177 results about "Radionuclide" patented technology

A structure and process for containing a nuclear reaction is disclosed. The structure and process may involve a nuclear containment structure including a plurality of metallic rings stacked axially to form the nuclear containment structure, the nuclear containment structure being configured to: mitigate, by using a physical barrier, the movement of radionuclides from inside a containment structure to a surrounding space; shield, by varying the alloy properties throughout the containment structure, an exterior of the containment structure from radiation produced inside of the containment structure; regulate, by transferring thermal energy, a temperature within the containment structure; regulate, by storing thermal energy, the temperature within the containment structure; shield, components within the nuclear containment structure from kinetic events external to the nuclear containment structure; and shield, the external environment, from kinetic events within the nuclear containment structure.

The invention relates to a radionuclide-labeled glucose derivative molecular probe and application thereof, and belongs to the technical field of nuclear medicine. According to the present invention, propargyl glycine is used to synthesize an amino acid skeleton, and the amino acid skeleton reacts with glucose azide to prepare a series of radionuclide-labeled glucose derivative molecular probes, such as [68Ga] Ga-NOTA-SDG, [68Ga] Ga-NOTA-DDG, [68Ga] Ga-NOTA-TDG, [68Ga] Ga-DOTA-SDG, [68Ga] Ga-DOTA-DDG and the like; the glucose derivative molecular probe labeled by the series of radionuclides has the advantages of being simple and rapid in labeling, high in radiolabeling yield and good in stability, can accurately position tumors, has good tumor uptake and lower non-target tissue uptake, and has great application prospects in preparation of broad-spectrum tumor imaging agents.

The invention provides an off-site full-range radiation risk evaluation method and system under a severe accident of a nuclear power plant, and the method comprises the steps: S1, building a radioactive accident risk evaluation basic model architecture, adapting to a model demand, and building a corresponding basic database; s2, constructing radioactive smoke plume characteristics after the accident; s3, calculating accident site space gridding and smoke plume transmission discretization; s4, sampling meteorological conditions, and simulating smoke plume transmission; s5, calculating the radiation dose in the early stage of the grid domain, and carrying out risk assessment; s6, performing risk assessment on the health effect in the early stage of the grid domain; s7, calculating the medium-and-long-term stage radiation dose in the grid domain, and performing dose risk assessment; and S8, performing health effect risk assessment and accident consequence assessment in the medium-and-long-term stage. According to the method, the whole transportation process of the radioactive smoke plume under a serious accident is simulated, loss effects such as dry and wet settlement and radioactive decay of the smoke plume are considered, and the activity of the radionuclide deposited on the ground and retained in the air in the off-site space range is determined.

The invention relates to the technical field of medical instruments, and discloses a breast cancer sentinel lymph node positioning system and method, and the system comprises a data collection subsystem, a space positioning subsystem and a central processing subsystem. The method comprises the following steps: acquiring an anatomical structure image, a radioactivity counting rate and fluorescence intensity after performing primary fluorescence, radioactivity and secondary pulse type fluorescence tracer injection in sequence; establishing a global coordinate system and generating an anatomical region of interest; constructing a weighted directed topological graph based on the initial fluorescence data, and obtaining a baseline flow velocity; extracting radionuclide dynamic characteristics based on the radioactive data; a functional flow rate attenuation coefficient is calculated based on the second pulsed fluorescence data. And a multi-factor scoring model is applied, topological information, dynamic characteristics and functional flow velocity attenuation coefficients are fused, and a comprehensive score is calculated to identify the sentinel lymph node. According to the method, the lymphatic flow topological structure, nuclide dynamics and functional flow velocity information are combined, and the positioning accuracy and reliability are improved.

