-end digital sampling circuit processes pulse waveforms output by the pre-amplifier, and digital pulse signals are output; processing the digital pulse signal by using a pulse counting method and a Campbell method in upper computer software to obtain a pulse counting rate value and a Campbell value, and calculating neutron fluence rates obtained by the two methods; and obtaining the neutron fluence rate as an output result by adopting different methods according to the size of the neutron fluence rate. According to the invention, a neutron fluence rate measurement system with a small size, high temperature resistance, radiation resistance and a wide range can be obtained, and real-time accurate measurement under the condition of large-range change of the neutron fluence rate is realized.

Lithium-containing thiostannate spinel compounds having the formula Li2M1+xSn3−xS8, where x is 0 or 1 and M is Mg, Fe, Mn, Ni, Ga, In, or a combination thereof; or the formula Li1.66CuSn3.33S8 are provided. Methods and devices for detecting incident neutrons and alpha-particles using the compounds are also provided. For thermal neutron detection applications, the compounds can be enriched with lithium-6 isotope (6Li) to enhance their neutron detecting capabilities.

A neutron flux detector comprising a source material and a solid state radiation detector is disclosed. The source material is configured to produce gamma photons greater than or equal to 6.8 MeV during neutron capture. The solid state radiation detector comprises a Schottky diode and an emitter layer comprising a Compton and photoelectron source material. The emitter layer is configured to receive the gamma photons produced by the source material. The emitter layer is spaced apart from the Schottky diode a distance such that a gap is defined between the emitter layer and the Schottky diode. The distance is selected such that only electrons produced by the emitter layer as a result of the emitter layer absorbing the gamma photons will contribute to a measured output signal of the Schottky diode.

A radiation detector is disclosed. The detector comprises a translation device configured to produce a voltage upon receiving incident radiation, and an operable switch configured to transition between a first state and a second state upon triggering by a threshold voltage. The translation device is connected to the switch such that the switch is triggered to transition from the first state to the second state upon the threshold voltage occurring across the translation device. The detector further comprises an interrogation circuit operable to determine whether the switch is in the first state or the second state, thereby determining whether the detector has received a threshold level of radiation associated with the threshold voltage. A system is also disclosed that includes a radiation sensitive device and a radiation detector.

The invention belongs to the field of geological exploration, and discloses a logging-while-drilling device based on the muon imaging technology. Comprising an outer shell, an inner cylinder and a muon detector, the inner cylinder is coaxially nested in the outer shell, and a relatively sealed detection space is formed between the inner cylinder and the outer shell; the muon detector is arranged in the detection space; the muon detector comprises a scintillator assembly and a photoelectric detector, the scintillator assembly is coaxially sleeved in the detection space, the photoelectric detector is arranged on the inner side of the scintillator assembly, the scintillator assembly is made of a scintillator material capable of resisting the high temperature of 140 DEG C or above, and a heat insulation layer is arranged on the inner wall of the detection space. The method can be widely applied to oil-gas exploration, mineral resource investigation and underground structure detection, and the detection depth and imaging precision of logging while drilling are improved.

Neutron detector comprising an optical fission chamber with a reflective optical cavity and a detection head coupled to the chamber comprising an optical lens and an optical fiber coupler. The invention relates to a neutron detector (1) comprising at least one sealed ionization chamber (2) with optical transduction, called an optical fission chamber (OFC) with an optical cavity whose operation is based on optical transduction and which integrates at one of its longitudinal ends, a porthole and an optical lens as an optical interface and, at the other end, a mirror for reflecting photons which take a direction opposite to that of the porthole. Figure for abstract: fig.9

An embodiment of the present application provides a method for measuring the cadmium ratio of a thermal neutron field, which includes placing a first object to be measured at a set position in the beam of the thermal neutron field, and monitoring the reactor power through a monitor. Placing the first object to be measured outside the beam, and measuring the quantity of activation products generated by the first object to be measured through a detector. Placing a second object to be measured at the set position, and monitoring the reactor power through a monitor. Placing the second object to be measured outside the beam, and measuring the quantity of activation products generated by the second object to be measured through a detector. One of the first object to be measured and the second object to be measured is a first material to be measured, and the other includes a second material to be measured and a cadmium material layer wrapped outside the second material to be measured, and the first material to be measured and the second material to be measured are of the same material. Calculating the cadmium ratio according to the measurement result of the detector and the monitoring result of the monitor. The method for measuring the cadmium ratio of the thermal neutron field in the embodiment of the present application has higher measurement accuracy.

