73 results about "Radioactive Activity" patented technology

The invention provides an online monitoring device for multiple nuclides in liquid effluent, and particularly relates to the technical field of radioactive substance detection. The online monitoring device comprises a shell, a cascade separation module, a fluid control module, a detection module and a control system, the cascade separation module, the detection module and the control system are arranged in the shell, the cascade separation module comprises a plurality of resin columns and a plurality of separated liquid collecting bottles communicated with the resin columns, and the separated liquid collecting bottles are communicated with the resin columns. The fluid control module comprises a liquid storage device, an injection pump and a control valve, the liquid storage device is respectively communicated with the plurality of resin columns through pipelines, the injection pump and the control valve are respectively arranged on the pipelines and are used for controlling the flow direction and the flow velocity of fluid in the liquid storage device, and the detection module is used for detecting the radioactive activity of nuclide; the control system is used for controlling operation of the cascade separation module, the fluid control module and the detection module. According to the invention, synchronous monitoring of multiple nuclides is realized, the multi-nuclide monitoring efficiency is improved, and the monitoring cost is reduced.

The invention discloses a dynamic calculation method and system for a radiation field in the cutting process of an activity non-uniform radioactive source, and belongs to the field of radiation field calculation. The method comprises the following steps: acquiring three-dimensional activity distribution data of a radioactive source; constructing a voxelization model and coupling the voxelization model with an activity grid to distribute activity; establishing an LBVH hierarchical structure of the shielding body; a point kernel is generated in a self-adaptive mode through ray tracing and average free path comparison; carrying out radiation field calculation by adopting point nuclear integration, and realizing multilayer ray tracing in the radioactive source, between the sources and a shield in the process; on-line cutting and moving operation of the voxel model is supported, and the radiation field is dynamically updated. According to the method, the cutting operation of the activity non-uniform radioactive source can be simulated, the problem that the radioactive activity of each cutting product cannot be represented after cutting by a traditional method is solved, the radiation field dynamic calculation in the activity non-uniform radioactive source cutting process can be simulated, initial modeling of the radiation environment only needs to be carried out once, and the efficiency is improved. And repeated modeling is not needed.

The invention relates to the cross technical field of nuclear reactor engineering and corrosion engineering, in particular to a radioactive isotope tracing-based fuel cladding tube surface scale on-line measurement device, which comprises an injection control module for injecting a radioactive 59Fe solution into a coolant and adjusting and controlling the 59Fe content by analyzing the Fe ion concentration; the test module is used for testing the surface of the fuel cladding based on the coolant injected with the radioactive 59Fe solution; and the measuring module is used for collecting the radioactive activity of 59Fe in scale generated on the surface of the fuel cladding after the test and speculating the scale mass surface density. According to the method, the content of the 59Fe in the coolant is controlled, and the radioactive activity of the 59Fe in the scale deposit on the surface of the fuel cladding is detected on line, so that the mass of the corresponding 59Fe can be accurately calculated, the scale deposit mass is speculated, the scale deposit information on the surface of a cladding pipe is obtained in real time, and the dynamic change condition of the scale deposit is reflected in time.

Disclosed is a system and a method enabling the behavior of a radioactive substance within a structure to be determined. More particularly, the disclosed is a system and a method wherein it is possible to determine, along a column, the position and / or distribution and / or movement speed of the substances moving within the column and / or the time when the substances will leave the column and how much of the substance reaching the column outlet will begin to leave the column and with what activity, and consequently, it is possible to provide optimization and automation for proceeding to the next process step.

The invention relates to a 14C urea pharmaceutical composition for helicobacter pylori diagnosis and application of the 14C urea pharmaceutical composition. The 14C urea pharmaceutical composition comprises 14C urea, 12C urea, citric acid, malic acid and mannitol. According to the invention, the 12C urea and the 14C urea are added together, so that the radioactivity can be ensured, the stability of the medicine can be improved, and the auto-radiolysis of 14C can be reduced. Besides, citric acid, malic acid and mannitol have a remarkable synergistic effect in the aspect of improving the radioactive activity and radiochemical purity of the composition, the stability of the 14C urea pharmaceutical composition can be improved, meanwhile, the activity of urease generated by helicobacter pylori can be improved, and the detection sensitivity and detection efficiency are further guaranteed.