The invention belongs to the field of wastewater treatment technology and energy conservation and environmental protection, and particularly relates to a preparation method and application of aerogel for selectively separating radionuclide cesium in water. The preparation method comprises the following steps: preparing graphene oxide by using a hydrothermal technology, preparing molybdenum disulfide / GO aerogel by using a hydrothermal synthesis and ultrasonic technology, and inlaying and growing Prussian blue (PB) on the surface of the aerogel, so that the problems that PB and MoS2 are easy to agglomerate and difficult to recover are solved, the GO aerogel is endowed with multiple functions, and the constructed PB / MoS2 / GO aerogel is used for selectively separating radionuclide cesium in water. The selective separation performance of the aerogel on the radionuclide cesium in water under different conditions is investigated, efficient enrichment and recovery of the radionuclide cesium are effectively realized by utilizing the ultralight and highly porous characteristics of the aerogel, and the application potential of the aerogel in the aspects of reducing secondary pollution and improving the environmental treatment effect is proved. An efficient and environment-friendly solution is provided for radionuclide wastewater treatment, and the method has a wide market application prospect.

A detector system for molecular imaging of a radionuclide comprises a 3D semiconductor detector comprising a plurality of sensor stacks of sensors made of a semiconductor material having an average atomic number Z below 40. A read-out circuitry connected to the pixels is configured to output, for each interaction induced by an incident gamma ray in the detector, a signal representative of a time, a position and an energy of the interaction in the detector. The interactions in the detector belonging to a same event induced by the incident gamma ray are predicted based on the output signals and used to estimate a direction of the incident gamma ray and reconstruct an image based on the estimated directions of incident gamma rays.

A method of generating an emulsion for treating a patient. The method includes generating a water-based solution carrying an agent which turns into a hydrogel by addition of calcium ions. Radium radionuclides are added to the water-based solution in a concentration sufficient for radiotherapy. After adding the radium to the water-based solution, the water-based solution is mixed with an emulsifier and an oil.

The invention is suitable for the technical field of gamma energy spectrum correction, and provides a lanthanum bromide detector gamma energy spectrum correction method based on LSTM, and the method comprises the following steps: carrying out the data preparation and pretreatment of a lanthanum bromide detector gamma energy spectrometer; constructing an LSTM network model, and training and testing the LSTM network model; and importing unknown actually measured original energy spectrum data of the lanthanum bromide detector gamma spectrometer into the trained LSTM network model, and predicting corresponding nuclide energy spectrum data. The problems of energy spectrum background, spectral line drifting, noise and the like existing in data collected by the lanthanum bromide detector gamma spectrometer are effectively solved. According to the invention, the method achieves the quick and stable correction of the gamma energy spectrum through the memory and processing capability of the LSTM network, can quickly obtain the nuclide spectral line used for determining the characteristic peak and peak area of the radionuclide, remarkably reduces the interference of human factors in the processing of energy spectrum data, and improves the accuracy and efficiency of data processing.

The invention discloses an energy spectrum analysis method based on sample migration, and relates to the technical field of nuclear radiation detection, and the method comprises the following steps: taking a large amount of simulated radionuclide gamma energy spectrum data obtained through a Monte Carlo numerical simulation method as a source domain; taking marked actually measured radionuclide gamma energy spectrum data obtained by measuring a radioactive source through a detector as a target domain; source domain and target domain samples with heterogeneous data distribution are mapped to a shared subspace with discriminative feature representation through feature alignment; aligning samples of a source domain and a target domain with tag space isomerism through tag alignment; and constructing a classification model of the low-count energy spectrum through tree decision. According to the method, effective expansion of the energy spectrum number is realized through data distribution and label space heterogeneous simulated energy spectrum data, and the strong dependence of a classification model on a large amount of actually measured annotated data can be reduced.

The invention discloses a nuclear accident external irradiation dose field calculation method based on a limited smoke cloud model and a computer readable storage medium, and solves the technical problems of poor matching performance, relatively long calculation time, relatively low efficiency and the like in the prior art. The method comprises the following steps: firstly, acquiring a radionuclide source item file released by a nuclear accident and corresponding original data; calculating to obtain a source item nuclide information data matrix; according to the gamma ray energy of each nuclide, calculating to obtain a gamma ray linear attenuation coefficient and an air energy mass absorption coefficient of each nuclide; calculating to obtain the grid number and the grid number of each grid penetrated by gamma rays of other grids in the smoke cloud activity concentration field and the ground deposition activity field; finally, the deposition external irradiation dose and the smoke cloud external irradiation dose of each grid are obtained through calculation, and therefore nuclear accident external irradiation dose field calculation is completed. According to the method, automatic and refined calculation of the external irradiation dose during the nuclear accident can be realized, and the calculation efficiency is relatively high.