The present invention discloses a method for obtaining the energy spectrum and angular distribution of a deuterium-tritium neutron source. The method adopts a D-T neutron radiation field measurement system based on a SiC detector, can simultaneously measure the companion particles of the two reactions, can distinguish them in the energy spectrum, and obtain the counting rates of alpha particles and protons; takes into account the influence of the D-D reaction caused by target deposition in the D-T reaction, obtains the energy spectrum and angular distribution of the D-T and D-D neutron sources, and effectively improves the accuracy of the energy spectrum and angular distribution; the method of the present invention has no requirements for target tube-related parameters such as a pipeline and a grating for the companion particles, is simple to operate, only needs to measure a set of data to obtain the energy spectrum distribution in the entire space, does not need to measure different angles one by one, and can measure all D-T neutron sources; and the method of the present invention can be applied to a DC beam or pulsed beam D-T accelerator neutron source with a reserved companion particle channel, especially when there is deuterium beam deposition on the target, and can more accurately obtain the energy spectrum and angular distribution of the D-T neutron source.

The neutron beam detection device of the present application is provided with: a first solar cell type detector, which has a conversion film attached to its surface that converts neutrons into any one of a charged particle beam of a photon, an alpha particle, a proton, a lithium nucleus, a gamma ray, and a beta ray, and generates a current by incident radiation; a radiation detector that generates a current having no sensitivity to neutrons as an output signal due to radiation incident; a current measurer that detects the current generated by the first solar cell type detector and the current generated by the radiation detector due to radiation incident as signals; and a flux calculation section that compares the current signals from the detector detected by the current measurer. The flux calculation section corresponds the detected current signals from the solar cell type detector and the radiation detector to a relationship between the flux of a prescribed kind of incident radiation and the detected current from the solar cell type detector and the radiation detector, which is obtained in advance, and calculates the flux of the neutron beam.

A neutron detector comprising an optical fission chamber with a reflective optical cavity and a detection head coupled to the chamber comprising an optical lens and an optical fiber coupler. The invention relates to a neutron detector (1) comprising at least one sealed ionization chamber (2) with optical transduction, called an optical fission chamber (OFC) with a single optical cavity, the operation of which is based on optical transduction and which integrates, at one of its longitudinal ends, a window and an optical lens as an optical interface and, at the other end, a mirror to reflect photons traveling in the opposite direction to that of the window. Figure for the abstract: Fig. 9

A method of compensating for high energy radiation damage to a semiconductor, in a semiconductor-based radiation detector, the method including: (a) detecting, using the detector, a radiation signal, wherein the detector comprises the semiconductor; (b) detecting, using a temperature sensor, a temperature of the detector; (c) detecting a leakage current in the detector; (d) determining a compensated leakage current based on the detected temperature; (e) determining a high energy radiation exposure based on the compensated leakage current; (f) determining, using the detected radiation signal, a compensated radiation signal based on the determined high energy radiation exposure; and (g) outputting the compensated radiation signal.

The present invention provides a gadolinium-based neutron / gamma discrimination system, comprising a gadolinium-based neutron detector, a gamma-ray detector, a preamplifier circuit connected to the detector, and a dual-channel digital multichannel spectrometer. The dual-channel digital multichannel spectrometer receives two pulses amplified by the preamplifier circuit. When the time difference between the arrival times of the two pulses is within a set threshold, the system counts neutron signal events and further extracts amplitude values, recording them in the neutron response energy spectrum and gamma energy spectrum, respectively. The present neutron / gamma discrimination system combines physical processes with electronic technology. A thin sheet of gadolinium-based material absorbs only internal conversion electrons and low-energy background gamma rays. The system then uses the temporal relationship between internal conversion electrons and secondary prompt gamma rays to further distinguish whether the signal detected by the gadolinium-based material sheet is internal conversion electrons or low-energy background gamma rays, thereby achieving neutron / gamma discrimination.

A neutron detector for detecting neutrons with energies from meV to tens of MeV comprising one or more nitride (BN) strips electrically connected in parallel or series. In some embodiments, the two or more BN strips are stacked on one another. In other embodiments, the two or more BN strips are disposed on a substrate with a gap between the two or more BN strips.