The present disclosure relates to a method, device, storage medium and electronic device for measuring radioactivity, the method comprising: obtaining an absorption factor of a shielding medium corresponding to each of a plurality of target radioactive objects in a target container and an absorption factor of the target container, the absorption factors representing absorption efficiencies of radioactive rays of the target radioactive objects; and determining radioactivity of the target radioactive objects according to the absorption factors of the shielding medium and the absorption factor of the target container. In this way, the radioactivity of the target radioactive objects is determined according to the absorption factors of the shielding medium and the absorption factor of the target container, thereby avoiding the problem of poor accuracy of the analysis result of the radioactivity due to non-uniform radioactivity of the plurality of target radioactive objects, and effectively improving the accuracy of the radioactivity measurement result when the plurality of target radioactive objects are non-uniform radioactive sources.

The present application relates to a kind of natural conditions under the estimation method of superuranic nuclide resuspension release source term.This application obtains the environmental characteristic parameter of the place of contaminated area, calculates the related parameter of superuranic nuclide pollution condition, calculates the surface covering layer factor of contaminated area soil, calculates the resuspension release source term of superuranic nuclide, can start from the characteristics of soil environment, nuclide characteristics, natural condition three aspects, based on the mass load theory of particulate matter, using the release flux of particulate matter in contaminated area soil and the level of radioactivity concentration in soil, the resuspension release source term of superuranic nuclide is calculated, realizes the resuspension release amount of superuranic nuclide in soil environment pollution under natural conditions is scientifically and accurately evaluated.

The utility model provides a gas purification system of a reactor. The gas purification system comprises a purification pipeline, and an inert gas adsorber, an iodine adsorber, a polonium processor, a filter and a radioactivity analyzer which are arranged on the purification pipeline. And the inert gas adsorber is used for removing inert gas in the mixed gas. The iodine adsorber is used for removing iodine in the mixed gas. The polonium processor is used for removing polonium in the mixed gas. And the filter is used for removing small particles in the mixed gas. And the radioactivity analyzer is used for detecting the radioactivity of the mixed gas. According to the gas purification system, aiming at the characteristics of radioactive substances of the metal fast reactor, a four-stage purification treatment device is arranged to remove radioactive gas in the mixed gas, and in addition, a radioactivity analyzer is further arranged and used for detecting and judging whether the radioactivity in the gas meets the emission standard or not; and the problem of radioactive substance release caused by leakage of the containment is effectively avoided.

The invention relates to the technical field of terminal sterilization and filtration in a radiopharmaceutical preparation process, in particular to a filter membrane sterilization and filtration device and system. Comprising the steps that a state sensing module collects liquid medicine radioactive activity data and filter membrane structure strain data and generates real-time data flow; the twinborn modeling module constructs a filter membrane digital twinborn body, and outputs an accumulated radiation damage prediction result through data assimilation; the prediction decision module generates a filtering parameter optimization sequence by using a deep reinforcement learning agent; the control execution module adjusts the flow velocity of the liquid medicine through fuzzy self-adaptive PID control and feed-forward compensation; and the verification feedback module performs an integrity test and transmits data back to the twin modeling module to realize model parameter correction. All the modules form a closed-loop control system through a high-speed data bus, so that the damage of radiation to a filter membrane material is effectively inhibited, and the integrity and the sterile guarantee level of a membrane structure in the filtering process are maintained.

A method and an apparatus for producing radionuclide-containing products having substantially the identical desired activity of radioactivity at different times of application, with respect to a given calibration time. The method according to the invention allows to ensure a consistent composition of the desired radionuclide-labeled pharmaceutical at all times of application within the shelf life through a single manufacturing process. With the present invention it is possible, for example, to provide [n.c.a. Lu-177]Lu-DOTATOC on each working day of the week with constant activity at the respective time of application in a single manufacturing step.

The present disclosure relates to a radioactive waste barrel activity measuring device, comprising a chassis and a conveying mechanism arranged on the chassis, the bottom of the chassis is connected with walking wheels, a dose measuring mechanism and a detector assembly are arranged on one side of the conveying mechanism on the chassis, the detector assembly is movably arranged on the chassis to be able to approach or move away from the conveying mechanism, a control system is further arranged on the chassis, and the control system is electrically connected with the dose measuring mechanism and the detector assembly respectively. In this way, the device can conveniently adjust the waste barrel and the radioactive activity measuring device for radioactive measurement, and move to the required measurement site according to the requirement. Through the control system, all running actions of the whole measuring device can be controlled, and the data collected by the dose measuring mechanism and the detector assembly are collected, the efficiency of the activity measurement of the radioactive waste barrel is improved, and the measurement operation of the low-activity radioactive waste barrel is conveniently and quickly completed.