The invention provides a spectrum stabilization method of a nuclear energy spectrum and electronic equipment, and the method comprises the steps: determining a channel address feature value corresponding to each channel address in an energy spectrum diagram; determining a target reference source characteristic peak according to the plurality of channel address characteristic values; determining a reference position with the maximum slope in a falling edge region on the right side of the target reference source characteristic peak; according to a predetermined proportionality coefficient and the reference position, determining a central position of a target reference source characteristic peak; wherein the predetermined proportionality coefficient is the ratio of the reference distance to the full width at half maximum; the reference distance is the distance between the center position of the reference source characteristic peak and the reference position; and based on the central position of the characteristic peak of the target reference source, correcting the energy spectrum diagram to obtain a spectrum-stabilized energy spectrum. According to the technical scheme of the embodiment of the invention, the central position of the characteristic peak of the target reference source can be accurately determined, the influence of statistical fluctuation and the interference of characteristic peaks of other radionuclides are avoided, and the stability of a spectrum stabilization result is ensured.

The present invention relates to MOFs for radiotherapy. More specifically, the present invention provides: a particle comprising an MOF, optionally at least one targeting moiety, and at least one beta-emitting radionuclide; a composition comprising the particles; said particles or composition for use as a medicament; said particles or compositions for use in a method of treating a proliferative disease; and a kit comprising a particle comprising an MOF, optionally a targeting moiety, and a beta-emitting radionuclide.

The invention provides a radioactive internal pollution detection method and system based on a dynamic efficiency scale, and the method comprises the steps: carrying out the detection of the radioactive internal pollution based on a pre-obtained change table of the total detection efficiency of a detector for the radionuclide in a human body along with the time; determining the total detection efficiency of the radiation detector on the radionuclides in the human body according to the interval time and the types of the radionuclides; and obtaining the activity of the internal pollution radionuclides based on the characteristic gamma ray full-energy peak counting and the total detection efficiency of the radiation detector on the radionuclides in the human body. In-vivo metabolism behaviors of nuclides are considered, so that the internal pollution detection accuracy is greatly improved. The problems that in the prior art, equipment is large in size, high in manufacturing cost, long in measurement time consumption, complex in operation and the like are solved. The method is suitable for rapid detection and classification of mass personnel in nuclear and radiation accident emergency sites.

The present application relates to radionuclide separation and purification technical field, specifically disclose a kind of separation and purification method for preparing lead-212 and actinium-225 by photo nuclear reaction method.It includes the following steps: radium target solution is sequentially passed through lead resin column and DGA resin column, and recovery liquid is obtained;First eluent is used to elute lead resin column, and lead-212 first purification liquid is obtained;Second eluent is used to elute DGA resin column, and actinium-225 first purification liquid is obtained;Lead-212 first purification liquid is introduced into first LN resin column, and third eluent is used to elute first LN resin column, then third desorption liquid is introduced into first LN resin column, and lead-212 second purification liquid is obtained;Actinium-225 first purification liquid is introduced into second LN resin column, and fourth eluent is used to elute second LN resin column, then fourth desorption liquid is introduced into second LN resin column, and actinium-225 second purification product solution is obtained.The present application effectively avoids the cumbersome operation steps such as medium conversion, improves the separation and purification efficiency of lead-212 and actinium-225 in radium target.

In a sodium-cooled fast reactor, a breached fuel pin releases fission and activation products into the primary sodium coolant and into the cover gas. A portion of the primary sodium coolant is diverted to a distillation system where its constituents can be separated based on volatility. Condensing these constituents enables precise measurement of their concentration without delays caused by waiting for the decay of radioactive sodium background scatter, enhancing the detection and quantification of the dilute constituents. Additionally, the method continuously determines the initial concentration of constituents in the feed. Quantification is achieved using detectors, facilitating the continuous assessment of dilute constituent concentrations in the sodium coolant feed stream.