A hybrid radiation measurement device and a neutron dose measurement method are disclosed. The hybrid radiation measurement device includes a housing and an inner core. The housing surface includes a window; a cadmium filter is placed on the window; the inner core includes a solid track detection element and a thermoluminescent element group; when the inner core is disposed within the housing, the thermoluminescent element group is located below the window, and the solid track detection element is offset from the area where the window is located; the thermoluminescent element group includes at least one incident thermoluminescent element and at least one albedo thermoluminescent element. The hybrid radiation measurement device of this application can measure both photons and neutrons.

To provide a technology that is advantageous in improving the cooling performance of a detector that detects radioactive rays and enhancing the sensitivity or the resolution of the detector.SOLUTION: A detector includes: a first member that detects a radioactive ray; a second member located around the first member; a third member having a first opening overlapping at least a part of a first region on which the first member is projected in a planar view including the first member and the second member; a fourth member having a second opening overlapping at least a part of the first region in the planar view and having higher thermal conductivity than the third member; and a penetration part provided in a third region overlapping the fourth member outside a second region on which the second member is projected in the planar view, connected to the fourth member, connected to the second member via a thermal conductive member, and having higher thermal conductivity than the third member.SELECTED DRAWING: Figure 1B

The invention relates to an online single-sphere neutron gamma-ray spectrum detector and a detection system. Wherein the detector comprises a spherical moderation body and a plurality of detectors; the plurality of detectors are arranged on the spherical moderator, the distances from the plurality of detectors to the sphere center of the spherical moderator are different, the plurality of detectors are arranged at the same depth from the sphere center of the spherical moderator, and the plurality of detectors at least comprise a CLYC detector. According to the invention, the plurality of detectors are arranged in the same spherical moderator, and the plurality of detectors are arranged at positions with different distances from the sphere center of the spherical moderator, so that large-range energy detection can be realized, the size is small, the weight is light, and carrying is convenient; moreover, the device comprises a CLYC detector, so that the device has the capability of detecting neutrons and gamma information at the same time.

An X-ray semiconductor detector (2) comprises a first i-type nitride semiconductor substrate (10), a second i-type nitride semiconductor substrate (15), a p-type semiconductor layer (20), an n-type semiconductor layer (22), a first electrode (25), and a second electrode (26). The first i-type nitride semiconductor substrate (10) includes a first gallium surface (11). The second i-type nitride semiconductor substrate includes a second gallium surface (17). The p-type semiconductor layer (20) is formed on the first gallium surface. The n-type semiconductor layer (22) is formed on the second gallium surface. The first electrode (25) is formed on the p-type semiconductor layer (20). The second electrode (26) is formed on the n-type semiconductor layer (22).

This invention discloses a ring-shaped trench silicon carbide neutron detector, comprising an array of ring-shaped trenches etched on the top surface of a substrate. The ring-shaped trench array includes a first ring-shaped trench and a second ring-shaped trench. The second ring-shaped trench is spaced around the periphery of the first ring-shaped trench. The first and second ring-shaped trenches are concentrically arranged with reference to the center point of the top surface of the substrate. The first and second ring-shaped trenches have the same width, and the trench spacing between the first and second ring-shaped trenches is the same as the distance of the second ring-shaped trench from the boundary of the top surface of the substrate. Both the first and second ring-shaped trenches are filled with neutron conversion material. This invention, through the construction of the ring-shaped trench array, increases the filling area of ​​the neutron conversion material and the contact area with the substrate region, thereby reducing the self-absorption effect of secondary particles and significantly improving the detection efficiency of the detector.

The invention relates to the field of neutron dose monitoring, in particular to a wide-energy-region neutron surrounding dose equivalent monitoring method and device. The device comprises an outer-layer moderator, a thermal neutron absorption layer, a high-energy neutron compensation layer and an inner-layer moderator, wherein the high-energy neutron compensation layer is made of a heavy metal material which can generate a remarkable neutron multiplication reaction with ultrahigh-energy neutrons; the thermal neutron detector is a radiation-resistant solid-state semiconductor detector and comprises a silicon carbide semiconductor substrate and a < 6 > Li-enriched lithium fluoride thermal neutron converter tightly attached to the surface of the silicon carbide semiconductor substrate; an initial thermal neutron detector group used for collecting thermal neutron fluence information under different moderation degrees is arranged in the device. And based on the initial thermal neutron detector group, performing position optimization and weight coefficient output, and selecting a plurality of groups of thermal neutron detectors in a first depth order, so that the total dose response is flat in a wide energy region and dose equivalent monitoring is performed.