The utility model discloses a movable radioactive iodine continuous monitor and detection device, and belongs to the technical field of nuclear facilities and air detection. According to the movable radioactive iodine continuous monitor, the reliability and the accuracy of detection data are improved by arranging the detection device, the detector can conveniently monitor at different places by designing the main wheels, and the working efficiency is improved. According to the detection device, the size of the iodine box is set to be consistent with the size of the probe of the NaI detector, so that the NaI detector can monitor radioactive iodine in air more comprehensively, the reliability of a monitoring result is improved, and resource waste caused by the fact that the arranged probe is too large is avoided. By using the lead chamber, external interference is effectively shielded, and the accuracy of monitoring data is improved. And more accurate radioactive activity concentration data can be provided through a multichannel analyzer. The problem that the monitoring result of an existing monitor is unreliable due to the defect of incomplete monitoring is solved.

The present invention relates to the field of nuclear medicine technology, and discloses a targeted compound bimolecular probe, comprising a component A and a component B respectively labeled with radioactive metal nuclides; component A is a molecular probe targeting fibroblast activation protein (FAP), and the molecular probe targeting fibroblast activation protein (FAP) includes a trans-cyclooctene group (TCO); component B is a molecular probe targeting integrin α v β6 molecular probe, targeting integrin α v The β6 molecular probe includes a tetrazine group (Tz); wherein the radioactivity ratio of component A to component B is (1-2): (2-1). The present invention has the following technical effects: the TCO in the FAP-targeting molecular probe and the integrin α v The Tz in the β6 molecular probe can undergo a bioorthogonal reaction, achieving a tertiary amplification effect, increasing tumor uptake, prolonging the retention time of the targeted compound bimolecular probe, improving the target / non-target ratio, and enhancing nuclear medicine diagnostic and therapeutic effects. The present invention also discloses a method for preparing and using the targeted compound bimolecular probe.

The application relates to a kind of measuring device for measuring the alpha radioactivity activity concentration in spent fuel wastewater, comprising a support base, the top of the support base is provided with a liquid receiving tray, the outer side of one end of the liquid receiving tray is hung with a liquid inlet pump and a liquid outlet pump at intervals, and the outer side of the other end is hung with a sealed junction box; the liquid inlet pump is also connected with a constant liquid level pre-tank in communication through a pipeline; a support frame is arranged in the inner cavity of the liquid receiving tray, the top inner side of the support frame is fixedly connected with a liquid storage pool, the liquid storage pool is respectively connected with a first liquid level meter, a second liquid level meter and a third liquid level meter, and the bottom of the liquid storage pool is also connected with the liquid outlet pump in communication through a pipeline; the top outer side of the support frame is fixedly connected with a measuring pool, and the bottom of the measuring pool is connected with the top of the liquid storage pool in communication; the top of the measuring pool is provided with a film winding driving mechanism, a film is arranged on the film winding driving mechanism, and a detector is arranged on the top of the film winding driving mechanism. The application can effectively solve various problems of offline measurement, and realize online measurement of alpha activity concentration in high-concentration acidic liquid.

A method and an apparatus for producing radionuclide-containing products having substantially the identical desired activity of radioactivity at different times of application, with respect to a given calibration time. The method according to the invention allows to ensure a consistent composition of the desired radionuclide-labeled pharmaceutical at all times of application within the shelf life through a single manufacturing process. With the present invention it is possible, for example, to provide [n.c.a. Lu-177]Lu-DOTATOC on each working day of the week with constant activity at the respective time of application in a single manufacturing step.

The utility model relates to an improved radioactive activity counting and lifting bucket which comprises a handheld transverse partition plate, a middle partition plate, a bottom support and a connecting rod, the handheld transverse partition plate is installed at the upper end of the connecting rod, the bottom support is installed at the lower end of the connecting rod, and the middle partition plate is installed in the middle of the connecting rod and located between the handheld transverse partition plate and the bottom support. The placing hole is positioned right above the bottom support; the improved radioactive activity counting and lifting bucket further comprises an injector positioning sleeve, the injector positioning sleeve comprises a supporting plate and a positioning sleeve, the upper end of the positioning sleeve is connected with the supporting plate, and a positioning opening communicated with the upper end of the positioning sleeve is formed in the supporting plate. The supporting plate is arranged on the middle partition plate, and the positioning sleeves are arranged on the inner sides of the containing holes in a sleeving mode. According to the improved radioactive activity counting and lifting bucket, the position of an injector containing radioactive liquid medicine on the middle partition plate can be more stable, the situation that the injector shakes or inclines is prevented, the medicine dosage measurement error can be effectively reduced, and radioactive medicine pollution is prevented.