The application discloses a rare earth nanoparticle-organic photosensitizer composite material and application thereof. By introducing rare earth doped nanoparticles as an energy conversion medium, direct activation of excited triplet states of an organic photosensitizer and sensitized production of singlet oxygen under excitation of radionuclides (e.g., 64Cu, 67Cu, 90Y, 111In, 177Lu, etc.) are realized. 18 F、 99m Tc、 177 Lu, etc.) excitation, the rare earth doped nanoparticles respond to CL and the rare earth ions reach an excited state after various high-energy radiation. Subsequently, through an interface energy transfer process, the excited state energy of the rare earth ions is directly transferred to the excited triplet state of the organic photosensitizer, and then singlet oxygen is generated, realizing tumor cell killing. The nanocomposite material is surface modified to load iRGD peptides to realize in vivo tumor targeting. 18 F-FDG (fluorodeoxyglucose) can realize tumor targeting through abnormal glucose uptake of tumors. The combination of the two realizes double targeting of in vivo tumors, and enhances the anti-tumor efficacy and safety.

The invention relates to the field of radioactive waste treatment, in particular to a plasma treatment device and method for removing nitrogen oxides and adsorbing radionuclides. According to the device, a plasma furnace is connected with a secondary combustion chamber through a smoke exhaust pipe; a denitration furnace, a heat exchanger, a quench tower, a dust removal cloth bag, a washing tower, a flue gas cooler, a gas-liquid separator and an adsorber are connected behind the secondary combustion chamber; at least two adsorbers are arranged in parallel; one outlet of each adsorber is connected to the electric heater and the induced draft fan; and the other outlet of the adsorber is respectively connected to the plasma furnace and the smoke exhaust pipe. The method comprises the steps that the radioactive waste is incinerated; carrying out melting treatment on inorganic matters generated by incineration treatment; gas generated by incineration treatment enters a secondary combustion chamber to be subjected to secondary plasma incineration treatment; the generated flue gas is subjected to denitration treatment, heat exchange cooling treatment, extremely cold cooling treatment, dust removal treatment and at least two-stage washing treatment; and carrying out low-temperature cryogenic removal treatment, and discharging after heating. According to the invention, zero emission of nitrogen oxides is realized.

The invention discloses a radiation hazard source identification method with a low counting rate, and relates to the technical field of nuclear technology application. Comprising the following steps: establishing a nuclide library containing radionuclide information; setting a peak position interval corresponding to each radionuclide based on the nuclide library; carrying out statistics on the peak position count of the peak position interval and the high-energy area count of the high-energy area, and caching the peak position count and the high-energy area count obtained through statistics as environment background data; estimating parameters according to the cached environment background data; acquiring a peak position counting rate and a high-energy region counting rate in real time; according to the parameters, the peak position counting rate and the high-energy region counting rate which are obtained in real time, the probability that the peak region is abnormal is calculated; performing early warning sequential test on the abnormal probability of the peak region; and comparing a result of the early warning sequential test with a preset early warning threshold value, and if the result reaches the early warning threshold value, outputting early warning information and identifying a corresponding radiation hazard source. According to the invention, rapid early warning and hazard source identification of abnormal radiation data are realized.

The invention relates to a method (10) for characterizing at least two radionuclides mixed in a radioactive fluid, the radionuclides being substantially pure beta emitters, the method (10) being carried out by a device (1) calibrated by an average number of photons detected per decay and a decay detection yield value for each radionuclide, the method comprising: measuring (14) an average number of photons produced per decay for the radioactive fluid, determining (16) a number of decays for each radionuclide from the average number of photons detected for each radionuclide and the average number of photons detected for the radioactive fluid, and determining (18) the activity of each radionuclide from the number of decays and the calibration value for each radionuclide. [Fig. 1]