The present invention belongs to the field of nuclear radiation protection technology and provides a radioactive gas detection device based on a silicon drift detector. The device comprises a silicon drift detector and a sampling chamber. The sampling chamber has a circular hole at the top. The silicon drift detector is mounted above the sampling chamber with a beryllium window facing the circular hole. The gas to be measured enters the sampling chamber through an inlet pipe, and the gas in the chamber flows out through an outlet pipe. The outlet pipe is arranged offset from the inlet pipe. A preamplifier circuit board and a signal processing board are mounted above the silicon drift detector via a fixed holder. The silicon drift detector, preamplifier circuit board, and signal processing board are disposed within a detector housing, which is fixedly connected to the sampling chamber. A connector is provided at the top of the detector housing. The device can simultaneously calculate radioactive activity and perform nuclide identification. The device uses simulated data and real X-ray data to calculate gamma background counts. The corresponding functions can be achieved using only one detector, greatly improving detector utilization.

The invention provides a radioactive activity meter which comprises a shell, the shell is of a composite structure, a shielding cylinder is detachably arranged in the shell, a containing cavity used for containing a glass container of radioactive medicine is formed in the shielding cylinder, a positioning clamping piece is arranged on the upper portion of the shielding cylinder, and a fixing frame is arranged on the outer side of the shell. An installation assembly is arranged on the fixing frame and comprises a connecting piece matched with the positioning clamping piece. According to the invention, rapid and efficient installation and positioning operation of the shielding cylinder can be realized through cooperation of the installation assembly and the positioning clamping piece, and efficient detection operation of particle activity of radiopharmaceuticals is greatly improved.

The present application relates to the technical field of system optimization, and discloses a method and system for performance collaborative optimization of multiple sets of separation systems based on federated learning, which comprises the following steps: obtaining the radioactivity value of the current batch of raw material liquid of each set of At-211 separation system, analyzing the radioactivity value based on a deep neural network, and generating a working condition sensitive feature vector representing the influence of different batches of radioactivity value on the performance of each separation system; through the working condition sensitive feature vectors, working condition attribute factors, activity working condition vectors and collaborative optimization coefficients generated in each step, the influence of different batches of radioactivity value can be accurately mined, the correlation and difference of each system working condition can be determined, the fine adjustment of differentiated aggregation weight can be realized, the distribution out-generalization failure of cross-activity migration can be avoided, the process parameters can be ensured to adapt to the demand of different activity raw material liquid, the collaborative ability of multiple sets of At-211 separation system can be improved, and the separation efficiency and product purity can be improved.

The invention relates to the technical field of radioactive source preparation, in particular to a preparation process of a 137Cs source for a blood irradiator. The process comprises the following steps: step 1, preparing raw materials: using 137Cs product liquid extracted from high-level liquid waste, wherein the radioactive activity of the 137Cs product liquid meets the basic requirements of a cesium source for a blood irradiator; step 2, cesium source pellet pressing: loading a 137Cs raw material into a mold, pressing a pellet in a hydraulic machine, and then drying and sintering at high temperature to obtain the cesium source pellet; and step 3, welding and sealing: the cesium source core block is arranged in a stainless steel source box, and the source box and the sleeve are sealed through an argon arc welding technology. The invention provides a process for purifying and transforming 137Cs product liquid extracted based on high-level liquid waste. A prepared radioactive source has the characteristics of high uniformity, high radioactive activity, small size, high difficulty, high sealing performance and the like; meanwhile, the 137Cs product liquid is successfully converted into the radioactive source meeting the standard of the blood irradiator, and autonomous preparation of the 137Cs source for the domestic blood irradiator is achieved.