Methods and apparatus are disclosed for concurrent digital autoradiography of alpha-particle events and positron events using a spatially resolving charged particle detector and a gamma-ray detector. Alpha particles and positrons are detected by the charged particle detector. Positrons are identified based on coincidence with gamma-ray events. A positron-emission autoradiograph is formed based on positron positions. Alpha particles, both coincident and anticoincident with positrons, are identified based on energy deposition patterns. An alpha-emission autoradiograph is formed based on alpha-particle positions and energies. Separation of coincident alpha and positron events recorded by the charged particle detector improves position detection accuracy. Validation of positron imaging for alpha-emitter distribution or dosage, based on correlation of respective autoradiographs, is described. Variations for particle identification based on isotropy, energy, coincidence, or anticoincidence are presented. Disclosed techniques are applicable to alpha-emitting radionuclides with complex decay chains.

Disclosed herein are methods for radiotherapy treatment plan optimization for irradiating one or more target regions using both an internal therapeutic radiation source (ITRS) and an external therapeutic radiation source (ETRS). One variation of a method comprises iterating through ITRS radiation dose values and ETRS radiation dose values to attain a cumulative dose that meets prescribed dose requirements. In some variations, an ITRS is an injectable compound that has a targeting backbone and a radionuclide, and images acquired using an imaging compound that has the same targeting backbone as the injectable compound can be used to calculate the radiation dose deliverable using the injectable ITRS, and also to calculate firing filters for delivering radiation using a biologically-guided radiation therapy (BGRT) system. Image data acquired from a previous treatment session may be used to adapt the dose provided by an ITRS and / or ETRS for a future treatment session.

An image acquisition device 1A comprises a measurement unit 10 and a processing unit 20. For each coincidence event in which a pair of gamma-ray photons generated through an electron-positron pair annihilation event in a positron emitting radionuclide 81 are detected in coincidence by a first detector 11 and a second detector 12, the processing unit 20 determines a position where a gamma-ray photon has been Compton-scattered in a subject 90 selectively if the gamma-ray photon has reached the second detector 12 after being Compton-scattered in the subject 90 and the energy of the gamma-ray photon represented by a signal outputted from the second detector 12 is lower than a first threshold value lower than 511 keV, on the basis of a position and a time of the gamma-ray photon detection by each of the first detector 11 and the second detector 12, and the position of the positron emitting radionuclide 81. Thus, an image acquisition device and an image acquisition method are achieved that make it possible to acquire a tomographic image representing anatomical information pertaining to a subject without performing image reconstruction processing.

A detector system (1) for molecular imaging of a radionuclide comprises a 3D semiconductor detector (100) comprising a plurality of sensor stacks (110) of sensors (120) made of a semiconductor material having an average atomic number Z below 40 A read-out circuitry (!26) connected to the pixels (125) is configured to output, for each interaction induced by an incident gamma ray in the detector (100), a signal representative of a time, a position and an energy of the interaction in the detector (100). The interactions in the detector (100) belonging to a same event induced by the incident gamma ray are predicted based on the output signals and used to estimate a direction of the incident gamma ray and reconstruct an image based on the estimated directions of incident gamma rays.

Disclosed are compounds represented by the following structural formula (I): or a pharmaceutically acceptable salt thereof. The variables of structural formula (I) are described herein. The compound of the invention can be attached to chelating groups for radionuclide binding and are therefore suitable for radioimaging and / or radiotherapy applications, for example the disclosed compounds can be radiolabeled with a positron emitter such as 18F, 68Ga or 64Cu, and used for positron emission tomography (PET). Alternatively, the compounds can be radiolabeled with an alpha particle emitter such as 223 Ac, a beta particle emitter such as 67Cu or 177Lu, or an Auger electron emitter (e.g. 111In, 67Ga, 99mTc, 195mPt, 125I, 123I and 161Tb) for use as a radiotherapeutic.