The application discloses a method for estimating radioactivity concentration of desalinated water of a nuclear energy seawater desalination device, which comprises the following steps: firstly, according to the technological process of the nuclear energy thermal method seawater desalination device, the flowing direction of the medium and the migration direction of the radioactive substances are sorted out; then, according to the approach of the radioactive substances entering and removing in the intermediate loop fluid, the nuclide balance equation of the intermediate loop fluid is established, and then the radioactivity concentration calculation model of the intermediate loop water and the radioactivity concentration calculation model of the flash steam of the flash tank are obtained; then, according to the technological scheme of the MED device, the radioactivity concentration calculation model of the steam and the condensed water generated by each heat exchanger is established; then, the radioactivity level of the steam and the condensed water generated by each heat exchanger of the MED is calculated; finally, whether the desalinated water can meet the radioactivity control index is evaluated. The method can accurately calculate the radioactivity activity concentration of the desalinated water, and provides a reliable basis for the safety evaluation of the nuclear energy thermal method seawater desalination device.

The embodiment of the application provides a multifunctional food and water radioactivity detection device, wherein the multifunctional food and water radioactivity detection device has a radioactivity detection mode, and the multifunctional food and water radioactivity detection device comprises: a shielding main body; a Marin cup arranged in the shielding main body, the Marin cup being used for storing a sample to be detected; and a detection device, in the radioactivity detection mode, the detection device having a spectrum stabilization stage and a measurement stage, when the detection device is in the spectrum stabilization stage, the detection device is located outside the shielding main body, and when the detection device is in the measurement stage, at least part of the detection device is located in the Marin cup. The technical scheme of the embodiment of the application is characterized in that the detection device is located outside the shielding main body in the spectrum stabilization stage, and at least part of the detection device is located in the Marin cup in the measurement stage, so that in the radioactivity detection mode, a natural radioactive nuclide can be used for spectrum stabilization, thereby avoiding the adverse effect of the shielding main body on spectrum stabilization and ensuring the accuracy of the activity detection result.

A method for determining the position (zk) or radioactive activity of gamma ray sources (Pk) of an object, wherein: b) interaction data, or 'electron recoil's, are acquired with a Compton camera during movement of the object; c) these electron recoil's (REi) are grouped into groups (Gj) relative to the same gamma ray; e) for each group and each possible sequence (Sjk) of electron recoil's in the group, a Compton cone (CCjk) is constructed and its intersections (Pjkl) with the object's axis (z) are calculated; f) for each group, for each possible sequence (Sjk) of electron recoil's, a weighted fraction (Rjkl) of gamma ray emitted for each intersection is estimated; g) from the weighted fractions, a radiation curve (CR) representing the activity density of the sources is established; h) their position or activity is determined from the radiation curve. Figure for the summary: Figure 6A

The present invention relates to the field of medical equipment technology, and discloses a dynamic detection method and system for the spatial distribution of Compton detectors. The method comprises determining a high-radioactivity region of a scanned object; determining a detector distribution spatial region relative to the scanned object; determining a radioactive attenuation degree based on the detector distribution spatial region; determining the number of Compton detectors to be arranged in the detector distribution spatial region based on the radioactive attenuation degree; and detecting the high-radioactivity region using the number of arranged Compton detectors; wherein the number of arranged Compton detectors increases with an increase in the radioactive attenuation degree. In this manner, the detection time can be shortened, image reconstruction artifacts can be reduced, and the difference in resolution of images reconstructed at different angles in three-dimensional space can be reduced, thereby reflecting a more realistic radionuclide distribution. Higher-quality radionuclide distribution images can be reconstructed at lower activity levels, providing selectable angle detection information for reconstructing three-dimensional images.

The invention relates to the technical field of system optimization, and discloses a federated learning-based performance collaborative optimization method and system for multiple sets of separation systems, and the method comprises the steps: obtaining the radioactivity value of the current batch of raw material liquid of each set of At-211 separation system, and carrying out the analysis of the radioactivity value based on a deep neural network, generating a working condition sensitive feature vector representing the influence of different batches of radioactive activity values on the performance of each separation system; through the working condition sensitive feature vector, the working condition attribute factor, the activity working condition vector and the collaborative optimization coefficient which are sequentially generated in each step, the influence of different batches of radioactive activity values can be accurately mined, the working condition association and difference of each system are determined, the differential aggregation weight fine adjustment is realized, the out-of-distribution generalization failure of cross-activity migration is avoided, and the system safety is improved. It is ensured that technological parameters adapt to the requirements of raw material solutions with different activities, the cooperative capacity of multiple At-211 separation systems is improved, and the separation efficiency and the product purity are improved.

The application provides a method and system for real-time evaluation of adsorbent performance for radioactive gas treatment, and relates to the technical field of environmental monitoring. The method comprises the following steps: deploying a plurality of radioactivity sensors, temperature sensors and humidity sensors at different depths in the axial and radial directions of the adsorbent bed layer, collecting the radioactivity concentration, temperature and humidity at each position in real time through the sensors; and arranging ventilation impedance sensors and organic volatile sensors at the inlet and outlet of the adsorption unit and inside the bed layer, and synchronously collecting the ventilation impedance and the concentration of organic volatiles. The application realizes the global three-dimensional monitoring of the adsorption performance of the adsorbent, the accurate correction of multiple factors and the automatic discrimination of the safety threshold.

The invention belongs to the technical field of radiation monitoring, and particularly relates to a method for correcting the detection efficiency of I-129 in a radioactive iodine sampling carbon box, the specific distribution depth of radioactive substances can be obtained by measuring the counting rate of a to-be-detected activated carbon powder sample, and then the detection efficiency value at the depth is obtained through a function relation. The method is used for I-129 radioactive activity concentration calculation. And the detection efficiency of each carbon box sample after correction can be quickly obtained. According to the technical scheme, calculation of complex distribution parameters can be avoided, a correction value can be rapidly obtained from a curve for an actual sample through the constructed correction curve and function, accurate measurement of the I-129 radioactive activity concentration in the iodine sampling carbon box is achieved, an effective basis is provided for index emission control and atmospheric environment safety evaluation, and the method is suitable for industrial production. And the method still retains non-destructive analysis of the carbon box sample, and is simple to operate and high in timeliness.

This invention relates to the field of radiation monitoring technology and discloses a method and system for monitoring radioactive separation and leakage prevention based on multi-sensor data. The method includes: Step 1, dividing the monitoring area and establishing a step-by-step reference model; Step 2, performing background counting measurements and baseline initialization; Step 3, simultaneously acquiring radiation, liquid level, pressure, and accumulated liquid data, determining the net radiation measurement value, and generating a unified spatiotemporal state vector; Step 4, determining the abnormal quantity of the liquid path and the liquid event judgment result; Step 5, determining the actual normalized activity distribution ratio, activity migration distortion, and activity exceeding the prohibited zone boundary, obtaining the radioactive trajectory anomaly judgment result; Step 6, obtaining the system joint state and determining the priority anomaly source segment; Step 7, determining the total measured activity of the system, the radioactive activity loss judgment result, and the monitoring result. This invention achieves automatic monitoring and location of liquid anomalies, radioactive migration anomalies, and radioactive leakage risks.

The invention relates to a real-time monitoring device and method for beta radioactive activity in fluid, and the device is connected with a radioactive waste liquid decay pool through an extraction pump, and is used for extracting the fluid in the radioactive waste liquid decay pool; the solid scintillation cell is installed at the downstream of the draw-off pump, solid scintillation particles are placed in the solid scintillation cell, the solid scintillation cell is a quartz glass cavity, the scintillation particles are of a spherical structure with the particle size of 50-300 micrometers, and beta particles in the fluid emit scintillation light through the solid scintillation particles; the triple photomultiplier is arranged around the side wall of the solid scintillation cell and is used for capturing scintillation light from three directions and converting the scintillation light into three electric signals; and the coincidence counting module is connected with the triple photomultiplier and is used for receiving the three electric signals for coincidence counting and determining the beta radioactivity. The probability of interaction between the beta particles and the scintillation material is improved, synchronous detection of the triple photomultiplier is realized, the detection precision is improved, and online monitoring of the fluidity of fluid is realized.

Method (100) for carrying out a release process of a, in particular at least partially radioactive, measured material (10), comprising the following steps: - Recording (111) preliminary activity data (11) of the sample (10) depending on a preliminary investigation (110), - Preparation of the preliminary evaluation (112) of the preliminary investigation (110) of the measured material (10), - Planning (113) a subsequent part of the release process depending on the preliminary evaluation (112) and on a method of carrying out a decision measurement (140) for the release (150) of the measured item (10), - Performing the decision measurement (140) in a clearance measurement facility (1), wherein the decision measurement (140) is carried out using the method in which a detection (141) and an evaluation (142) of a radioactive activity distribution in the measured material (10) is carried out, wherein the preparation of the preliminary evaluation (112) of the measured material (10) is carried out taking into account the preliminary activity data (11) with regard to a specification criterion (111.1) for specifying at least a part of the release process by the measured material (10), wherein the release process begins with the dismantling of the measured material (10) in a nuclear facility (8), wherein the preliminary investigation (110) takes place in the nuclear facility (8) to record (112) the preliminary activity data (11).