The invention discloses a double-target small-molecule probe, a radionuclide-labeled double-target small-molecule probe as well as a preparation method and application of the radionuclide-labeled double-target small-molecule probe. The medicine < 68 > Ga-DOTA-PRE / RGD is obtained by labeling < 68 > Ga radionuclide, and is applied to PET / MR (positron emission tomography / magnetic resonance imaging) imaging of pancreatic cancer tumors. A radioactive product with high purity and high stability (greater than 93%) is obtained through a simple marking method; the obtained double-target radioactive small molecular probe 68Ga-DOTA-PRE / RGD has a huge potential for the accurate diagnosis of pancreatic cancer.

The invention discloses a radionuclide identification system and method, and the system comprises a sample placement module which is used for placing a radionuclide sample; the absorption sheet module is positioned above the sample placement module, is provided with a plurality of absorption sheets with different models, and is used for weakening absorption of rays emitted by nuclides; the acquisition module is positioned above the absorption sheet placing module and is used for acquiring information of rays passing through the absorption sheet; the processing module comprises a control unit and an analysis unit, the control unit is used for controlling the absorption piece module to switch the absorption pieces, and the analysis unit is connected with the acquisition module and used for analyzing the ray information; according to the invention, the limitation of traditional equipment in the aspect of switching the absorption sheets is overcome, the absorption sheets are automatically switched by adopting an automatic control mode, and the experiment efficiency can be improved.

A method of generating an emulsion for treating a patient. The method includes generating a water-based solution carrying an agent which turns into a hydrogel by addition of calcium ions. Radium radionuclides are added to the water-based solution in a concentration sufficient for radiotherapy. After adding the radium to the water-based solution, the water-based solution is mixed with an emulsifier and an oil.

The invention discloses a radioactive inert gas detection device which comprises a gas chamber, a plastic flash detection assembly used for detecting the total beta activity concentration of radioactive inert gas, and a lanthanum bromide crystal detection assembly used for detecting the types of radionuclides in the radioactive inert gas and the activity concentration of the radionuclides. The plastic flashover detection assembly comprises a first plastic flashover detector, a second plastic flashover detector, a third plastic flashover detector and a first signal processing circuit board. The bottom of the third plastic flashover detector is provided with a lead shielding plate. The lanthanum bromide crystal detection assembly comprises a first lanthanum bromide crystal detector, a second lanthanum bromide crystal detector and a second signal processing circuit board; lead shielding plates are arranged on the first lanthanum bromide crystal detector, the second lanthanum bromide crystal detector and the tops of the first lanthanum bromide crystal detector and the second lanthanum bromide crystal detector. The device can overcome environmental background interference, improves the detection efficiency, and can accurately measure the total beta activity concentration of the radioactive inert gas in the exhaust pipeline and the types and activity concentrations of all radionuclides under the condition that the activity concentration of the radioactive inert gas is low.

An image capture device 1A comprises a measuring unit 10 and a processing unit 20. The processing unit 20 determines, for each coincidence event in which a first detector 11 and a second detector 12 perform coincidence detection of a pair of gamma-ray photons generated by an electron-positron annihilation event in a positron-emitting radionuclide 81, a position at which the gamma-ray photon is Compton-scattered in the object 90, assuming that the gamma-ray photon arriving at one detector of the first detector 11 and the second detector 12 is a gamma-ray photon arriving without Compton scattering in an object 90, and the gamma-ray photon arriving at another detector is a gamma-ray photon arriving after being Compton-scattered in the object 90.based on a detection position and a detection time of the gamma-ray photon by each of the first detector 11 and the second detector 12 and a position of the positron-emitting radionuclide 81. In this way, an image acquisition device and an image acquisition method capable of acquiring a tomographic image representing anatomical information of an object without performing image reconstruction processing are realized.

The invention relates to wet preparation of a radiation therapy source and a method for accumulating radium radionuclides. The method comprises providing a first solution comprising a thorium radionuclide and a thorium binding extractant, wherein the first solution does not bind to radium; allowing a portion of the thorium radionuclides in the first solution to decay to radium atoms; and collecting radium atoms resulting from the decay. The collected radium atoms may be contained in a solution into which a brachytherapy source is immersed in a manner that collects radium atoms onto the brachytherapy source.